ML17277A640
| ML17277A640 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 07/31/1983 |
| From: | WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| To: | |
| Shared Package | |
| ML17277A639 | List: |
| References | |
| RTR-NUREG-0892, RTR-NUREG-892 IEB-79-01B, IEB-79-1B, NUDOCS 8307070013 | |
| Download: ML17277A640 (801) | |
Text
'83070700i3 830630 PDR ADOCX OS0003e7 A PDR
&Nj'-'2'V RED)',. 0588'N.VI RON@K.NTA'L EQU I PME NT:QUALI.FltAT'IQN RE.P.G.RT.
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EQUIPMENT QUAL,f'$-'II,'.:I',06f,"-- >>>>'4T RE PORT Volume 3 July 1983
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HNP-2 JUSTIFICATION FOR INTERIM OPERATION REPORT JULY 1983
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Page i JIO REPORT TABLE OF CONTENTS
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1.0 INTRODUCTION
2.0 RESULTS/CONCLUSIONS 2-1 3.0 SCOPE OF ANALYSIS 3-1 3.1 Accidents Creating a Harsh Environment 3-1 3.2 Post-Accident Environmental Conditions 3-1 3.3 Regulatory Guide 1.97 Requirements 3-2 3.4 'Impact of Non-Safety Equipment on Safety-Related 3 4 Equipment 4.0 SAFETY-RELATED SYSTEMS 4-1 4.1 Approach 4-1 4.2 Systems Reviewed 4-2 4.3 Correlation Between FSAR Table 3.2-1 and Systems 4-6 Reviewed 5.0
~ METHODOLOGY 5-1 5.1
~ Approach 5-1 5.2 Safe Shutdown Analysis 5-3 5.2.1 Assumptions 5-3 5.2.2 Accident Definition 5-4 .
5.2.3 Safety Sequence Analysis 5-5 5.2.4 Failure Modes and Effects Analysis 5-8 5.2.5 Use Code Definition 5-9 5.3 Minimum Required Set of Equipment 5-11 5.3.1 Assumptions 5-11 5.3.2 Preferred Safe Shutdown Paths 5-12 5.3.3 Selection of Minimum Set of Equipment 5-16 5.3.4 Results 5-16 5.4 Component-Specific Justifications 5-17 5.4.1 Approach 5-17 5.4.2 Cri teri a 5-18 5.4.3 Results 5<<20 TABLE A TABLE B EQUIPMENT JUST I FI CATIONS
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Page ii JIO REPORT LIST OF FIGURES 2.0 RESULTS/CONCLUSIONS Figure 2.1 RCIC Line Break Safety Sequence Diagram Figure 2.2 RWCU Line Break Safety Sequence Diagram Figure 2.3 AS Line Break Safety Sequence Diagram Figure 2.4 MS Line Break Safety Sequence Diagram Figure 2.5 RFW Line Break Safety Sequence Diagram Figure 2.6 Large Break LOCA (Recirc.) Safety Sequence Diagram Figure 2.7 Large Break LOCA (MS) Safety Sequence Diagram Figure 2.8 Small Break LOCA Safety Sequence Diagram Figure 2.9
~ Control Rod Drop Safety Sequence Diagram 5.0
~ METHODOLOGY Figure 5.1 Safety-Related Electrical Equipment Exposed to a Harsh Environment Figure 5.2 Sample Safety Function Path Diagram Figure 5.3 Sample Safety System Auxiliary Diagram Figure 5.4 Sample Safety Sequence Diagram Figure 5.5 Sample Failure Modes and Effects Analysis Figure 5.6 Unqualified Safety-Related Electrical Equipment Exposed to a Harsh Environment Figure 5.7 Preferred Safe Shutdown Paths (Table A)
Safety Sequence Diagram Figure 5.8 Alternate Safe Shutdown Paths (Table B)
Safety Sequence Diagram Figure 5.9 Safety-Related Electrical Equipment to Have
(}ualification Documentation Prior to Fuel Load or End of First Refueling
Page i ii JIO REPORT LIST OF TABLES SECTION 2.0 RESULTS/CONCLUSIONS Table 2.1 Accidents Analyzed Table 2.2 Accident Break Locations and Affected Areas Table 2.3 Systems Identified on Safety Sequence Oiagrams Table 2.4 Auxiliary Support Systems Nhich Help the. Safety System to Achieve its Function TABLE A Equipment on Preferred Safe Shutdown Paths TABLE 8 Equipment on Alternate Safe Shutdown Paths
Page l-l
1.0 INTRODUCTION
To obtain an operating license for Washington Public Power Supply System (Supply System) Nuclear Project Number 2 {WNP-2), the Supply System is required to provide documentation that establishes the environmental qualification of all safety-related electrical equipment. NUREG-0588, Category II, "Interim Staff Position on Environmental gualification of Safety-Related Electrical Equipment" provides the basis for determining the adequacy of the safety-related equipment's documentation.
The Environmental gualification Program for WNP-2 is in process, and most of the components have been shown to be qualified; yet it is unlikely that all safety-related electrical equipment's qualification will be documented prior to fuel load. The NRC Staff's final rule 10CFR50.49, "Environmental gualification of Electric Equipment Important to Safety for Nuclear Power Plants," states in paragraph (i) that the applicant for an operating license shall perform an analysis to ensure that the plant can be safely operated pending completion of environmental qualifi-cation. Therefore, a Justification for Interim Operation {JIO) analysis was performed which provides the results and the basis for the safe operation of WNP-2 until the Environmental gualification Program can be comp 1 eted.
The JIO analysis establishes that, upon documentation of the qualification or component-specific justification of a minimum set of safety-related electrical equipment, WNP-2 can be safely operated pending completion of the Environmental gualification Program.. This minimum set of safety-related electrical equipment consists of the equipment located in a harsh environment that is required to accomplish the following six safety functions:
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- 1. Emergency Reactor Shutdown.
- 2. Primary Containment Isolation.
- 3. Reactor Core Cooling.
- 4. Containment Integrity.
- 5. Core Residual Heat Removal.
- 6. Prevention of Significant Release of Radioactive Material to the Environment.
Accomplishing these six safety functions will ensure the safe shutdown of WNP-2. Safe shutdown includes accident mitigation as well as achieving and maintaining cold shutdown. The equipment in a single preferred path that accomplishes the required safety functions was selected as the minimum set requiring documentation of qualification or justification prior to fuel load of WNP-2. Category I and II. post-accident monitoring instrumentation required by Regulatory Guide 1.97 as well as the safety-related electrical equipment which perform or support the safety functions are included in the minimum set. The redundant safety-related electr ical equipment required for defense in depth, diversity of function, and electrical separation will have qualification documentation prior to the completion of the first refueling outage or November 30, 1985.
The JIO analysis for WNP-2 was accomplished in six steps:
Step 1 identified the FSN Chapter 15 accidents that potentially cause a harsh environment inside the primary containment or reactor building.
These accidents include three Loss-of-Coolant Accidents (LOCAs), four High Energy Line Breaks (HELBs), and the Control Rod Orop Accident.
Break locations for LOCAs were not necessary for this analysis since the effect inside the primary containment is not dependent on the break location. Break locations for the HELBs, and the equipment which could be exposed to the harsh environment, were determined and are discussed in Section 2.0.
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Page 1-3 Step 2 determined the environmental conditions associated with the LOCA and HELB accidents in Step l. (The radiation conditions resulting from a LOCA were used to envelop the Control Rod Orop Accident.) These environmental conditions provide the basis for the qualification program and are used to determine the qualification status of the safety-related electrical equipment in Step 5.
Step 3 consisted of the performance of a Safety Sequence Analysis (SSA) for each of the LOCA and HELB accidents defined in Step 1. The SSA identified the safety systems, and their associated equipment, required to achieve each safety function. The analysis was performed for equipment located inside the reactor building and primary containment potentially exposed to a harsh environment. Each SSA identified all credible and redundant paths to accomplish each safety function. It also described the system actions, inputs, and interlocks. The SSA is discussed in Section 5.2.3. For the Control Rod Orop Accident, a Safety Sequence Oiagram was constructed based on the Protection Sequence Oiagram given in the MNP-2 FSAR .
In Step 4, A Failure Modes and Effects Analysis (FMEA) was performed on the safety-related electrical equipment not required to function for safe shutdown. This analysis identified the safety-related electrical equipment that could fail in a manner detrimental to the safe shutdown of the plant. Section 5.2.4 describes this analysis. Since this equipment must not fail in a manner detrimental to plant safety, documentation of its capability to withstand the potentially harsh environment is provided. Lastly, equipment whose failure under accident conditions is not detrimental to plant safety has been determined. This equipment need not be qualified for any accident environment and is not included in this report.
Page 1-4 In Step 5, the single-path minimum set of safety-related electrical equipment r equired to accomplish the six safety functions was identified.
. This set of equipment, which includes equipment identified in Step 3, and the equipment identified in Step 4 requiring qualification, will have if complete qual ication documentation or component-specific justif ication provided prior to fuel load. This will ensure one qualified or justified path to safe shutdown for all the accidents identified in Step 1. The selection of the equipment on this preferred safe shutdown path is discussed in Section 5.3.3. The equipment to be documented as qualified or justified prior to fuel load are included in Table A. The balance of the safety-related electrical equipment identified in Step 3 and Step 4 are included in Table B and will be documented as qualified prior to the completion of the first refueling outage.
In Step 6, component-specific justifications were developed for the equipment on the preferred safe shutdown path whose qualification documentation would not likely be completed pr ior to fuel load. The justifications were developed based on the operability requirements of the equipment and criteria derived from the final NRC rule 10CFR50.49, paragraph (i). Assurance that HNP-2 can be safely operated pending completion of the Environmental gualification Program is demonstrated by these component-specif ic justif icati ons.
Page 2-1 2.0 RESULTS/CONCLUS IONS Oocumentation of the environmental qualification or component-specific justification, for the minimum set of safety-related electrical equipment identified by this analysis, will ensure the capability of safely mitigating the WNP-2 harsh environment producing accidents. The analysis performed to document this conclusion, as required by the NRC's final rule 10CFR50.49, "Environmental gualification of Electric Equipment Important to Safety for Nuclear Power Plants", will ensure that the plant can be safely operated pending completion of environmental qualifications.
This analysis determined all viable safe shutdown paths which could accomplish the required safety functions under accident conditions. A single preferred path was then selected such that the equipment on that path comprise the minimum set requiring documentation of qualification or justification prior to fuel load. This report presents the qualification documentation or component-specific justification for the preferred safe shutdown path equipment. Table A identifies this minimum set of equipment. The remaining safety-related electrical equipment in the alternate shutdown paths are listed in Table B and will have complete qualification documentation prior to the end of the first refueling outage. Interim operation is justified since a fully qualified or justified preferred safe shutdown path has been identifed for each safety function. The preferred path equipment is necessary and sufficient to ensure safe shutdown. Results from major tasks in the analysis are described and presented below.
The FSAR Chapter 15 accidents that potentially cause a harsh environment in the primary containment or reactor building are identified in Table 2.1. The areas of the plant affected by each break are identified in Table 2.2 by zone. The methodology for determining the accidents and areas affected is discussed in Section 5.2.
Page 2-2 Safety Sequence Diagrams (SSDs) were prepared for each accident. They identify the systems required to accomplish the necessary safety functions for each accident and are included as Figures 2.1 through 2.9.
Table 2.3 lists the safety systems delineated on the SSDs, and Table 2.4 identifies the auxiliary support systems necessary for all safety systems.
Page 2-3 TABLE 2.1 ACCIDENTS CONS IDERED
- 1. Reactor Building - High Energy Line Breaks Reactor Core Isolation Cooling (RCIC) Steam Line Bream, Accident Codes A, B, C, E Auxiliary Steam System Pipe Break Accident Codes D, M Reactor Water Clean-up (RWCU) Line Breaks Accident Codes F, G, H, I, J, K, L Main Steam Tunnel (either steam or feedwater line) Pipe Break Accident Codes N, 0
- 2. Primary Containment Loss-of-Coolant Accidents Reactor Recirculation Line (RRC) Break (Large break)
Accident Code P Main Steam Line Break (26")
Accident Code g Small Main Steam Line (2" or less) Break Accident Code R
- 3. Primary Containment and parts of Reactor Building Control Rod Drop Accident Accident Code S
'i TABLE 2.2 ACCIDENT LOCATIONS AND AFFECTED AREAS Page 2-4 Acci-dent Location(s ) Acci dent Pr ofi 1 es Codes Accident T e/Location Affected Pressure/Tem erature)
A HELB - 4" RCIC (13) - 4 R422L, R441I 5,6 RCIC Pump Room HELB - 4" RCIC (13) - 4 R422L, R441I Room Above RCIC Room C HELB - 4" RCIC (13) - 4 R422M, R441J Room Above RN-2C D HELB 4 AS (1 1 ) 2 R471Ae B De J 9,10,11 Southeast Open Floor Area R501, B, F, H, K, g R522; C, H, J, K, P HELB - 4" RCIC (13) - 4 R501P, R510S 12,13 TIP Room F HELB - 6t RWCU (2) - 4 R501P, R510S 14,15 Room Above TIP Room G HELB - 6" RWCU (2) - 4 R5220 16 Valve Room "N" of Cont.
H HELB - 4" RMCU (1) - 4 R522F, G 17,18,19 RWCU Pump Rooms I HELB - 6" RWCU (1) - 4 R522F, G 20,21,22 Valve Room Above RWCU Pumps R522B, B 23,24 J HELB - 6't RMCU (1) - 4 R 25,26,27 RWCU HX Room R548P
- 6" (2) -
572'548B, K HELB RMCU 4 R548g 28 Valve Room "N" of Cont.
L HELB - 6" RMCU (1) - 4 R54%, H, K 29,30 Valve Room "S" of Cont.
M HELB - 3" AS (11) - 2 R572B,C, F, G, N 31.32 Southeast Floor Area R548C, G, K, M, P, R El.
Page 2-5 TABLE 2.2 (Continued)
ACCIDENT LOCATIONS AND AFFECTED AREAS Acci-dent Location(s ) Pccident Profiles Codes Accident T e/Location Affected Pressure/Tem erature)
N HELB - 26" MS R5010 0 HELB - 24" RFW R5010 P LOCA - 24" RRC Containment 1,2 q LOCA - 264 NS Containment 1,2 R LOCA - Small Steam Line Containment 1,2 S Control Rod Drop Accident Containment, R5010 None LOCA (P,'g,R) - Primary Reactor Building (except 4 Containment Class 1E motor control center rooms)
Page 2-6 TABLE 2.3 SYSTEMS IDENTIFIED ON SAFETY SEQUENCE DIAGRNS Abbreviation S stem Naos CAC: Containment Atmosphere Control CAS: Control Air System CEP: Containment Purge Exhaust CIA: Containment Instruaent Air IXS: Containment Monitoring System CRA: Containment Recirculation Air
$ 0: Control Rod Drive (Hydraulic )
CSP: Containment Purge Supply CVB: Containment Vacuum Breaker E ~
El ectri cal EDR: Equipment Drains Radioactive FDR: Floor Drains Radioactive FPC: Fuel Pool Cooling HPCS: Hi gh Pressure Core Spray HY: @draulic Control LD: Leak Detection LPCS: Low Pressure Core Spray LPRM: Local Power Range Monitor MS: Main Steam MSLC: Main Steam Leakage Control PI: Process Instrumentation PSR: Process Sampling Radioactive System RCC: Reactor Building Closed Cooling RCIC: Reactor Core Isolation Cooling REA Reactor Building Exhaust Ait RFM: Reactor Feedwater RHR: Residual Heat Removal ROA: Reactor Building Outside Air RPS: Reactor Protection System RRA: Reactor Building Return Air RRC: Reactor Recirculation RWCU: Reactor Mater Clean-Up
Page 2-7 TABLE 2.3 SYSTEMS IOENTIFIEO ON SAFETY SEQUENCE OIAGRAMS (Continued)
Abbreviation S stem Name SGT: Standby Gas Treatment SLC: Standoy Liquid Control SPTM: Suppression Pool Temperature Monitor SRM: Source Range Monitor SW: Standby Service Water TIP: Traversing In-core Probe
Page 2-8 TABLE 2.4 AUXII IARY SUPPORT SYSTEMS WHICH HELP THE SAFETY SYSTEM TO ACHIEVE ITS FUNCTION Safety Auxi1 i ary
~Sstem ~Sstem CAC E, RRA, SW CAS E CEP E CIA E
(ÃS E, RRA, SW CRA E CRD E CSP E EDR E FOR E HP CS RRA HY E LD E LPCS E, RRA, SW MS(l ) E, CIA MSLC E PI E PSR E RCC E RCI C E REA E RFW E RHR E, RRA, SW ROA E RPS E RRC E RWCU E SGT(2) E SLC E SPTM E SRM E TIP E (l) RHR (LPCI) or LPCS Interlocked with ADS valves.
(2) REA Differential Pressure e Transmitter signal is used to control reactor ouilding pressur e.
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PROFILE 4 REACTOR BUILDING SERVICE PRC CONDITIONS DUE TO A LOCA IN PRIMARY CONTAINMENT .APERTURE
'b, CARD Rceetor iscstidi~ 'fcmperetssre
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240 220 200 Pmax = 15.45 psia RHmax = 100% 41 min.
180 I-160 W
I-140 120 100 1.0 10.0 100 1000 10000 TIME (SEC)
PROFILE 5. 4" RCIC LINE BREAK IN RCIC PUMP ROOM (EL 422). RESPONSE IN RCIC PUMP ROOM (EL 422). 821772. l 7 A
240 220 200 180 I-160 I-140 120 100 1.0 10.0 100 1000 10000 TIME (SEC)
PROFILE 6. 4" RCIC LINE BREAK IN RCIC PUMP ROOM (EL 422). RESPONSE IN ROOM ABOVE RCIC PUMP ROOM (EL 444).
210 Pmax 15.41 psia RHmax = 100% 47min 190 E'70 K
LU I-150 130 110 100 0.1 1.0 10.0 100.0 1000.0 10000 TIME (SEC)
PROFILE 7. BREAK OF 4" RCIC LINE IN ROOM ABOVE RCIC PUMP ROM (EL 444).
RESPONSE IN ROOM ABOVE RCIC PUMP ROOM (EL 444) (A) AND RCIC 8217%46A PUMP ROOM (EL 422) (B).
230 220 200 Pmax = 15.1 psia 180 = 95.8% 15 min RHmax I-K 160 I-140 B
120 r L 100 1.0 10 100 1000 10000 TIME (SEC)
PROFILE 8. BREAK OF 4" RCIC LINE IN ROOM ABOVE RHR PUMP 2C ROOM (EL 444).
RESPONSE IN ROOM ABOVE RHR PUMP 2C ROOM (EL 444) (A) AND RHR PUMP 2C ROOM (EL 422) (B).
170 160 150 Pmax 15.22 psia RHmax 100% 83 min 140 130 120 110 100 1.0 10.0 100.0 1000.0 10000 TIME (SEC)
PROFILE 9. 4" AS LINE BREAK IN THE SOUTHEAST OPEN FLOOR AREA (EL 471).
82$ 77247A RESPONSE IN ALL OPEN FLOOR AREA (EL 471).
170 160 150 140 130 Pmax = 14.7psia RHmax 80%
120 110 100 10000 1.0 10.0 100.0 1000.0 TIME (SEC)
PROFILE 10. 4" AS LINE BREAK IN THE SOUTHEAST OPEN FLOOR AREA (EL 471).
RESPONSE IN ALL OPEN FLOOR AREA (EL 501). S21'172.19A
170 160 150 140 I-Q.
130 I-Pmax = 14.7 psia RHmax 80%
120 110 100 1.0 10.0 100.0 1000.0 10000 TIME (SEC)
PROFILE 11. 4" AS LINE BREAK IN THE SOUTHEAST OPEN FLOOR AREA (EL 471).
RESPONSE IN ALL OPEN FLOOR AREA (EL 522). 821712.18'
360 350 340 320 300 280 E'60 Pmax = 14.86 Psia 240 RHmax = 100% 6.5 min 220 200 180 160 140 120 100 1.0 10.0 100.0 1000.0 3000 TIME (SEC)
PROFILE 12. 4" RCIC LINE BREAK IN T.I.P. ROOM (EL 501). RESPONSE IN T.I.P. ROOM (EL 501).
380 360 340 320 300 280 260 240 220 I- Pmax = 14.86 PaIa 200 RHmax = 100% 6.5 min 180 160 140 120 100 0 1.0 10.0 100.0 1000.0 3000 TIME (SEC)
PROFILE 13. 4" RCIC LINE BREAK IN T.I.P. ROOM (EL 501). RESPONSE IN VALVE ROOM ABOVE T.I.P. ROOM (EL 510.5).
240 230 220 210 200 Pmax = 15.1 psia RHmax 100% 61 min 190 E'80 170 160 I-150 140 130 120 110 100 0.1 1.0 10 100 1000 10000 TIME (SEC)
PROFILE 14. 6" RWCU LINE BREAK IN VALVE ROOM ABOVE T.I.P. ROOM (EL 510.5).
RESPONSE IN VALVE ROOM ABOVE T.I.P. ROOM (EL 501).
240 230 220 210 200 190 E'80 Pmax RHmax
= 15.1 psia 100% 61 mIn 170 160 150 140 130 120 110 100 0.1 1.0 10 100 1000 10000 TIME (SEC)
PROFILE 15. 6" RWCU LINE BREAK IN VALVE ROOM ABOVE T.I.P. ROOM (EL 510 5)
RESPONSE IN VALVE ROOM ABOVE T.I.P. ROOM (EL 510.5).
230 220 200 Pmax 16.23 psia RHmax 100% 50 min Z'80 I-160 140 120 100 0.1 1.0 10 100 1000 10000 TIME (SEC)
PROFILE 16. 6" RWCU LINE BREAK IN THE VALVE ROOM NORTH OF CONTAINMENT (EL 522) RESPONSE IN THE VALVE ROOM NORTH OF CONTAINMENT (EL 522).
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245 235 225 Pmax = 33.5 psIa 215 RHmax 100% 17 min 205 195 185 175 165 155 145 135 125 115 105 0.1 1.0 10 100 1000 10000 TIME (SEC)
PROFILE 17. 4" RWCU LINE BREAK IN RWCU PUMP ROOMS (EL 522). RESPONSE IN RWCU PUMP ROOMS (EL 522). 821772 12A
245 235 225 215 205 195 f'85 Pmax = 16.5 psia RHmax 100% 17 min 175 165 155 145 135 125 115 105 0.1 1.0 10 100 1000 10000 TIME (SEC)
PROFILE 18. 4" RWCU LINE BREAK IN RWCU PUMP ROOMS (EL 522). RESPONSE IN
-VALVE ROOM SOUTH OF CONTAINMENT (EL 522).
245 235 225 215 205 Pmax 19.5 psia 195 RHmax 100% 17 min 185 175 165 155 145 135 125 115 105 0.1 1.0 10 100 1000 10000 TIME (SEC)
PROFILE 19. 4" RWCU LINE BREAK IN RWCU PUMP ROOMS (EL 522). RESPONSE IN VALVE ROOM ABOVE RWCU PUMP ROOMS (EL 535).
230 210 o 190 170 Pmax = 27.8psia RHmax 100% 17 min 150 130 110 100 0.1 1.0 10 100 1000 10000 TIME (SEC)
PROFILE 20. 6" RWCU LINE BREAK IN VALVE ROOM ABOVE RWCU PUMP ROOMS (EL 535). RESPONSE IN RWCU PUMP ROOMS (EL 522). eamon
230 210 E'
190 LLI 170 I- = 20.6psia Pmax RHmax 100% 1'l min 150 130 110 100 0.1 1.0 10 100 1000 10000 TIME (SEC)
PROFILE 21. 6" RWCU LINE BREAK IN VALVE ROOM ABOVE RWCU PUMP ROOMS (EL 535). RESPONSE IN VALVE ROOM SOUTH OF CONTAINMENT (EL 522).
230 210 190 I-170 I- Pmax = 27.9 psIa 100% 17 min RHmax 150 130 110 100 0.1 1.0 10 100 1000 10000 TIME (SEC)
PROFILE 22. 6" RWCU LINE BREAK IN VALVE ROOM ABOVE RWCU PUMP ROOMS (EL 535). RESPONSE IN VALVE ROOM ABOVE PUMP ROOMS (EL 535).
230 B
210
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/ RHmax = 100% 61 mIn r
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110 100 0.1 1.0 10 100 1000 10000 TIME (SEC)
PROFILE 23. 6" RWCU LINE BREAK IN VALVE ROOM ABOVE RWCU PUMP ROOMS (EL 535). RESPONSE IN SOUTHEAST (A), SOUTH (B), SOUTHWEST (C) AREAS (EL 522).
230 210 190 I- 15.0 psIa Pmax K RHmax 100% 61 min 170 I-150 B
130
/ A 110 /
100 0.1 1.0 10 100 1000 10000 TIME (SEC)
PROFILE 24. 6" RWCU LINE BREAK IN VALVE ROOM ABOVE RWCU PUMP ROOMS (EL 535). RESPONSE IN CRD EAST (A), CRD EAST WALKWAY(8) AREAS (EL 522).
230 220 210 Pmax = 18.1 psia RHmax = 100% 83 min 200 190 180 E'70 160 150 140 130 120 110 100 0.1 1.0 10 100 1000 10000 TIME (SEC)
PROFILE 25. 6" RWCU LINE BREAK IN THE RWCU HEAT EXCHANGER ROOM (EL 548) ~
821772 1SA RESPONSE IN RWCU HEAT EXCHANGER ROOM (EL 548).
230 220 210 200 190 Pmax 14.7 psia RHmax = 100% 48 min 180 170 160 150 140 130 120 110 100 0.1 1.0 10 100 1000 10000 TIME (SEC)
PROFILE 26. 6" RWCU LINE BREAK IN THE RWCU HEAT EXCHANGER ROOM (EL 548). 82lll2.14A RESPONSE IN NE AREA OF EL 548.
240 220 200 160 Pmax = 14.7psia RHmax 100% 33 min 140 120 100 0.1 1.0 10 100 1000 10000 TIME (SEC)
PROFILE 27. 6" RWCU LINE BREAK IN THE RWCU HEAT EXCHANGER ROOM (EL 548).
RESPONSE IN NW AREA OF EL 548.
240 Pmax = 15.95 psIa 220 RHmax 100% 13 min 200 f
180 K
I-160 140 120 100 0.1 1.0 10 100 1000 10000 TIME (SEC)
PROFILE 28. 6" RWCU LINE BREAK IN VALVE ROOM NORTH OF CONTAINMENT (EL 548).
RESPONSE IN VALVE ROOM NORTH OF CONTAINMENT (EL 548).
230 210 Pmax 17.7 psia RHmax 100% 51 min.
190 150 130 110 100 0.1 1.0 10 100 1000 10,000 TIME (SEC)
PROFILE 29. 6" RWCU LINE BREAK IN VALVE ROOM SOUTH OF CONTAINMENT (EL 548).
RESPONSE IN VALVE ROOM SOUTH OF CONTAINMENT (EL 548).
230 210 190 170 K
uj I-150 Pmax 14.7 psia RHmax 100% 51 min.
130 110 100 10,000 0.1 1.0 10 100 1000 TIME (SEC)
PROFILE 30. 6" RWCU LINE BREAK IN VALVE ROOM NORTH OF CONTAINMENT (EL 548).
RESPONSE IN SOUTHWEST AREA (EL 548).
170 160 15.22 psIa Pmax R"max = 100% 58 min 120 100 1.0 10.0 100.0 1000.0 10000 TIME (SEC)
PROFILE 31. 3" AS LINE BREAK IN SOUTHEAST OPEN FLOOR AREA (EL 572). RESPONSE IN ALL OPEN FLOOR AREA (EL 572).
170 160 f 140 K
W Pmax 14.7 psia RHmax (80%
120 100 1.0 10.0 100.0 1000.0 10000 TIME (SEC)
PROFILE 32. 3" AS LINE BREAK IN SOUTHEAST OPEN FLOOR AREA (EL 572). RESPONSE IN ALL OPEN FLOOR AREA (EL 548) s ( El 606) .
l' I
I I
J 1 l
E e
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XW 0%a<<VR REACTOR (XXLDING WAS CALCDLATKD;>NMi ~3 2 INDICATIVE GF ACTVAL ZONE, ~ <<33 ! SAFETY, RELATED EVENT Y ZONES
<<H.3> +K> >HI Y I I
LPCS WATER ( 'HPC~ I ZONE C FDR-Wl I LPCS PUMP HPCS WATER H SIP<<P PIP P LEG PUMP I I I)
(LPCSIP LEG PUMP '(MPCS<<P<1) i'<<DP. P AA>/ ~EACTOR BLDG LP~P/I LO-TE-279 (LPCS P 2> (HPCS P'3<< (FDR-P.48>/ (ED~%) EGVIPMKNT ERA'Ã !
HI'AT KXC<<<<ANGER (cDR <<<<X 2 ~ 1.1xl06 rads LO-TK-270 I !
i.I/
LPCS-FCY-11+
j
<< CRD <<O IA LPCS P-I+ QSUBZONE JI Sl IRA ELEV CGNTRDL ROD DRIVE Lpcse 2+ PSR-Y-003/A PV MP ,I 2.lxl06 radS D +V(SIR. RACX E~ II (CIO-P "IA>
H PCS'<<((MP LSCLERMP CRD IR I> FIR-LS-I5 PSR-Y49/A JRQM~ I,
<N<< (4.> I
~ I ZONE ZONE K Ox> IR 'r D'P&O-I2 ~ Rc lees-9s 5(MP PIP(P (FDR P 3 ~CONTROL RGD DRIVE
~ 1.5xl05 rads PVM P ~ 1.6xl06 rads RHR PUMP rRD P IR)
(RHR~2C)
RQQM IH<<D
!. ~
IPCSe-I<< RCIC-PS-9A RMR WATER HPCS-P-3+ Rc lees-9 B LEG PVMP (RHR P 3) IIICS-Y-l+
HPCS-Y-12+
ZONE L I
~ Rc IC~ I6 REACTOR BLDG FDR-LS-l6 ~ l.2xlDT rads CGNDENSA.K SUPP' PUMP (CAPP 3>
6 38 Tb/RI35+
ZONE E RCIC~RO- I+
2-la-6I+
RCIC-OT-I a
~ 9.la 103 rads aclce-3 RCIC Y-Ia
~AVX N > 2-IR-6 It RCIC MISC. J5(MP I52/2! RCIC-Y I&
EGVIPMKNT RC Ic-YQ5+
(RCr TAN) ZONE l (Rcr.p-a> RCIC-Y-2&
(lrr p 2) R>Rwe/2B (BC'<<x I) F ~ 2.5xl06 rads RCIC-YA+
PCIC PVMP (RCr P I)
A TVRBR<<K (RCIC.DT I) RADWAS K BLDG a>R-P-29+
R>R-Y45+
PRC L CGIDENSA svppLY pvGI I RCIC-Y-Sla RCIC ~ << (COND ~1 WATER LEG PUMP (RCr P 3)
FDR-LS42
// CDNDKNSA, E FID BACXWasH LD-TK DA LD-TE-27C LO-TE-Sl CARD PUMP (COND ~ 5<< LO-TE-5b I
S(PPIIKSSIGN RCIC LS-'ll POOL CLEAN VP PUB (PPC P 3>
f SIEIZONE 11 PSR-Y-003/B RCIC~3 Rc lees-'I She avoca re 0<
~ 2.6x106 rads RC ICeS-3l ifkpertmre Card RHR PUMP RC IC-PS-7 STAIR RGGM R 2 G P SR-Y-003/b Rc IC-TI5-IOK i) KLEVLR>R I STAIR SUMP PIP<<P <<
(FCR P >8)
ZONE J ZONE M R>6(&-P/2A ~ R)E(~/2C
~ 2.0xl06 rads ~ 1.9xl06 rads GENERAL NOTES:
I ~22 St&ET 2 OF 2 FOR CORIPOIKIITS EL'22'-3'>B(-P-2A+
SEE DWCL RHR PVMP RHR PUMP SUM<<<<P V<<<<D OF 12STED ~SITES. (RHR.P 2A) (RHR.P 2a) (FDR-P 2> a>Re-2c+
- 2. ~ S)ENTS'>ES WORST TAROET FROll CALCULATION. RHR Y 6At a>R-P-3>
- 2. RR E>EICTPKS NOT SISTALLED EOW%KIIT THAT HAS NOT %%% StATIALLYLOCATED. REACTOR BUlt:DNG 8807OYOOIS -OZ
- d. RI<<ALL
~
- d. ~ OENTPKS WORST ThRKY QAl(MA DOSE N 8 IIONTH DEPOT ACCE)ENT DOSE R 8 SRNTH ARSON%
ACCE)ENT E~IT
+ 40 YEAR NOI%IAL OPERATIONB DOSE.
RECER(ES APPRO>(S(ATELY SAl% DOSE, FDR-LS-I 3 RADIATlON ZONE MAP >(AI(. NONE REACTOR BUILI3(NG EL.
- 4. EO~NT tART NNCIERS FOLLOWED PUBLIC POWER 422-3'ASHINGTON RM<<N<<a IKk
~ Y + ARE C~OSITE EOUIPQENT.
~ONES Rftl%SENT ECWIPSIENT CROUP>NQS *ND SUPPLY SYSTEM M-422 S NOT A PHYSICAL RED>ON NI SPACE.
NNO I (V2 v S
I ga h c A p I pa g."
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ICAL&
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ac RCIC~T/Zb le<~
aclc~r/b
~/2 LPCS-PM+
'c RC IC445t RCIC~5 RCIC<~
ZONE 0 IC~
~/I HPCS-P-I+
H PCS-P<<3+
ac le-s+
RCIC~Y/5 aclc-Y %
HPCS~P/3 RCIC~V/sb H PCS-Y>>l+
HP~I Z(XK M
. HPCS-V 12+
RHRW-Zc+
IPCS~IZ IR~/Zc RIR P-3+
ZONE E kHt~P/3 E IR-61+
Eca-spY-I 9 FW-SPT-3 ZONE I RIR-P-Zb+
RHt~/Zb ae-we+
RIVt~bb ZONE J RSI-P-ZA+
R%&W/ZA RN-YW+
R%~
ZONE L
. E-Jb-rb/RA39 RCICMV-CRI PRC" 3 RCICALT~
RC ICC0H0- le RC IC-LS-10 APERTURE RC Icier-It RCIC-CHIR-C002 CARD .,y Rc le<ID-C002 RCIC-SE-COOZ RCIC-SS-I RCIC-SMÃQ RC IC4-3+
RCI C-II-P/3 kpgg~Q RCIC-V 'I+
RCIC~I
>neve, C'q" RCICms-T/I RC IC-V-19 RCICM-10 COMPONENT EOUIPMENT LIST FOR scut HOHE COMPOSITE EOUIPMENT.SHOWN OH RADIATION ZONE MAP, REACTOR BUILDIHG EL. 422'-3" WASHINGTON PUBLIC POWER SUPPLY SYSTEM M-422 2
&CD 2 EY
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SAFETY RELATEO EQUIPMENT INDICATES AREA WHERE DOSE RATE'UTSIDE OF REACTOR BURZWO WAS CALCULATEIX
- BY ZONES NOT PIDICATNE OF ACTUAL ZONE, FL3 H.7 K N; NB ZONE B CD-SPY-6A RCIC Y-lap
~ U~I2 CD SPY~ RCIC Y-22p 0 ~ 1 Sxl06 rads LtCS-Y-l+
CD SPY FA C IA-SPY-28 CIA-SPY-SA RCIC-Y-31+
RCIC-Y-Sap RCIC-Ykap
+,;.Ii F ~ ~ ~
LtCS-Y-12+ RRA-f C~
CIA-SPY'IA-SPV-9A RRA-FC-5+
SH-Y~ CIA-SPY-9b STAIR ELEV. LO-TE~
LO-TK~
A FAN CCRL VNIT OD H/ f SUSZONE 01 LO-TE-SA IRRUFC 4> ~
" PSR V Xab/I LO-TENS
~ I.1x105 rads RRA~R/6
~. .4.1, ,,
CIA tx-IA ZONE C SU-V-34 THRV 1SA p HtCS-TT3-23 PSR-V-Xbs/1 FAN COIL VNIT IRRA FC I) l.lxl06 rads PSR-V-X88/2 ZOPF J D r ~ Rig~21 EOR-Y-l 9+
~ 3.lxlO rads
'j 8/ gear EIVI-V-20p FOR-Y-3+
ZONE F 100~648 RTSI-FCV-64CP
//// Csp Tx 1 FIR-Yap 1.2x10 rads RHR-V-21+
THRV 10
~~
LI ~ ! r/ r ItCS-Y-1 0+ RHt Y<p
~ rlr HPCS-Y 11+
'I ~ R%-FEY~/ RRA-FC-1+
II ItCS-Y-1 Sp RHR-Y-lb+ SV Y-24C+
//
HPCS-V-23+
RRA-FC-3p RRA-FC-4+
SV V 2lb+
t SH-Y-54+ 015 LT-I IIRNCATES AREA wttERE DOSE
!'t' 4 B ltCS-LS-lA RATE OUTStDE OF REACTOR sx CRS-LT 2 Hpcs-LS-la I 8VEONO WAS CAI.CIJLATED. RRA~FR/4 HPCS LS 28 ItCS-LS-2A I NOT RIDICATnrE OF ACTUAL ZONE.
I LO-TE-188 RC IC-LS-15A I
I LO-TE-180 RCIC LS-158 I
. ~ f//f!t CIA 11 ISB THRU SB ZONE D
/ LO-TE-28A RRAQHS-FK/I I ~o 8 2xl0 radS I LO-TE-28C
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CIA-SPY-11A 'ZONE G
~ RIVT~4A CIA-SPY 118 I I
I r/ CIA-SPV-12A ',
~ 9.9xl05 raos 4-I I
r ~CIA TII-RB THRU 18 CIA-SPY 128 I
I FAN COIL UNIT l CIA-SPY 13A FPC-Y-153+
I I
L Ft RA/FBI
//
1A 3' CIA-SPY-138 CIA-SPV-14A CIA-Spv-1 la FPC-Y 154p FPC-V-156+
RHR-FCY44A/
/ ~ FAN COta UNIT FAN CO'I. UNIT IRRA FC.SI
'F CIA-SPV 15A CIA-SPV-158 RIVI-Y-al/
RRA-FC-2+
APERTURE PO~t/ Q CIA SPV 168 CIA-SPV-llS 5II-V-24A+
CARD 51A IR CIA-SPV-188 STA:R CIA-SPV-198 LO-TE-18A t
I;Ii 'rr c" CIA-SPV 1A LO-TE-18C
/// t/
1
' ~ / I/ ~ / CIA-SPV-lb L 0-TE-28S CIA-SPY-2A LO-TE-280 CIA-SPY-28 RRA4015 FH/2 CIA-SPV-3A kV~EEMe 0 REACTOR BUILDING EL. 441'-0 CIA-SPV-38 I CIA-SPY-4A I ZONE I LO-TE&b C IA-SP Y%8 l.OxlO rads GENERAL NOTES: CIA-SPV SA ; ~
II 1.op~,oo, $ , ARE IDENTIFIED IN GENERAL NOTES CIA-SPY 58 2,3,4,587 ON DRAWING M-422 SHEET 1. RADIATION ZONE MAP ssAIL NONE
- 2. SEE DWG. M-441 SHEET 2 OF 2 FOR COMPONENTS REACTOR BUILDING EL. 441-0 OF LISTED COMPOSITES. WASHINGTION PUBUC POWER SUPPLY SYSTEM M 441 w d DRTT I
I
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40 C,",9, J ~ \
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11 ZONE 8 ZONE I LPCS-V-lt RCIC V-19+
LPCS~I ~ 'RC I C~ 19 LPCS-Y-12t RCIC-V-ZZ+
LPCS~IZ RCIC~ZZ RRA-FC-5+ RCIC-V 31+
ARAN FH/5 RCIC~31 SH-V-44t RCIC-Y 59+
SH~44 RCIC~-59 RCIC-V-69+
RCIC~-69 ZONE C RRA-FC<t EN-Y-19+ RRA+-F H/6 EN-POS-V/19 EN-V-ZO+
ECR-POS-V/20 FN-V-3+ ZONE J RIR-FCV-64ct F N-POS-V/3 RHR-HO-64C FN Y4t FOR-POS-V/4 RIR-V-ZI t HPCS-V-10t RHR~Z I HPCS-HO- IO RIR-Y-4ct HPCS-Y-11t RHR~4C HPCS-HO-11 RRA-FC-1+
HPCS-V-15t RRA~FH/I HPCS~ IS SH-V-24ct HPCS-V-23t SH-HO-24C HPCS-HO-23 RRA-FC-4t RRA-H-FH/4 SH-V-54t MO-54 ZONE F RIR-FCV-648+
RMO-648 RIFI-V-48t RIR&0-48 RRA-FC-3t RRA~F8/3 5'N-V-248 t SWO 248 ZONE G fPC-V-153+
FPCA 153 FPC-V>>154t fPc~-154 FPC-V-156+
fPC~156 PRC RIFI-FCV-64At RARA-'64A APERTURE RIR-Y-4At RHIM-44 CARD RRA-FC-2+
RRA& FH/2 SH-V-24At SQM 24A COMPONENT EOUIPMENT LIST FOR scuci NONE COMPOSITE EOUIPMENT SHOWN ON RADIATION ZONE MAP REACTOR BUILDING EL. 441'-0" WASHINGTON PUBLIC POWER SUPPLY SYSTEM M-441 S
W j I
~ ~
I
t 5
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~IAFETY RELATED EQUIPMENT BY ZONES 1
ZONE A 0 SuBZOXE O1 ZONE I CAC EIO FCV/38
~,1 RIR~27A P SR-Y-XSZ/ I RCIC~86 ~ 9.1xlO rads
!) 7) ~ 5.0xl05 rats I a I.lxl06 rads ~ 4.8xl06 rads I EACH CY-38+
E-IR-POOI+ PSR-V-XSZ/I RCI C-Y-l 10+ CAC Y 17+
I E-IRW9) PSR-V X82/2 RC I C-V-113+
'f ~ .I 34) ) RHt-Y-27AD X82/7 RCIC-V<8)
PSR Y 5!A>
Kj EEEV 5!AIR PSR V XSZ/8 ACIO NDCS STSTEI)
C;- PI Y X269 PSR Y X83/ I PSR-Y XS3/2 ZONE J ttSLC~t IastR RACK 2 '
~
II)$ 1AUDCD!A! E~R pora ( RC P!IMD P> ~ INSTR r 'R PC29
- l. A: INS!R SACK E-IP.POC 4 c'. '
RACK E.IR-OS ~ l.la 107 rads
' BUBZONE A1 ZONE E tPCS STS!EM INSTR RACK
+. ! ~ PI I I~t) ~
I ~ ~
~ PSR-V-XS4/I AIR~118 CEP-V-3AD ERR POOI E MC al!A Cf P-V-4AD 6.0xl05 rads 1.7xl06 rads r ~
pI I
~ o CSP-V4+
I t D)pD DP E DIltR PSR-Y-X84/ I CAC-FCY 4A+ 15LC-it A+
ACK i I E~R DC"J xSLCA~
PSR V X84/2 CAC-Y-l+
~ i I i -i-* ZONE B RIR-Y-125A+ I5LC-X-OD
~
I ~
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/ 3) ~
Sl PPDESSICN E-IRCK&
I RHt-Y-1258 D ltSLC V IA+
~ 5.0xl05 rads RIR V 248+ XSLC-Y-lb+
RHt-Y-268+ !5LC V-IC+.
I CAC-FCY-3A+ RHt-Y-27b+ ttSLC-V-IOD r CA~
J
.! 'I MC CAC-FCY-4b+
I CAC-Y-13+
~ I r.: CAC-Y-b+ RHt-LS-IOA l5LC-FT-3A
~
~ ~ I CSP-Y 10+ RHt-LS-10$ l!SLC-FT-38 FAN IMSLC-F% a B CSP-Y 9+ R!R-LS-IOC I5LC-FT-3C
'-':-;i F~ I t SIC 3OA E-IRIS) RIR-LS-100 'SLC-FT-30 J ~
E-IR-P006t I
IN 5!R. RACK I ~ 'I E-IRPOIO)
E IR~
r
/ E-I R-F024+ ZONE F
~ AIR~ZSA 4 BUBZOXE J1
~ CEP Y 3A 2.2x10 rads ~ 1.3xlO rads ZONE D I CSP-AO-3 RIR-V llA+ CEP-Y-38+
Jl ~ 7.lxl04 rads RHt-Y-124A+ CEP-Y48+
RIR-Y-1 248+
E-8-1$ /R363) R!R Y 2lA+
E MC 52N)ISI RRA FC.I2
~ I ~ ..'.
~ '"
r II )tR E 'D PC?2 PgmP PACK EM 1$ /R364+
CSP-V-3D R!R-Y-26AD P E'MC 52!IAlAI
~t r~D txstR E IR PCC9 RACK -F- D CSP-Y-at CSP-Y-S+ R!R-LS-I IA
- ',,:;."...APERTURE CSP-V-7+ RHt-LS-11$
..5;D STS!EM RACK E-IR-P009+ RIRZS IIC E-IRAZD ft tCCK E IR po!r E-IR P017r A!R-LS-1 10 D E- IR-P022+
G 5tAIR I E-PP-7AE+
ZONE H f SUBZONE,IB
~ ttSLC-Fx- I+
S!AIR E-PP-SAED
~ HC 52/IA+
EI ~ 1.7xl0 rads I ~ 6.2xl03 rads
~ I~
I I
Ah ~h CAS-Y-l53
<<CSP-Y-93 ERIC-I 2+
ttSLC-FX-It kvaggMg 0~
I I
~ tCSP V.98 ZONE M o gipon PI-V-X265 I !MtlcATEs AREA IT<RE oosE I ROA-SPV 12 Located at clew.
AAIE Ovtsià Or DrACTCD I Sx-V-840 REACTOR BUILDING EI . ~())KI TIAS CALC4.ATE!I I 480'0 aoo Te 471'-O'ENERAL
+~t ~cAIITE or Actcat rg Ir SW-Y-842 rane I SI!-Y 844 NOTES: Sx Y 846 1.e/be+,oo, 0, ARE IDENTIFIED IN GENERAL NOTES 2,3,4,557 ON DRAWING M-422 SHEET 1.
NONE a
- 2. SEE DWG. M-471 SHEET 2 OF 2 FOR COMPONENTS RADIATION ZONE MAP OF LISTED COMPOSITES. REACTOR BUILD!NG EL.471-0 or)r<isg Do T WASHINGTON PUBLIC POWER SUPPLY SYSTEM '4- 471 1"lit
~
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l II I ~ L'dot Q l W~~ ~
1 I 4 ~
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1 2P-VI a @weber ZONE A CACNLV FCV2880 RHI-Y-1258+ NSLC-V-lb+
E-)R-POOI+ R%% 125$ NSLC~)d CACWLV-FCV3880 LP CS-F15-4 R% Y 248+ NSLC-Y-) C+
CACWLV FCY4880 LPCS-FT 3 A%~248 NSLC~) C E 38 TB/8364<
LPCS-P IS-1 R%-Y 268+ NSLC-V-) &
CAC-RLY-4A/CR)
LPCS-PS-9 CACWLY-4A/Qt R% ~268 NSLC~) 0 E-)R-P029+ R% Y 278+
CACALY-FCV)ABD RCIC-DP 1 5-138 CACPLY-FCV2ABO A%~278 RC ICE )5-78 CACRLY-fCV3ABO RCIC-PS-128 RC IC-PS-12D CAC4LT-FCV4ABD ZONE F 4 SUBZONE Jl CSP-V-3i R% V-)IA+
RCIC-PS-22b CEP-V-38+
CSP-POS-V/3 RHR~) IA RCIC-PS-22D CEP-POS Y/38 CSP-v-4+ R%-Y-124A+ Vdb+
R%-V-27A+ CEP R%~27A CSP-POS-Y/4 R%~) 24A CEP-POS-Y/4$ baal)
CSP-V-5+ R% V-12484 CSP-Y-&
CSP-POS-V/5 R%&-)24$ CSP-POS-Y/6 CSP-Y-li ZONE 8 R%-V-24A+
CAC-FCY-3A+
CSP-P05-V/7/PI A%~24A CSP POS V/7/P2 R%-Y-26A+ 4 SUBZONE J2 CAC-EBON/3A CSP-POS-V/7/P3 R%~26A E-IRAt+
CAC-FCVAB~
CSP POS-Y/7/P4 CEP-SP V-3A CAC-EHO-FCV/4$
CSP-POS-V/7/P9 CEP SPY-3$
CAC-Y-)3+ ZONE H E-IR-PO& SPWA CAC~) 3 ~ RRA-fC-12+ CSP HS LITS-44$ CSP-SPY<8
'AC-V4+ ARAN-fH/12 E-IR-P017+ RCIC SPY<5 CAC~B RCIC-OP)5 13A RCIC-SPY<
CSP-V-)Di RCICNPIS 7A ZONE I CSP-POS-Y/10/P) RCIC 5PV4l RCIC-F IS-2 RCIC Y-))b CSP~S-Vl)0/Pt CSP-POS-V/10/P3 RC I C-F)-3 RCIC~
RCIC-PS-12A RC IC-Y-113+
CSP~S-Y/10/P4 $ SUBZONE J3 RCIC~6 NSLC-FN-1+
CSP-POS-Y/)0/P9 RC IC<5-12C RCIC-V%8+
CSP-Y-9+ RC I C-PS-t0 NSLC~FR/I RCIC~B CSP-POS-V/9 RC I CPS-21 E-IR-65+ RCIC-PS-21A CSP-SPV-IDA RC IC<5-22C ZONE J ZONE M CSP-SPV-108 RCIC-PS-6 CEP-V-3A+ CAC-F CY-3b+
CSP-SPY-3 RC ICWIA CEP-POS-Y/3A CAC-E KCV/38 0 ~,)>>
s CSP-SPY-7A RC I C-Pl-5 CEP-Y~+ CAC-Y-)7 i CSP-SPV-7$ RCIC+T-7 CEP-POS-V/4A CAC~) 7 ECR-SPY-20 RCIC-PIN CSP-V 4+
FDR-SPY-4 E-IR-P022+ CSP-POS-V/8/P I E-)R-P006+ NS-DP 15-118 CSP~-Y/8/Pt RRC-PS-IBA NS-OP 15-8108 CSP-POS Y/8/P3 E-IR-POID+
NS-DP IS-1)C NS-CP 15-88 l5-CP IS-9d CSP-POS Y/8/P4 CSP~S-V/8/P9 PRO
$ 4+
NS-DP15-8)
NS-OP)SIC NS-DP I 5-9C OC RRC-PS-188 EPP-7AE+
E-PP4LE+
IISLC NSLC-H-A NSLC-TE IDA APERTURE E
NS-L IT 5-44A IR-P024+
NSLC-H-b+
N5LC-H4 CARD ZONE E IPCS-F)5-6 NSLC-TE-) Db CAC-FCV~ A+
HPCS-FT-5 NSLC-H-C+
CAC-EIa-fCV/4A IPCS-PS-12 NSLC-H-C CAC-VA+
NSLC-TE-)OC CAC~4 NSLC-H4+
AvaBable On ZONE D R%-V-))bt E-X-TB/8363+ R%~) )8 NSLC H4 ~]perhu'e Card CAC<LY-4$/CR) N)LC-TE-)00 R%-V- I 25A+
CACALY-48/Qt NSLC Y )A+
RHRW )t5A CA"PLY-FCVI880 NSLC~)A COMPONENT EQUIPMENT LIST FOR scut NONE COMPOSITE'OUIPMENT SHOWN ON RAOIATION ZONE MAP REACTOR BUILOING EL. 47'I'-0"
~
WASHINGTON PUBLIC POWER
'UPPLY SYSTEM
~tl lX
~ ~ (~~ Cg 4
'0 0
/
II E~
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I . 5
- r. I 533. SAFETY RELATED EQUIPMENT IOCI - III BY ZONES
'ON OIWSna I ~
ZONE B ZONE M ] TIP-SV-3 f -IR-P030+ ~ RIIR-IO-538 TIP-SVu
~ a.lxl05 rads 1.0K ID J
rads e TIP-SV-5 RIKI INSTR RACK AKNf CATI MATISTEAM FLOw POIB~ t-:R.PO33 LISIR RIV ~ TIP-Y-l E 'R I f E IR PO'3(~ E 'P POTS~ E- IR4018t RIR-Y 168+ TIP Y 2 E-IR-POZst RIIR V 11St TIP V 3 E IR4030t RIVI Y 530t TIP-Y-a I
I STAr,.'.
A ELEV STA R E-IR-P032+ TIP-Y 5 ZONE DRIVE MECNAIASM X NO IC 0'IS-PO5-V/Z80/2
<TTPI ZONE F ZONE Q Q,QI- RRC~I 6A
~ 6.25106 rads TIP-Svk I a.6510a rads ~ 5.1xlOa rads
~
l5 V 19+
ITS-V-28lt E-IR-6at TIP-SV-6 l5-V-288t E-IR46t I5-Y-28Ct TIP I Rill p DETAIL A F' RRC-Y-16A+
15-Y-280t I
ZONE S {DETAILA C-3)
, NOTE- SEE TOKE S
'EIIIWRKIHT I
- .,"'y /
E~
IISTR RACK ac 9 ZONE I RS-V<1At IIS-V<18t I
Located at 510'6 eler.
re ff aoo
~ K RS-V %1Ct Lr'ERST
~ ESP~I zfxies 'P'nd 'C s AX
~ 1.5x106 rads I5-Y<10t I
I~ f
- RISTR RACK I5LC-V-IDt I ~ RCIC~4 FAN CCKL IRKT / E~ err 15LC-Y-ZN I
FIPAFC AM, I AL jf j 6 I CSP-Y-It CSP Y 2+ ITSLC-V-Z It-
~ 2.6xl0 rads Ir I'.~
I
~ UNNE L ffj RIVI-Y-Dt ITSLC Y-Z"+
E.MC TC RCC-V-10at I
rf 15LC Y-20t OT t I RCC-Y 21+
I~
~
I I ~
IISLC-V-3lt iI I RCC-Y st I5LC V 31L FAN COL IIIRT IRRIVFC o I'r I
'lf,
~
I I
r CSP V
~ TCSP-V-91 96 I5LC V-3r+
I5LC V-30t I
RCIC-Y4+
RIEI-Y-53At RN-Wt
~j:
~ 'Y""'-
jtr fI I Af't
~
ITSLC-Vdt I
I
~
ri I
~ ~
"j" I ZONE K I5LC Y-5t I
Wt E-IR-63+ I5LC V 9+
MSLC-FII 2~~ r H fr ~ 1.9510a rads RAI-Y<flt
/ /
~
'I I RFM V418t ITIN ST~MM FLOW L E-IR-63+
PISTIL RACK PARTIAL. PLAN EL 510'-6" E-IR-P015+
E IR PCIS LOCA EO ABOvE E IR P021+ LD-TE-29A PRE~ CAB KI TONES PANO C ~
E IR FOSI E-IR-P031+ LO-TE-298
' ITSLC-Pll-Zt LO-TE-2$ C I
I RRC-Y-168+ LD-TE-ZSD I
LO-TE-31A PRE AhIP CAB EMR PO33 M LD-TE-3 18 I
I I II PRC
~
RRC-Y-ZO LO-TE-31C I
LO-Tf-31D K;
i SUBZONE K1
~ PSR-Y-X11A/2 I54E-3II l54f 30I APERTURE C 4C IA CIA Crte f STAIR
~ l. Ix106 rads I54E 3C I
STA R J~
II SKID I I I I~
ELtv II IJ PSR-V-X11A/2 I54E 30 CARD 3 . .. .. .... 5- PSR Y XZIA/6 RNR tISTR RACK INSTR RICK E<R~? I ERR e3 ZONE P I INDICATES 'ARE3 WNEREIOOSE- $ SUBZONE K2 TIP Y 1 r
z RA'IE ODTSIOE'OF REACTOR: er BIRLONO WAST TIALCIRATEO.
NO'I RKR/ATIVEOF ACTOAL EONEI
~ f-IR-P033+
II 9.1x103 rads
~ l.lxl06 kvBQgge '0~
I TIP-SY p~b REACTOR BUILDING EL. 501'-0'ENERAL E IR4033t 1
NOTES TIP-SY-Z
,1.e~~,oo. ii ARE IDENTIFIED IN GENERAL NOTES 2,3,4,557 ON DRAWING M-422 SHEET t. ec'ADIATION scuI NONE
- 2. SEE DWG. M-501 SHEET 2 OF 2 FOR COMPONENTS ZONE MAP REACTOR BUILDING EL. 501' OF USTED COMPOSITES. OKAKIKO NO I~
WASHINGTON PUBLIC POWER SUPPLY SYSTEM M-501 6 TII! I w 2 I
aSOV 700.>q 'jj'g"" '
I fQ 'P Q>> pc Cf
~v 8 9' m
m~ ~O II N
t,
>l y*
g 'g i
ZONE 8 Ns-DP IS-BA E-IR 4018+
Ns-DPIS 9A NS-CORN-Y 280/ J2 RHI-DP Is-12A E-IR-P021+ Ns-ccaa Y280/J3 RHI-FIS IDA RHR DPIS 128 Ns-POS-V/280/I R%-FT-15A RIQ FIS 105 l5-POS-Y/280/2 R%-Ps '16A R%-F IS-IK as-POS-V/280/3 RIQ Ps 19A R%-F T-158 Ns SPV-28D2 E-IR-P025+ R%-FT-15C l5-SPY-2803 as~ Is-1 10 R% Ps 165 NS-V<7A+
~PIS 8100 R%-Ps-16C as~<7A as&IS 80 NS TATS+
RIQ Ps 198 NSAP IS-9D R%-Ps-19C asm<78 E-IR-PO& NS-Y<7C+
E-IR-P031+
SRH-ENt-)A SRH-E/Hp-Ib as~67C E-IR-P032+ NS-V670+
HSLC-FH-2+
SRN-E//p-I c HSLCW-FH/2 as~70 N5LC V 10+
RRC-V-168'RC-HO 168 HSLC~IO ZONE F iSLIBZONE K2 NSLC V 2A+
E-IR-64+ E-IR-P033+ NSLC~2A Csp-DPT-5 sRN-E/ap-10 NSLC V 28+
CSP-DPT-6 NSLC ~28 CSP-RLT-ARCSPV5 ZONE M NSLC-V-2C+
RHR-V-168+
CSPNLT-ARCSPV9 l5LC~2C R%%-168
<<CSP-RLV CRS NSLC-Y-20+
R%-V-lib+
<<CSP4L T-CR6 R%%-1 78 NSLC ~20 CSP-SPV-4 HSLC-Y-3A+
R% V-538+
CSP-SPV-5 l5LC~3A IV%~538 NSLC-V-38+
CSP-SPV-9 RFV-SPV-32AI ZONE 0 NSLC~38 RFN-SPV-32A2 Ns Y 19+ NSLC-Y-3C+
RPl-SPV-3281 as&-19 aSLCA-3C RFR-SPV-3282 Ns-Y-28l+ NSLC-Y-30+
E-IR-6& Ns-CONN-V28A/J I NSLC~30 05-PT-3 NS CORN-Y284/J2 NSLc-TA+
CSP-SPV-I as-CQw-V28A/J3 NSLC~4 RRC-V-Idly NSLC-V-s+
Ns-POS-Y/28l/I RRC~16A NS-POS-V/28A/2 HSLC~S Ns POS V/28A/3 NSLC V 9+
ZONE I NSLC~9 Ns-SPV 28A2 CSP V I+ Y45A+
l5-SPT RFR 28A3 CSP-POS-Y/I Ns-Y 288+ RFH~65A CSP Y 2+
RA-Y-658+
NS-CORR-V288/JI CSP-POS Y/2 RFH~658 Ns-CON II-T288/J2 R%-Y-84 Ns-CONN-Y288/J3 RHR~8 ZONE K E-IR-63+
Ns-POS-Y/288/I Ns POS-Y/288/2.
ZONE RCC-V-104+
~< pRC CEP-SPV-4A CEP 5P'V 48 NS-POS-Y/288/3 Ns-SP V-2882 NS-SPV-2883 RCCA-l04 RCC-V-21+
APERTURE CHS-PT-4 Ns-Y-28C+
RCC~21 RCC-Y-s+
C'AR D NS-COHR-Y28C/Jl CSP-OPT'CSP-RLY-ARCSPV6 RCC~5 Ns CONN Y28C/J2
~ ~CSPWLT ER4 RCIC W+
Ns-CORN-V28C/J3 CSP-SPY-2 RCIC~8 Ns-POS V/28C/I RHI-Y 53A+
CSP SPV 6 NS-POS-T/28C/2 R%~53A RCIC-SPV-26 RCIC-SPY-5 l5-POS-Y/28C/3 RICI-TAO+ a>~>o g>
E-IR-PDI 5+
NS SPV 28C2 NS-SPY-28C3 RVCUMA0 ~Apertuee cga Hs-DP I 5-1 IA Ns-Y-280+
NS-DP I 5-810A Ns CONN T280/Jl COMPONENT EOUIPMENT LIST FOR COMPOSITE EQUIPMENT SHOWN ON RAOIATION ZONE MAP REACTOR BUILDING EL. 501'-0" WASHINGTON PUBLIC POWER SUPPLY SYSTEM 48 pvpvp 0'yg :TI"
f I+ l ca.
h
.~
A II t~
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1 g
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ZZ5 - Ih<
SAFETY ~TED EOUN3MENT
-BY ZONES II1 I ~ ~ ~ <
N3<ITN ; I!NB.) ZONE B ZONE H ZONE 0 LPCSN(3 5 E-IR-P039+ RIF(~20 fri 4' ~ 6.ax 105 rads ~ 8.3xl05 rads ~ 1.7x 10 rats CRDWQI(93 Total) CIA-V-3054 RIN YAZAc LPCS-V-54 E-IR49c RTSI Y42C4 8
SINK ' E-IR4027+
Elf; I I
.:C 4
. FANC(N sfap ZONE C Hpcs-v-ac OBK (RRA4C < E-IR-P002, LD-TE-3OA Sff-Y-755+
~ 5.8x105 r d LO-TE-305 INSTR RACK
>>-Y-X250 LO.TE-30C E IR 7O
-:- r I CRD-IR-34 Pl-y 7251 LO-TE 300 CRD tc<DD FNAC 88A "EVA(LRACK. Q: E-IR-F0024 Pl 'V a253 n E IRN<(ND E-la~ PI-Y-Xzsz ZONE p 0'WCQ
~ <STR RACK F IR 7 r r Pl-y x263 CRD-Y-IOt r h Pi-V-Xzba ~ 5 3xlOl rads FACTOR yg~cL XV.P.IA ltvEL AND PRESP Hy-Y-17A PI-Y-R266 IN(TR RACK PI¹(P RWCII KISTR. RAC<<
C IR PCOO P,-'t F-KI P(307 NY-V-ISA Pi-V-X267 CIA-V-30Ec HY-V 19A E-IR-73+
.,~ r, .
Hy-V-20A 4 St%ZONE Hl K IRWOzst
~ ~ r c '
0 ~ <
". I c
g 4
,~
rr REA< TOR VESSEL LFVELAPRESS NT-V 33A Hy-V 3lA HY-V 35A
~
PSR-y-IBO/2 9-3xl05 rads CRD-SPY-182
.'dy tNS'1 R RACK HY V 36A PSR-Y-QO/2 CRD-SPY-9 I I (tat ? E IR.POOS NSLC-PT-10A
' I :
I .=":G II ILSLCW-108 ZONE J I
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INSTR.RA(3(L C IR PC<87 ""i <
L I
~
ZONE D RRAAhF H/10 2.3x10 radS
~
CRD-KU I92 1.0xl05 rads Totall NSLCPT-IOC
.NSLC PT 100 NSLC-PT-12A I I NSLC-PT-128
< Ltli CRO-Iel(92 Total )
INSTR RA"K I LEVEL dPRESS
'<STR RACK
'r<: I RRA-FC 1OI NSLC-P T-12C g-IR 71 NSLC-PT-120 HT-V-175
~ c ~ PI-Y-2256
/'EVA(L RACK F IINIKXIO "
HV Y-lbb Jg HY-Y-19B P I-Y-2257 ROA-SPV-10 HY-Y-208 pl v-a258 I 2, PI-Y-1268
<o RPT Sw(iR Ny V 335 P<STR RACK EvR 74 (C C8 KPTAA(SN IIR
~
Tr ZONE F
~ CSTR RACK a IR poo4
~<h <
I
.~ r ~ ~ < <",
r
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LD-TE-2A B.la 10 Pads Hy-V-3lB HY-Y-355 HY-Y-365 I
SUBZONE Pl
~ PSR Y I73/2 RP1 SW(KE RVCU-Y-a4 i (Ms Rpt48(sN taa 1.5x'105 rads KLN CDL (Aft ZONE K
'"'RC OtRA FC II) 'T<r<<
ast(FRACK F I4C 78 NR (TS< ~ K IR-7l+
Yr PSR V I73/2 8 I <to
~ <<.
LD-TE IA ~ 2.0xl0 radS C I PISTR RACK 77 I¹4 c<R<4 CDNTRCL<r RE(AY RACK <A. <
LD-TE-lb C IR PDOS LO-TE-IC CIA-V-20+
SAIC 784 "~
I STA<R CRD I40D(XSSM I
3 hstA<R LD-TE-ID LD-TE-2A E-IR-71+
E-IR-Tl+ APERTURE
~
~4
}M ELEV II M 2b LD-TE-2C E-IR-POOl>
Sff-Y-75A4 a,., CARD
~
4 h
I Ih M-20 LO-TK-3A RAG<ATION ='RADIATION14CNITDR AC IP¹1 (RRA AC.(g i IIOt¹TQ<<
Sw.c R-47 I
i SW.SRW3 M-35 NSLC-PT-11 LO-TE-3C ILSLC-PT-13 I
ROl-SPY-1'I REACTOR BUILDING EL. 522'-0 LO-TE-30 AVaBELMe .Qa ZONE G ZONE N
~kpeeue C ra GENERAL NOYES: ~ Rfbf~zb BC-7b+
T.~~,oo. t. ARE IDENTIFIED IN GENERAL NOTES ~ 1.3xl06 rads ~ 6.7x 103 rads 2,3,4,5IL7 ON DRAWING M-422 SHEET 1.
RIN-Vdzb+ RRAhFC-114
- 2. SEE DWG. M-522 SHEET 2 OF 2 FOR COMPONENTS NONE RADIATION ZONE MAP OF USTED COMPOSITES. REACTOR BUILEVNG EL.522'-0" WASHINGTON PUBLIC POWEP. (Nkwho <Kh r A(V SUPPI.Y SYSTEM M 522 8 ~ (5
~
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{tr cat ZONE B F, RIB{-V>>lddc A ~ CH5 SRL)4>> ~
RC IC~64 RHR V-49>> E IR 68<<
~ 3.lr106 rads ~ I.or loia rads RIB{ V48Sc ~ 8 Sx)0 l'ads 4rcCtr<<>>.3A !c 4rtccw t
't( istc {+RAT ! 4{GEKE4 <<0{S-5R )ac
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PS A RHR V 16A+ RCIC-LS 5 E-JS-IB/LR68/2<<
c 4t
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<<as-H-SR14/T5 RHR-V-7 58 ZONE Q
<<05-H-SP)4/I6 ~ CAC AO-)5 RSO(: wATE4 D LO-TE-IE <<QIS-H SR)4 I
fll ~ 1.04106 rads PCPttPS l TTPt <<05-H-SR)4/Td l TA.eaC LO-TE-IF
~
~
{7.-.
~ 44
- .. ~
LD-TE-2E LO-TE-ZF 44SUBZO E FI ZONE K CAC FCV 18+
CAC-V-15 I: ~ CCN~ ~ CAC-EHO-FCV/2A STAttG ST LD-TE-3E ~a S,or)0 I'44s INSTR PA{tE' LO-TE-3F I ~ '1.Or)OS raos
~ I. OSUBZONE 01 Ctrs SR 33 t S>>SR {3 RCIC-LS-6 ~-FC-17<<
REACTOR e{rtG CLOSEO COOct/Ot WATER ttEAT
,'R ",
I t tt-.
{Cacti
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~
CEP4tOS-Y/) 8 1.2xl06 l a4S EXCHANGER r S SR 3{ {AheCE GAS Alla XE!Pr ZONE C CH5-RPS-HIPBO REA-AD-8+
'Pr '"A"t'CA S ER" """" ": ROA-AO-15 <<as-TQsc f
CEP-V-IA<<
- t7 r 1 CEP-V-18+
/'.'r
~~ ~ 1.) x)0 radS ~ <<CHS-IS-SO ROAN AD/19 CEP-V-ZA<<
~ ~ C IBP.; <<as- T~ ~
CEP Y 28+
ROA SPV-15 <<aS Tc dB rtt ~ 't 4
~
ROA-SPY- I 7 -05.TI~ ZONE L
>>r INS, R SLC Y 4A OIS-'15-80 t 348
'IBLM)
'll E tR ta<< SLC-TAB <<<<05- TS-SA l ).Sx)05 radS Q:.'" ~
t I) OIS-IS-9 8
<<<<CHS-TS-9C
~
%%ZONE E RCC-V-129+
' OIS-TS 90 05-SR-)3+ RC C-Y-)30<<
~ 1.0r106 rads <<SH-PS,)0)5 RCC-Y-131+
<< <<ST{-V c06 RIBt Y )348<<
ch <<05-SR-)3<< <<Sc{-V-209 I RRAPC 19+
<<Q{-Y-210 tF Cll kRA-FC 20+
<<ST{-Y-Z)1 Q{-V-187A<<
,h r r
<<Q{5-H 5R13fl5 Ql-I-Idid+
I C' D t <<05-H-SR) 3/'16 ZONE G Q{-Y )BSR<<
t E-LR-67<<
t
<<QIS-H-SR)3/T7 t S>V-)888<<
>> ~ 1.2x104 rads
<<Q5 H SR)3/TB G- r!
~ !77 ~ E-IR47<<
ZONE M 4 4 SUBZONE E1
~ a 5.04104 radS
~ SIN~) 34A ZONE.4 ~ 4,lx)0 rads INSTR. RACK CAC-00-FCY/28 E-;R-Ge <<<<RRA-FC-15 L4 C
~ 1.64106 rads RIB{-V-)34A<<
I
<<CHS4{ HS-HLP I
<<as-Ts-4A l
l CACPCY-28+
CAC-Y-l)+I ZONE N ICONS-4 PRC O NQQ
+ I E:E I' lI STA,c 0!S-Ts-ad
~ <<as-T5-4 C RCIC Lp
{
kHR-V-23+
13+ ~
~
RIBL 2.)x10
((0 3A rads APERTURE
~
rlr.
C r ~
<<OIS-TS-40
<<<<05 TS SA ZONE J RIB{-I 3A<<
RIBL WA<<
CARD RNR rE ' C 4E C>>hC; l c PGC 4 <<OIS-TS-SP ~ RIB)~ 9 ttcttatCGEF RQR vw t c cpC-N T ECC tA AGE raale REM C RC. Pc p c*a'e
/<<S E{tC4ANGE<<
RPR Py 8
<<SH-PS-LO) l ~
t 3.)x)06 rads RIBL-WQ<<
AVaI'Liable Oa
<<Ql Y ZOI
<<Sl{-V 20l RIB)-V-1)5<< Spectre.Cgrg REACTOR BUILDING EL. 548'-0 <<SH-V-212 RIN-Y 116+
RC I RIB{ YAOA
<<<<5V V 213 RIBt Y-38+
GENERAL NOTES: RHR-YW RHR Y 75A 1.@pep,oe, t, ARE IDENTIFIED IN GENERAL NOTES 2,3,4,557 ON DRAWING M-422 SHEET 1.
RAOIATION ZONE MAP NONE
- 2. SEE DWG. M-548 SHEET 2 OF 2 FOR COMPONENTS REACTOR BUILEHNG EL. 548-0 OF LISTED COMPOSITES.
WASHINGTON PUBUC POWER Tt<<C<<tha h>>
- 3. ttTHESE ARE ESTIMATED DOSES. DOSE WILL BE SUPPLY SYSTEM 548 M g 6 VERIFIED UPON EQUIPMENT INSTALLATION.
83 OV V'0 0 13-'l't ' t<< {I
r f
fl pt&
t hl N gg,1+ F% JJ II f ' j t
ZONE 8 RN-V-3$+ ZONE P RCIC-W4+ RHR~38 E IR-68+
RCIC~ RN CEP SPV-2A TAO'HR~A RN-V-)BAi 0 CEP SPY 28 RNA-)6A RN V-488+ CIA-PROB-)8 RN-V-)7A+ RHR ~88 CIA-PS-21$
RNA 'IFA RN V49+ CIA-P5>>228 RN~ 9 CIA-Pl'-21b RN Vdbb+ <<CIANLY 2)b ZONE E CHS-SR-13+ RHR~ <<CIA-105-1 8
<<CHS-AY I CHS-PT-2 0IS-AY 3 05+7<
CHS-PT-8 ZONE K RELRLY CR2 4 SUBZONE E1 CAC-FCV-2A+
REA-SPY-V/2 RRA-FC-1 5+ CAC-EHD-FCY/2A EWS-)b/IR68/2+
<<RRA~FH/15 CAC-Y-2+
<<CIA4LY-228 CAC~2 ZONE F REA-AD-8+
0IS-SR-)4> REAM-AD/8 ZONE 0
<<0IS-AY 2 CAC-fCY 18+
REA-POS-AD/8 t
~ 015 AY 4 CAC-EH)-F CV/lb CAC-V-15+
CAC~)5 0 SUBZONE F)
RRA-FC-1 7+ ZONE L
<<ARAN-f8/) 7 RCC-V-129+
WSUBZONE 01 RCCA-) 29 CEP-Y-)At ZONE G RCC-V-) 3>
CEP-P05 Y/IA E-IR-67+ RCC~)30 CEP-V-IB+
CEP-SPY-IA RCC-V-)31+
CEP-POS-Y/18 CEP-SPY-18 RCC~)31 CEP-V-2A+
RN-V-)348+
CIA-FROG-)A
/ RN~)348 CEP-POS-Y/2A C ILPS-2)A RRA-FC-19i CEP-V-28'EP-PDS CIA-PS-22A V/28 CIA-PT-2)A RRAW fH/19 CIAO-2)A RRA-FC-20 CIAALY 22A RRLH fll/20
~ ~CIA-TDS-)A SR Y 187Ai OIS-PT-) SH &4'-18 7A CHS-PT-5 Sv-V-)878+
CHS-PT-7 Sv~-)87$
SH-Y-188A+
ROAST CRIA ROA SPY 100 SM~)bg Sv-V-) 888+
ZONE H 5WO-)888 CAC-FCY-28'AC-EIO-FCY/2$
RCIC-V-13+
RN-Y-)34A+
RHRA-) 34A APERTuRE RN-V RCICM-) 3 2> ZONE N CARD RN~23 RHR- Y-3A+
RN~IA Aho AvaHable pg RN V484+
ZONE J APCgfgge CEIL/
RN-V-))5+ RN~SA RN-V-68A+
RHR~H4 RN-V 116+ RN~
RNA-93 COMPONENT EQUIPMENT LIST FOR sew( ~ NONE
.OMPOSITE EQUIPMENT SHOWN ON RADIATION ZONE MAP REACTOR BUILDING EL. 548'-0" WASHINGTON PUBLIC POWER SUPPLY SYSTEM M-54 CIA 3 K I' EI t3 8 0 V 0 T 0 0 $ Q '-tg
aa 4
- y j
6 1 ~
>~I h ~ A l l
1
ITIVS SAFETY. RELATED EQUIPMENT Ztst -Vt BY ZONES Ov ONIAVPO ZONE B ZONE H 56T-OV-IA)+ SCT-F U-IA+
~ CACTI-6 SCMt~/)$ 2+ SGT-OY- IA2v SCT-FU-lb+
~ 7.65105 rads
~ 9.lxl03 raos SCT-DY-)A)v 56)4V-)$ 1+
RRA-FC-)4A DY-)82v CAC FCY IA+ SGT CAC-YW SCTWY-lb 3+
sl' ~ ir E A. I N-2 ZONE I it A .'l PEA FN 28 FILTER UN'lS ZONE C a OR~ 7$
~ kOA-fN-)A ~ ). Ix)06 rads 0 SUBZONE N2
~ 3.0x)0 rads I
fPC-L)S-)A
~ I OR-YRlb+ 9.74105 rads Sot OV ~ AS I ~
I REA-DPT-'IA2 OR Y 528+
REA-DPT IA3 OR Y 738+ SCTPN IA)+
REA SR-22 kEA-OPT-1$ 2 OR-V-748+
I 4 SCT-FN-IA2P PEA FNQAITOP)
Rf AMN18tatsTTI
~
~rSs RAREA4E 9A OR V478+
SC TON-Ib I+
RE ILIRN A IP, PARTAKE-98 PLE N tpr SCT-FN-)$ 2A
~ %EAAE-9C ISUBZONE NS
~ AREA4E-90 kEA-OPT-I $4 SGT-EIO-IS)
~ 3.0x)05 radS ZONE D ZONE K
~ FPC-TE-7 56TTFN-)AI+
STAtCOV GAS ARAN FN/13 FILTER Otts'I pSGT CESHT
~ 1.0x)09 rads 56T-FN-IA2+
~
PILSf.,' ~ S.lx)04 rods I I 56T-FN-181+
I
~-TE-7
~
i 56T-FN-182+
TIAsrtfp A.P E IVC-%r RRA FC )3t OPE PATTIVS I HRECa 8:. D PAIIEL I OFPSOTst t
.I tli t
I E-CP CACIIAI SGT CONTI I~ I, ~,
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CAC-FT-78 R% V<7At SGT Fs-2A2 CAC-LT-18 R%~<7A 56T-FU 18+
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CAC-FCV-3A+ RHI V 278+ ITSEC-V-IDV I ~
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t 2-IRONS+
CAC-Y-13+
CPC-W. Rta-XS-I OA ITSLC-FT-3A
/ stt RHI-LS-108 ITSLC FT-38 CSP-Y-I 0+
PAN 04SLC PN S B CSP-Y-9+ RHI-LS-IOC ITSLC-FT-3C r.5(C.SOA~ I Rta LS 100 RSLC-FT-30 E-IR-P006+
2-IRPOIOv INSTR RACK~
E IR~ ~ r E-I WOZaa I ZONE F 4 SUSZONE J1 V' ~ - ~ Rla IO 26A ~
~ ~ -:C cRI CP-Y-3A
\ t t r ~ 2.2x10 I ads ~ 1.3xlO rads ZONE D
~ r ~ CSP PD-3
'i'-
Rta-Y-I IAP CP-V-38+
~ l.1xl04 racs RIa-V-I24A+ CP-Y4bv RHI-V-1248+ CSP V 6+
t .-'-. E-JB-TB/R363s E-KC $ 2(INCI RRA PC I2
~ 1IEclxc PVKIP
't(STR PACK L
E Jb TB/R364+
CSP-V-3+
Rta-Y-24AV aa-V-26AP I SUBZONE J2 E KICr/2IIATAI P~&r " - ic- I E 'P PC22 D
CSP-Y<+
5v ~
E IHZP I.lxID5 Pk INSTR RACK E I R POOI2 /IV ~sr % p CSP Y CSP-Y 7+
RIa-LS-I IA J I I
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Rta<S-I I b x.cp varsar I AIR RHI-LS-I IC LCCK NSIR RACK E IR.POT T E IR POI7+
E- IR-P022+
RIa-LS-1 ID ',APERTURE
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ZONE H ""-*'-"-III6sa, CAR D I ~ EM<2/IAa E'EV sl i ' ~ I.lxlO rads
~ 6.24103 rads
"'4) r 453 lsss.ss I IIMSO AVSVtgbiO
~
CAS V ARABIC-I2+
<<CSP-V-93 card 0'pattufe vvCSP-Y-98 I ZONE M PI v-x265 I VIKNCATESIAREA wvERC NOSE ROA-SPV 12 Located at elev.
RATE'OVTKIOE IO<<"PPAC TORA 5 Sti-Y-840 REACTOR BUILDING EL. 471'-0 OIPLCSS(SsVIAS CA%4IIZATXTI r 480'0 aacve
.got~cattvE op4T-TOAL io'i(L I SII-V-842 Xcee 'I Stl-V-844 GENERAL NOTES: SII-Y 846 1.4/alba,oo. 4, ARE IDENTIFIED IN GENERAL NOTES L 2,3,4,587 ON DRAWING M-422 SHEET 1.
NONE
- 2. SEE DWG. M-471 SHEET 2 OF 2 FOR COMPONENTS RADIATION ZONE MAP REACTOR BUILDING EL.471-0 OF LISTED COMPOSITES. Lvkvsvo IO AIV WASHINGTON PUBLIC POWER cSUPPLY SYSTEM M 471 e T.I(t 1 (P st 0'1 8"-
~
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CACWLV-FCV3880 RHRA 1258 NSLC ~lb LPCS-F 15-4 R%-V-248+ NSLC-V-IC+
CACRLY-FCV4880 LPCS-FT-3 E-38-TB/8 364+
R%~248 NSLC~IC LPCSWIS-1 C ACYL V-4A/CRI R%-Y 268+ NSLC-V I&
LPCS-PS-9 CACWLV 4A/012 A%~268 NSLC~ID E IR-P029+ R%-V-278+
CAC WLV-fCVI ABO RCIC-DP I5-138 CAC4LV-FCY2ABO 8%~278 RCIC~IS 78 CAC<LV-FCV3ABO RC IC-PS-128 RCIC-PS-1 20 CACNLV FCV4ABO ZONE F 0 SUBZONE Jl CSP-V-3+ R%-V-IIA+
RCIC-PS 228 CEP-V-38+
R CITS-220 CSP POS V/3 8%~1 IA CEPS/38 85 m.
CSP-V-4+ R%-V-I24A+ CEP-Yab+
R%-Y-27A+
CSP POS V/4 8%~124A CEP~S-Y/48 8%~27A CSP-V-5~ R%-Y-1248+ CSP-Y+
CSP-POS-Y/5 R%% 12lb CSPAS-Y/6 CSP V-7+ R% V 24A+
ZONE S CAC-FCY-3A+
CSP-POS-V/7/PI R%~24A CSP-POS-V/7/P2 R%-Y-26A+ 4 SUBZONE J2 CACAO-FCV/3A CSP-POS-V/7/P3 8%~26A E-IRQR+
CACWCVAB+
CSP-POS-Y/7/P4 CEP-SP V-3A CAC EN-FCY/48 CSP-POS-V/7/P9 CEP-SPV-38 CAC-Y-1 3+ ZONE H E-IR-P009+ CSP-SPHtk CAC~I3 ~ RRA-fC-1 2+
HS-LITS-448 BRAW-FN/12 CSP~V<b CAC-Y4+
E-IR-P017+ 25-CAC~B RCIC-5PY 25 RCIC DPIS-13A RCIC-SP VA CSP-Y-IO+
RCICQP IS-7A ZONE t CSP-POS-V/10/Pl RCIC SPV-54 $6 RCIC-F 15-2 RCIC-Y-110+
CSP~S-Y/10/P2 CSP-POS-V/10/P3 ICOSA RCIC-F'I-3 RCIC ~<0 RC IC-PS-12A RCIC-Y 113+
CSP ~5-V/IO/P4 QSUBZONE J3 RCIC~6 NSLC-Fa-I+
CSP-PO5-V/10/P9 RC I CPS-12C RCIC-V4B+
CSP-Y-9+ RCIC-PS-20 NSLC~FN/I RCIC~B CSP-POS-Y/9 RC ICPS-21 E-IR-69 RCI C-PS-22A CSP-SPV- IOA RC I CPS-22C ZONE J ZONE
'C'LV3:
CSP-SPY-1 08 RCIC-PS-6 CEP-V-3A+ M'AC-FCY-38+
CSP-SP V-3 RC CEP-POS-V/3A CAC EWCY/38 CSP-SPY-7A RC I C-PT-S CEP-VW+ CAC-V-17+
CSP-SPY-78 RCICWT-7 CEP-POS-V/4A CAC~17 ECR-SPV-20 RCIC-PT< CSP-VQ+
FDR-SPV-4 E-IR-P022+ CSP-POS-V/8/P I, E-IR-P006+ NS-DP15-118 ESP~-Y/8/P2 RRC-PS-IBA I5-OP I'll 08 CSP-POS-V/8/P3 E-IR-POI 0+
NS-CP IS-I IC NS-DP IS-810C l&CP IS-88 NS-DP 15-98 RRC-PS-188 NSLC CSP~-Y/8/P4 CSP-POS-Y/8/P9 H4+
~"""
P R C NS-OP I SAC E-PP 7AE+ NSLC-i@A NS-DP 15-9C NS LIT~A W'PME+ NSLC-TE 'IOA NSLC-H-B+
'APERTURE E-IR-P024+ NSLC H4 HPCS-FIS-6 HPCS-FT-5 ZONE E CAC-fCY-lA+
NSLC-TE-108 NSLC-iK+
CARD CAC-EIN FCY/4A HPCS-PS-12 NSLC H-C CAC-VA+
CAC~4 NSLC-TE-IOC o A~BEIBMC On ZONE 0 E-JB-TB/R 363+
R%-Y Iles RHR~Ilb NSLC-H4+
NSLC-H-D
'Aper.tume ~g CACWLV-48/CRI NSLC-TE-100 R%-Y-125A+
CAC ALT-48/Gt2 NSLC Y IA>
RHRA-125A CACWLV-FCV1890 NSLC~IA COMPONENT EOUIPMENT LIST FOR ~c NONE .
COMPOSITE EOUIPMENT SHOWN ON RAOIATION ZONE MAP REACTOR BUILOING EL. 471'<<0" WASHING1'ON PUBLIC POWER SUPPLY SYSTEM M-471 3
~~
I 0 ~ 1, ~ C 4
1 '
I33fr SAFETY RELATED EQUIPMENT LOIN- IAI BY ZONES TIN OMAYYO
. ZONE B ZONE MI TIP SV-3
~ 6-IR-P030I RHRW538 TIP-SYAa
~ 4.7xl05 rads I.OK106 rads TIP-SY-S
'lfe INSTR RA(K PRE AIAICAS K-IR PC33 MxtlSTEAM FLOII TIP Y-I E IP POISE PISIR RA/K
~ F: IR.PC30 I
K '" POTS~ f- IR&018/ 'iieet.V-I6S+ TIP-V-2
~ ~
f IR-P025I RIIR-V 178+ TIP-Y-3
~ f 6-IR4030 Rtat Y 538< TIP Y-l
, ~
SI/.Ri 4 ELEV l STAiR I E-IR-F032+ TIP-Y-S B ZONE 0 DRIVE MECNAMSM X NC IC "D) 3 E NS-PO5-Y/280/2 ITYPI ZONE F ~ l.2x 106 pads ZONE 0
~ RRC~16A TIP-SV4
~ = l.6xlOl rads ~ 5.7KIOl rads
.'C NS-Y-19+,
NS-Y 28A<
/ E E-IR-66/ NS-V-288l'S-Y-28C+
p DETAIL A RRC-Y-16Al NS Y 280I NCTF'EE TONE S OKIIOW RKINT
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I RCC-T-5+
CSP-V-96 NSLC-Y-38+
x RCIC-YW FAN COL NSLC-V-X+
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- 5 NSLC-V-30+
f RTCI TW I'I;I T
NILE-VA+
~
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I MSLC FN 2~~~ ~ 7.9xlOl rads RFV-VISA+
/ j; RFTI- Y458+
MAIN STEAM FLOW
~ i
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~ ISTIL RACK E IR POIS
~ X I PARTTAI: PI.AN EL 510'-6" E-IR-PDI 5+
LOCATKO A8OVE- E-IR402l+ LD-TE-25A PRE MAIII CAB E IR PO31 K L E-IR-P031l LD-TE-298 NSLC-FN-2/ LD-Tf-29C I
RRC-Y-Ibbl LO-TE-290 LD-Tf-3 I A PRE AMP CAB E IR-PO33 M I I LO-TE-3 Ib I Cdh I""'" P RRC-V-20 LO-TE 31C LO-Tf 310 Ag?ST'TAIR 0 SUBZONE K1 NS-RE-3A NS4E-38 R C f PSR-V-I77A/2 CIA% LA CIA% 18 SKI O i~ r 'Jh STAIR
~ 'I.lxl06 rads NS-RE 3C NS4E 30 j,pERTURE
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f I PSR-V-I77A/2 PSR V I77A/l I;ARD
~ ISTR RACK INSTR RACK
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KWR~?I ERR 63 ZONE P I
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~ E-IR-P033F TIP-V-l I
I WXSICAL NOT PKXITATAIEOF ACTOAXITIOIIE. I/ 73 9.lxt03 rads
~ 1.)xl06 rads Avails)o g~
L @PORC Cga g REACTOR BUILDING EL. IR493+
TIP SY-lh 501'-0'ENERAL E
NOTES: TIP-SY-2 1.PM.Oo I tl ARE IDENTIFIED IN GENERAL NOTES 2.3.4.557 ON DRAWING M-422 SHEET 1.
scuE NONE
- 2. SEE DWG. M-501 SHEET 2 OF 2 FOR COMPONENTS RADIATION ZONE MAP
.REACTOR BUILDING EL. 501'-0" OF LISTED COMPOSITES. ok/willa No I RII WASHINGTON PUBLIC POWER SUPPLY SYSTEM M -50'I 6
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~7A l6-DP 15-80 RIQ PS 198 as-v<78+
I6<P IS-9D RHR PS 19C NS~<78 IR-POX& NS Y47C+
E E-IR-P031+
SRN-W'-lA SRH EAHP 18 NS~67C E-IR-P03Z+ NS-Y470+
SRH E&-IC l6LC-FH-2+
l5LC-H FH/2 ~70 NSLC-Y-10+
RRC-V-168+
RRC~168 NSLC~IO ZONE F WSUBZONE K2 NSLC-Y-2A+
E-IR-64+ E IR-P033+ aSLC~ZA CSP-DPT-5 SRN-EINP 10 NSLC-V 28+
CSP-DPT-6 NSLC~ZB CSP RLY ARCSPVS ZONE M NSLC-V-ZC+
RHR-V-1 68+
CSP-RLT-IRCSPV9 NSLC~ZC RHRA-168
<<CSP4LY CRS asLc-Y-5+
RW-V-178+
<<CSPWLY CR6 NSLC~ZD RHR&-178 CSP SPV 4 NSLC-Y-3A+
RIQ-V-538+
CSP-SPY-5 15LC~3A RHR&-538 CSP SPV-9 NSLC-V-38<
'FM-SPV-32AI ZONE 0 NSLC~38 RFH-SPV-32AZ NSLC-V-3C+
NS V-19+
RN-SPV-3281 NS-NO 19 aSLCA-3C RFH-SPV-3282 NS-Y-28A+ NSLC V 30+
K-IR-66+ NS COHH-YZBA/JI NSLC~30 D5-PT-3 NSLC-VA+
NS-CDRN-Y28A/J2 CSP-SPV-I NS COW YZBA/J3 NSLC~4 RRC-V-16A+ NSLC-Y-Si NS POS Y/28A/I RRC~16A l6 POS Y/ZBA/2 NSLC~5 NSLC-Y-9+
NS POS Y/28A/3 ZONE I NSLC~9 NS-SPY-ZBAZ CSP-V-I+ RFV-V-65A+
l6-SPY ZBA3 CSP-POS-Y/I 16-Y-28b+ RfV~SR CSP-Y-2+
RFV-YC 58+
NS-CORA VZBB/J I CSP-POS-V/2 RFH~658 RIR Y4+
l6 COIIN VZBb/JZ NS CDNII V288/J3 RHR~B ZONE E IR-63+
Yi 16-POS-Y/Zdd/I NS POS V/Zbb/2 ZONE S RCC-Y-104+
APERTURE NS-POS-V/288/3 CEP-SPV-4A CKP-SPV-48 NS-SPV-2882 NS-SPV-2883 RCCA 104 RCC-Y-21+ CARD g ~ A 05-PT-4 RCC~21 NS-Y-28C+
RCC-Y-S+
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l6-POS V/28C/I RN-Y-53A+
CSP-SPV-6 NS-POS-Y/ZBC/2 RHt~53A 4Peetux'e GEEK RCIC-SPV-26 l6-POS V/28C/3 RCIC-SPV-5 RCI YAP@
NS SPV-ZBCZ E-IR-P015+ RICJ~<0 NS-SPV-28C3 l5-DP I 5-1 IA NS-Y-280+
15-DP I 5-8 1 OA 16-CONN-YZBD/JI scut NONE COMPONENT EOUIPMENT LIST FOR COMPOSITE EOUIPMENT SHOWN ON RADIATION ZONE MAP.
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c'ENERAL NOTES:
1.eg~.oo. t. ARE IDENTIFIED IN GENERAL NOTES ~ 1.3x 106 rads ~ 6.7xl03 rads 2,3,4,58 7 ON DRAWING M-422 SHEET 1.
RIVI-VAzb f BRIC-I 1+
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NSLCMY~/)5 NS PS 48A l6-Ll$ -31 0 NS-LITS-26C REA-SPY-V/I E IR 741 NSLCmV-Ca/IC 16 P~
NSLCWLYM/)0 NS-PT-51A IIS-PS-20C astc-ps-20 t6LC<LY-CR/SAI RPS-PS-2D I6-PS-23C NSLCPS-24 l6LC<LY-CR/SA2 16-PS&1b l6LC-PS-25 NSLCWLY-CR/581 NS-P~70 NSLC-PS-60 NSLCWLY-CR/582 RPS-PS-2 C NSLCNLY-CR/)
NSLCmv-CR/SCI 16LCNLY-CR/3 NSLCWLV-CR/SC2 NSLcMY-CR/4 ZONE 0 NSLCMV-CR/501 RRA-FC-I 9 NSLCWLY~/5 NSLC<LY-CR/502 NSLC-TO-TK/2 RRAW-Ftt/)0 NSLC4LY-CR/6AI IR-P004a NSLCmv-Ca/6A2 NSC 15-100A ZONE F 16-L 1s-24A NSLCWLY-CR/65)
RRCU-VA+ l6LCMMR/652 NS-L I 5-3)A RltcU~4 NSLC<LY-CR/6C) 16-L 1 5-31C NSLCALV-CR/6C2 NS-L ITS-26A ZONE G NS PS 20A t6LCMY CR/6D) pres' (~
RIVI-V428+ I6-PS-23A asLCMY-CR/602 Nstcmv-ca/8 R
RIVI~25 Its-PS-47A ZONE H aS-PS-47C RPS-PS-2A NSLC&LY-CR/9 NSLC-TO-TK/2A APERTURE aSt.c-TO-TK/28 C)A-V-308+
CIA~305 Stl-Y-75Aa Sw0-75A NSLC TD TK/2C NSLC-TD-TK/20 CARD ..
E-IR-6& NSLC-TO-TK/3A ROA4LY-CR200 ZONE N aSLC-TD-)K/38 RDA SPY-200 RRA-FC-I la l6LC TD-TK/3C E-IR-P021+
NSLC TO-)K/30 RRAA FN/11 t6-LIS-248 l6LC-TD-TK/4A 4vaggg'tenue~~"
I6-LIS 375 asLC TD-TK/ib Its-L 1 5-31D l6 LIS-385 ZONE 0 16LC-To-TK/4C NSLC TO TK/4D RIV)-VQTAi 16-L ITS-260 E-IR-F026+
l6-PS 20S RJV)~42A 16<15-240 l6 PS-488 COMPONENT EOUIPMENT LIST FOR ~t NONE COMPOSITE EOUIPMENT SHOWN ON RADIATION ZONE MAP, REACTOR BUILDING EL. 522'-0" gaaellsg lO WASHINGTON PUBLIC POWER SUPPLY SYSTEM M-522 3
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~ YH Io ~r" 3F" ~ I.OX)05 rads 1 J ~ c it 'O-TE tSUBZONE 01 \ OK)C'CR ~ C{CSKG COG.Rbs WATKR teE>>kt i'R CIIIS so '.I 5 H 54-0 C+At& YKP ' ~PAR ..Cet'TK RC IC-LS-6 -FC-'17+ CAC-FCY-ZAt CAC-V-2+ ~ CEP +OS-V/I 8 1.2xl06 rads KxCHA)IGKR r F S.SR 2. ttCB<K GAS ktu.TEED >>>> ZONE C '- 54IITS-HTP80 REA-AD-bt CEP-V-IA+ ' i tjG k'NHILrZFBS D ROA-AD-15 ~TS-SC CEP-Y-18+ I 4 1.)x)04 rads <<CNS-TS-SO ROA~AD/)9 CEP-V-2A+ ':Q'-'I >>I TRIA CEP-Y-28+ er ROA-SPV-15 <<CHS-)5418 ere / ee I (C-t{02 II " ~ I' ~ ~ RDA-SPY-17 <<06-)SIC >> 3 JZ e~ ~ I ~te5 SLC-Y-4A <<CRS-TS4ID ZONE L Rr ~ J K->4 I Rivi~)348 SLC-Y>>48 ~ ~15-9A ~. 1.54105 rads I <<CNS-TS-98 3'BK~~ 4 tT. <<06-TS-9C ~ ~ I 20 reer ttZONE E , RCC-V-129) PII I 1 ~ t ~ <<CNS-TS-90 'I 06-SR-13+ e RCC-Y-130) t <<Stt-PS-10)5 I er ~ ).OX106 rads RCC-V-13 I+ ~ k )ei1 <<Slav 206 ~~ Rtbt V )34bt 2 eo r)C <<SV-Y 209 1'1 (e 1 I 06-SR-13+ RRAWC-) 9+ 1 r i ~,I I <<St(-v-2)0 RRA-FC-20) I <<Stt-V-2) I IJ'TI) <<r Stt-Y-)87A) I I t r~ .'t 'I I 06-H-SR)3/15 ZONE G SII-V-)878. I I = <<06-H-SR)3/T6 = ~ (.t e I E-IR-674 St(-Y-) bbA) . I <<QIS N SR)3/17 Sit-Y )8Nlt ~ 1.2x)04 rads <<06-H-SR13/TB K G ~771 E-IR47) ~ i( I" i ZONE M t(o ) 4 4 SUBZONE El 1 ~4 5.0X)04 rads RHI~)34A 'ZONE H ItistR RACK CAC-BO-FCY/28 K IR e I' %RA-FC-)5) 4 1.64106 rads APERTURE Ii
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~ S'TA1R CACPCY-Zbo SKI I6-HTP71 I I F'-:. II ZONE N 'I ek L QoJ- I STA'4 <<CHS TS-.4A QIS-TS&b RRCHS TSWC CAC Y RCIC Jt-) 3+ RHR-V-23+ )lo ~ R%~3A ~ 2.)x)06 rads KLKY I ~ QIS-TS-4D i e'. ~ 'I I ~ Rtdt-V-Sl) I Q6 TS SA ZONE J t ~FGK{. P.~ ' QIS-TS-SP ~ RIVI~9 Rtbt VW) 8o AV~~e E RHR IIKA FtKL PCC'- ~ KICHkkGKOS~>>' ~ R K)CHARGER+ 0 KXCHAIIGKR RIR-V@%4 pe~ec IIK'KAT C OC OtterOS <<Sit-P 5-10 I 4 eet 'k roC-HX-'k&TB 1' P.tka. 8 PteR HX lB ~ 3.)x)06 rads <<Sit V ZO) d <<SH Y 204 RW-Y-) 15+ REACTOR BUILDING EL. <<Sit-Y 2IZ RCIC~ Rtdt-Y 116+ 548'-0'ENERAL ~ PS' Rtd) WOA 213 R% Y 38+ NOTES: I Rt>> V 75A YW f.e/bm,eo, t, ARE IDENTIFIED IN GENERAL NOTES ~ RIVI 2,3,4,587 ON ORAWING M-422 SHEET 1. 5{kit NONE
- 2. SEE DWG. M-548 SHEET 2 OF 2 FOR COMPONENTS RADIAT)ON ZONE MAP REACTOR BIRLDING EL. 548-0 OF LISTED COMPOSITES.
WASHINGTON PUBLIC POWER {et)&Iles eeo R{1 ~ 4
- 3. ttTHESE ARE ESTIMATED DOSES. DOSE WILL BE ~1 ~
SUPPLY SYSTEM M 548 VERIFIED UPON EQUIPMENT INSTALLATION. ~ I ~ 'e <<et' ee 2 @30~ 'I ~ I ) ~ ~ ,l ) 7~ ~ I I ~ 7) Ir) I II( I' ~ ~ (Ir VI V) ) I ~ C I)I I Il ~ ~ I ~ ~ ~ n L. I ~I A. I I I C Cl I I ~ ~ ~ .P Or 'Jt CI ~ .Lr ~ Pr ~ (0 I LI 0 p ~ Ll v V I Ir Ar S ~ ~ ~ Q C 7')I ~I LI ~ 7I C7 Al I I P~ ( ~P ~ ~ )CI I D il CO ~ 'I lv o ln'I w ~ I pP PP ~ ~ A f)W SOS-Pl OII JNWflf ZONE 8 R%-V-35+ ZONE P l iPPb r."I. >.90 RCIC-Y~+ R%~35 E-IR4I+ 3 Ai RCIC~ RN YQOf CEP-SP V-2A f RN-V )6Ai RHR~O CEP SPVQS ~ ( lh RNA 16A RN V+88+ CIA4tD6 lb iim.r!7 RN-Y '17A+ 5%~68 CI~Z)B PfAI" RNA-)7A RN-V-49+ CITS-225 5%~9 CIW Zlb ZONE E RN-TABB+ <<CITY 2)b 6 5 f/i A l 'f CHS-SR-I 3+ RN~ <<CIA-TDS-) 5 ( ~ Af i>S <<06 AY I 06-PT-2 OIS-AT-3 0LrPWi 06-PT 8 0 SUBZONE E 1 ZONE K CAC-FCV-2A+ R~Y~ REA-SP'Y-V/2 a%2 A-P ~
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<<ARAN FH/15 CAC-V-2+ DIARY-228 "6 Jf' CAC~Z ZONE F REA-AD-8+ 0IS-SR )4+ REA~AD/8 ZONE 0 CAC-FCF-15+ 0IS-AY-Z REA-POS-AD/8 ii0IS AY 4 CAC-EIO-F CV/I 5 CAC-V-15+ i SUBZONE F 1 RRA-FC-17+ ZONE L CAC~)5 <<BRA~FR/) 7 RCC-V-)29+ WSUBZONE 01 - I 'IKSUEP RECT-) 29 CEP-V- IAi ZONE 6 RCC-V-) 30+ CEP POS-Y/IA 'f I~ IR-67+ RCC~) 30 CEP-Y-IB+ ~ E I f'~ CEP-SPY-) A RCC-V-131+ CEP-POS-Y/15 CEP SPV-18 RCC~) 3) CEP-Y-ZA+ C'f CIA-PROD-)A RN-V-1345+ CEP-POS-Y/ZA ~ ~ CIM'5 21A RN~)348 CEP-Y-2b+ RRA-FC-19+ i/5 v CIA-PS-22A CEP-PDS-V/28 CIA-PT-2'lA RRAW-FII/19 CIAO-2)A RRA-FC-20 C IA4LT-22A RRA~FN/20 .~ - I pl <<CIA-TDS IA Sw-V-)874+ 0IS-PT-I SM~)87A i 06-PT-5 OIS-PT-7 SH-V-1878 Sv~)878 Sv-V-)BBA+ PRC ROAST-CR)A ROA-SP V-IOO SH~IBB'v-V-1888+ ~ t ~ APERTURE ZONE H CAC-F CV-25+ SH~)885 CARD CAC<IO-F CV/28 CAC-V-II+ ZONE M RN-V-1344+ CAC&-I1 RCIC-V '13+ RN~)34A hO AVaLh&1e ()2, RC)Coo-) 3 Ap~me C R%-V-23f ZONE N RHEA-23 RN-Y 3Ai 5%~3A RN-YgBAi ZONE J R%-V-115+ RN~BA R% V+84+ R%~44 RN-V ))Bf RN~ R%4) 93 COMPONENT EQUIPMENT LIST FOR +at NQNE "OMPOSITE EQUIPMENT SHOWN ON RADIATION ZONE MAP REACTOR BUILDING EL. 548'-0" g4AelhC Ifo g fIV WASHINGTON PUBLIC POWER SUPPLY SYSTEM M 548 3 8'8 0 7 0 V0 0 1'8-'gg ' ' (k f f2 ".0 5 SEE DYcG Y{. ~ ~ ~ g yQgr'e'~ 0 l col QEQGQV>> HV'-~~- it ~ rc c ~ c I )pt / I -il { %cjoy ~ I i i:X-" tc:llYf'. ~ cc~ (y I( ~ )'C A + ~ ~ ~'cg 5c: h I ~ hcc 'iCY r' ~ ~ ~ ( JQJ wc I (Ifli i I li"l 4C .\ I ~ ( 5r+ )+I ~ iY )i ccl p cubi ,'P ~ Ic I%i'L I '( ~ ~ ~ ~ c ~ IttCi pc ch gy g%e ~ -tn-"- 4-IC <".Cc J j lct'c L \tel .J ~ 8f'~"O'Ic< hv C 1 'Ccc <yg c tc 'l ') 8 }v' iB tc")~!I ~ Ngl 2"t-tr'I c ( v cYjvjn) <I. "c tc>~ ~ j ~ CR- 'Lf ~ (cl It wc+ I 5- t-'1'c i E \' 0 ~ N 5-.(-:p! ~ ~ ir I ~ ~ ~ $~ a Ca c-I ~ cc ) I'l lcl~ wcclI
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/r PRC K ~ii BB> ,APERTURE ~ J Lr 2 .'l CARD A1so AvaBable Ou 'Aperture Card 6PS'-10- GENERAL NOTES .. REACTOR BUILDING EL 1.~M,oo, I. ARE IDENTIFIED IN GENERAL NOTES BPTgjs Jt<R>j'.gt 2.3.4.657 ON DRAWING M-422 SHEET 1. RADIATION ZONE MAP NONE REACTOR BUILDItfG EL. 606-10-. WASHINGTON PUBLIC POWER .2'U>PLY SYSTEM . M-606 '4 h 0>>"-'~ ~ ~ b3 o v o70 . Il 5 .t )I 'M( )l I t k I J I 7 ~ I M 7 I II~ ~ It I )k M gl 5 k IIII.I U ILI ;X*I ~ 'I A 5, Ck M ~ 'k 'j II 'I I Y' I )5 k II I-I I I ( 7 1 7 I I! f" I '!I ( ."l 7 tl ,7 .1 II. I () c I )1 f, 7,N I. 7;Q I k t'i U pg I" 'Q 'l I 'w M,I Q Q I IJ II p I (77 I ) I 7 M( I tt K f I t ACCICEHT OEF IH IT IlW PAGE 2-9 ~ THESE ARE FOUR BREAKS CONSIOEREOi ALL ON ThE RCIC TINSIHE STEAN SUPPLY LINE 44 l ThE LEAK DETECTION STSTEN PROVIDES tHE CAPABILITY OF DETECTING THE BREAKS ANO INITIA'IES ThE CLOSIHG OF ThE APPROPRIATE VALVES TO ISOLATE THE BREAK+ tHE PLANT CAH PROCEED TO HORtIAI SHUTOOMN OR COtITIHUE OPERATIOHe ;Kt RCIC BYLINE. BREAK CST LEVEL LQI CR QPPRESSIttt OES IRIDO TO RCIC PQL LEVEL hlGH ~SSICt4 ~ 'ROVIIXS hEAT SINK BUILOI I ttl MIN I ZE RELEASE CF POQ AN SIGNAL I QX.ATI Q4 RADIOACTIVITY A VALVES P Al A2 Bl 82 RES PRINARY CRYI ELL PRESQWIE I/2 THICE MATER QPPLY IN IT IAI VESSEL LEVEI tent DIFFERENT FRIT 5KACLL RB IHLET KEEP REACTQI BLOOo QPPRESS IO4 CORE RESTIRED d EXHAUST PR ItIARY CXLIW HVAC AT &Go PRESQJIEo CCYITAIttENT ISOLATION LPCS gLPC I MATER StPPI.Yo~ / A ISOLATES tH I/3 ISIXATIOH SIGNALS SAQCR QPPLY EXCESS 8 RIR PRE SQNE NS IV'S LEAKIWY QPPRESS I Ql RELIEF PIXL CQLIHG HAINTAIHS PQX EPKRGEHCY MATER INVENTINY TB%%RATVIE REACTIà NAKELP Attt SHUTOOIt4 SPRAT CXL it% NS I V i LCS DIRECTS LEAKIttl HIGH PRESQNE ALTERNATE PATH FOR SWTOQtt ( ECCS ADEQUATE YES l&~~) ~ l StgJTOQtt 'XX.Ittt AVAILABLE, COOLIE IUSEO IF SWTOOt4 COCt.lttt IPIAVAILABLE OR IP4XS I RABLE 1 A 8 t tXXIITOR POQ TEt% ~ A 8 STEAN TO SOTS HIGH -'$ 4$ TQN RHR QX'PXESSIQ4 PTXL COCLIHG tKBXO SOTS NAINTAINS REACTOR CXLIW CQIE CXLIHS BLOGi HEBo PRESSURE N g F ILTERS CONTANINANTS . VIA RIR hX AUTO~SSe Ftà A, 8 A 8 PRIOR TO RELEASE LQ4 PRESSURE ECCS tlAINTAIHS PIXX. PROV I CES FLQPATH FRON A 8 RCS, TO 5LPPAESSI QI PQL TEtPERATVIE PREVENT ION CF A 8 SIGe RELEASE OF RADAR NAT'L TO ENV IRQtKNT M MATER PIPPEO FRCtt POQ 'PRj~ TO Rttt hX TO RPV A 8 CAPITA lttKNT I INTEGR I TY COLD: SIWTIXXtt ACHIEVED LPCS CQ4E SPRAY LPCI ~T tRESIRIAL CCRE REtXIVAL 'APEpTUpF VMS HIGH CORE FUXX)1tXI A 8 C CARD REACTOR AlSO AVailtlblC U4I CORE COOI.INB 'perture Cord FIGURE 2. I REACTOR CORE ISOLATION COOLING .SYSYRN LINE BREAK . /* SAFETY SEQUENCE DIAGRAM 9$ OVOIO 6 yB F 4 t ~P 1 L a v , 4 ~ (tÃJ f.l l~ ACCIDENT ~INITIDN SEVEN RHCU LINE BREAKS ARE COHSIOEREO IN thE ANALYSIS ISEE TABLE PAGE 2-IO 2+1 1 ~ ISOL.ATION IS INITIATED BY thE LEAK GETECTIOH SYSTEN OH RHCU AREA hlGH TEHPERATURE DELI'A TEMPERATURE OR DELTA FLOH IHE PLANT CAN PROCEED TO HORHAL. SHUTOOKN OR CONTINUE OPERATICH RIICU LINE'BREAK CSt LEVEL LOH CR QPPRESS ION CL.OSE POOL LEVEL hl GH QH%ESS I CN HlNINIZE PIXY PROVIDES I%AT Sf HC BUILDING RELEASE CF I SOLATIOH Al A2 81 82 I/2 GRAN SIGNAL THICE A 8 VALVES CST ~ PR IHART HATER SLPPLY QPRKSS ION .'N I TIAL VESSEL LEVEL P NO lÃYhELL FRESHIE OIFFUKHT FRCH &T~ RB I)4XT RAG I OAC't I V I TY KEEP REACTOR BLCOs CORE CLXLING RESTLYLEO I EXHAUST AT NEGUS PRESQREo LPCS.LPCI HATER hVAC PR I tLARY RELY o )PCS TES ISOLATES ON I/3 HSERT RODS CMTAI~NT BACKLOG QPPL.T " A I QLAl'ON SIGNALS EXCESS 8 SPPRESS fLYJ PRE SSLRE HSIV '5 REL.IEF PLXL CIXLIW HAINTAINS PIX. LEAKING'ES ENERGENCY uhtER INVEHtORY N IEJPERATVLE REACTCR NAKEDLY JQ4) SWTDSA SPRAY COOL ING HSI V LCS hl GH PRESQRE ALTERNATE PAtH FOR SHJt004i DIRECTS lEAKING ECCS AOEOUATE STEAJI TD SGTS SHJTOCSA COOLIE (USED IF SWTOQA tCYJI TOR COCLllQ COCLI JXJ WAVAILABLE CR POOL TEJY ~ A 8 YES AVAILABLE tP4XS IRABLEI A 8 Q4JTOOfA SIJAKSS I CH POOL IXXLI I% lKEGEO SGTS HAINTAINS REACTOR COOL. I NG BLDG+ HEBE PRESSURE ~ Vlh le hX CORE COLLINS YES FILTERS CONTANINAHTS PRIOR IO RELEASE AUTO-CEPRESSo FOR A 8 LOH PRESQNLE ECCS PROVIDES F~ATH FJKN Hh INTA IHS POOL A 8 RCS TO QPPRESSIOH POL TBPERATlÃE PREVENT I DH OF A 8 SIGi RELEASE CF RADAR HAT'L TD ENV I RODENT HATER PLPPEO FRCtl PLXL TO RJR hX TO RPV 8 CCtfthl t4KNT ) I NTEGR I T Y CLLO Q4JTOQOI ACHIEVED IL"'" PRC CORE RESIDUAL CCRE SPRAT &AT REtCVAL HIGH YOUP% CORE FUXOIHG 'APERTURE A 8 C CARD REACTOR CORE COOLING ~o Available 0~ MpcztUZ'e Cary FIGURE 2 2 REACTOR HATER CLEANUP SYSTEM LINE / s aAWAK SAFETY SEQUENCE DIAGRAM l 88 0 V 0 V 0 0 1.8 -3B Is s uI s ~ ~s h ~u u ~ up ACCIDENT DEFINITION PAGE 2-l l TMO BREAKS ARE COHSIOEREO FOR ThlS EVENTS OCCURRIHG OH 3 ANO 4 AUXILLIARYSTEAN SUPPLY LINES TO thE REACTOR BUILDING hEATING SYSTEHi HITIGATIOH OF tHIS ACCIDENT MILL REQUIRE OPERATOR ACTIONS IH RESPONSE 'TO A LOM-PRESSURE ALARHi TO ISOLA'fE THE BREAK ANQ PROCEED TQ COLD ShUTOOMHe AS LINE BREAK CST LEVEL LCAI DR RPPRESSIDH PCXl. LEVEL HIGH SUPPRESS I OH POOL PROIIIOES hEAT SINK P CST HPCS PRIHARY WAELL PRESRRE MATER SUPPLY INITIAL DIFFERENT FRCtt IEtiKLL CORE LPCSsLPCI MATER CCDLING RES TNIED YES SUPPLY o hPCS BACKUP SUPPLY CVB EXCESS PRESSURE QPPRESS IM RELIEF PQQL CQIX.ING ~ HA IHTAINS POOL MATER INYEHTORY H tEHPERATURE HAKEUP ANQ SPRAY COOL.ING 'ALTERNATE PATH FOR SHUTDQMN SPTH REACTOR SHUtQQMN COOLING tUSEO IF SHUTDOWN tXXtI TDR COOLING COOLING UNAVAILABLE OR POCL TED+ CORE COOI ING AVAILABLE URGES IRABLE I A 8 SUPPRESSION POOL , SHUTOOMH COOLING N EOEO COOLING CCRE CIXLING YES H YIA RHI hX NAIHTAIHS POOL PROVIDES FLRPATH FROtl t EKPERATURE RCS TO SUPPRESSION POOL A B QR MATER PUNPEO FROH POOL TQ RHR hX TO RPV H B CDNTA I tJKNT CKQ StAJTD&tl ACHIEVED A INTEGRITY PRC CORE RES IDUAI. hEAT REtCYAL APERTURE: CARD UBo Availab}e 0+ Aperture Carp FIGURE 2 ~ 3 AUXILIARY STEAM SYSTEM LINE BREAK-SAFETY SEQUENCE DIAGRAM -r, .SSQV0700 y B a a K 4 ACCttXNT IXFINITION PAGE 2-12 'fhlS ThE STEAN ~e EVENT CONSISTS CF A BREAK IN h HAIN STEAti LINE l2S I fNSIOE ~ BREAK IS ISLXJLTED BT tLSIV CLOSLÃE IRON RECEIPT CF A HIGH STEAN FLQI SIGNALS AFTER t&ICH TIE PLANT PROCEEDS TO CQJI SHJMQAo NS LINE BREAK CST LEVEL LQL CR QPPRESS ION OKSIQKD TO PLXL LEVEL HIGH QPPRESS I CH REACTOR NIHIHI ZE POQ. 'ROVIDES &AT SIW BUILDING RELEASE OF CLOSE ISOLATION ~ AH SIGNAL YAI VES lf RAOIOACTIV I TT A P 8 CST Al A2 81 82 A PRIHARY MATER SLPPI Y INITfAI CRY~ PRESSURE I/2 TMICE VESSEL. LEVEL HO DIFFERENT FRCH ICTl4JJ. RB INLET KEEP REACTOR BLDG+ QPPAESS ION CORE RESTORED 4 EXHAUST COOLIE KVAC AT NEGUS PRKSSLNEi PR I HART LPCSiLPCI MATER vn IQLATES ON I/3 CONTAlttKNT SPRY+~ ISOLA'fI OH SIGNALS BACKLP QPPl.Y 8 I SILATION FLITTER SR PRE SRRE HS I V 'S LEAKINGY QPPRESS ION RELIEF PQX. COOLING HA INTAINS POQ A EtKRGENCT MAtER IHVENTQLY H TBPKRATLXLE REACTOR HAKKLP AM) A 8 SWTOSt4 SPRAY CQLING HSIV'CS HIGH PRESSLRE VESSEL CEPRESQNI ZEO DIRECTS LEAKING TO h POINT Al'ksICH SPTH STEAN TO SOTS ECCS AOEQIATK YES DEPRESS REQIIRKO I ZATION SWACKED CQLING INI TIATED IS ICHI fOR PQL TEIV ~ A 8 A 8 NIGH REDUCE REPRESS ION PQX. COQ.IW l4XOED SGTS HAINTAIHS REACTOR REACTOR BLDGi HEBE PRESSURE PRESQNE YES FILTERS CONTAHIHANTS ADS A 8 IRIOR TO RELEASE AUTO~PRESS+ FOR A 8 LQL PRKSQRE ECCS OEPRESQXI IZED tlhtNTAIHS PDQ A 8 TO PR5, 135 PSIG IEtPERATURE PREVENT ION OF A 8 S I G RELEASE CF RAGE tLAT'L TO ALTERHATE PATH FOR SWTKSR ENV I ROttKNT SI4ITÃSA CQLIIXI LUSKD IF SHJTOOW CQLINI COOLING WAVAILABLEQl AVAILABLE IP4KS IRABLE ) CCNTA IDENT %R INTEGRI TY SINTDON4 COQ.I W CINE CQLING I'="' R C H Ylh AR HX CORE SPRAY A HIGH YRIPK CORE FLOOD I IS A PROV IDES FLLSPATH FRCH RCS TO QPPRESSICH PQL APERTURE A 8 C CARD MATER PLP%%D FRQL POQ 10 RIVI HX TO RPV REACTOR CORE COOLINS A 8 AIso Avazlable Olz CLO SMITOQA ACHIEVED '%pnme ~a CORE RESIDUAL tEAT REHOVAI , Og FIGURE 2 ~ 4 n HAIN STEAH LINE BREAK SAFETY SEQUENCE DIAGRAH 83'ovo700 y g - g5 "c qv' p Jlq f 4, C~ le~ J C.,i P e ~P' ACCIDENT IXFINIT ION PAGE 2 13 a THIS EVENT CCHSISTS OF A BREAK IH A REACTOR FEEC44ATER LIP< 124~$ INSIDE THE STEM TIPtKLo THE BREAtf IS ISOLATEO BY ft4IOARO A$4) QJTBOARD LINE CHECK VALVESa AFTER $ 4$ ICH THE PLANT PROCEEDS tO CCLD S$ 4JTDOHHo RFII LINE BREAX CST LEVEL LQI OR QPPRESSlftt REACTCR DESI GIZMO TD PQL LEVEL HIGH QPRKSS I GH H IHIHIZE I SQ.AT ION PIXL ~ PROV IDES HEAT SI$ 4C BU I L01$41 RELEASE OF RAO I OACT I V I TY RAH SIGNAL, A VALVES P CST ttaCS PR INARY CRYIELL PRKSQXIK Al A2 81 82 I HI T I AL MATER QPPI.Y DIFFERENT FRY METI4JJ. RB INLET I/2 TM ICE QtaPRESS I 0$ CORE VESSEL LEVKI RESTORED EXHAUST KEEP REACTCR BLDG ~ PR I NARY COOL! NG HVAC At HKGo PRKSQREa CONTA MKNT LPCSaLPC I MATER YES JQXATKS DH I/3 I SILATION QFPLYa~ A IQXJI 1'ON Sl GHALS HSERT RODS BACKLta QPPLY EXCESS A 8 Rttt PRESSUIE NSIV'S LEANINGS' QPPRKSS IQ$ RELIEF P$ XL CDQ.IHG HA INTA INS PQX. H TKttaERATLRE EHERGENCY MATER INVENTORY REACTOR HA$$ KUP A$41 A 8 S$ 4JTOQtl SPRAY CIXLIHG HSI V LCS DIRECTS LEAMI$41 h I BH PRKSSVIE VESSEL CEPRKSQ&I ZEO FURT$ 4tt TO A POINT AT $ 4$ ICH SPTtl STEAN TO SGTS KCCS AOEQJATE tCNI tDR DKPRESQtt I Z At l(tt S$ 4JTC0$ 4$ COCLI $ 4s I S YES RKQJIRKD aN! TIATED PQX. TKttaa A 8 A 8 QPPRKSS ICH POQ. COQ. IHG NEEDED SGTS HA INTAINS REACTOR BLGGo NEGa PRESSURE TES FILTERS COHI'AHINANtS PRiOR 'fO RELEASE AUTO-DEPRESS a Fat A 8 LQI PRKSQRE ECCS HAINTAINS PIXL OKfaRESQH I ZED 8 'fD PRaa 135 PSIG TEtPERATIXIE PREVENT I OH OF A 8 SIGa RELEASE CF RAD~ HAT L TO RCS PRESQRD 100 PSIG ALTKRHATE PATH FCR SHL$ TO0$ 4$ ENV I RCttKNT QCJTDQtf CIXX.ING IUSEO IF S$ 4JTDQtf CDQ.I$ 4$ CDCL1$ 4$ WAVAILABLEOR AVAILABLE tP4JKS I R ABLE ) CATA I ftKPIT ( INTEGRI TY S$ 4JT00$ tt COQ 4l CQIE CQL Ittf 1$ H VIA RHt HX CQ$ E SPRAT PROVIDES FLQtaATH FRQ$ hlGH VQAPC RCS TO S~$ ESSIQI PIXX H CORE FLO001$ 4I A 8 C MATER PlPfaED FROH PIXL TO Rttt hX TD RPV A 8 ~o Av~ahle O REACtOR CORE COOLING COLD S$ 4$ tlXWI ACHIEVED 'lqpmh e 6'ORE RESIDUAL HEAT REtCVAL FIGURE 2 ~ 5 REACTOR.FEEOWATER SYSTEH LINE BRLAK SAFETY SEQUENCE OIAGRAH 88 07070 0 I 8 -3/P te'l, KP '> P,> .".) m ' hm C'r'O v- ACCI(KHT GEF INIT ION 124 I IN ~ THIS EVENT CONSISTS OF A BRKAIC IH A RKCIRCCATIGH PtPP QX:TIOI I.INE CRYl4lJ BETlKKH AE REACTOR VESSEL h)47 tC)TGR~ATEO VALVE RRC-V-23, h)X) IS ThERKFCNK DNSIGERKD MW-ISOLABLE DE HA)H LARGE BREAK LOCA PAGE 2-14 PMAlKTER TRANSIENTS Ht)CH CHARACTERIZE ThE EVENT ARE UXI RKACTIXt LEVEL h)X) PRKSQXIEo h)X) hlGH ORAKU PRKSQXIEo lhE PLANT PROCKEOS TO I REC IRCo ) CCLO SWTOQtlo CST -LEVEL LCXt CXt SLFPRESS I GH PR I HARV GESIGPKO TO POOL LEVEL HIGH REACTOR ~SE A CONTA IMKNT IXWAIHS INEAX IXIILO I NG HINIHI ZE VOLLEY III Otal. RELEASE OF A Al A2 Bl 82 I/2 SCRAH SIGNAL MICE ISOLAT IOH VALVES TIP CLOSE IQLATIGH P CST ~ MATER PRIHARY QPPLY IHIT IAL CORE VESSEL LEVEL PROV I GES P RB INLET RAO I OACT I V I TY KEEP REACTCà ELOOo I CLOSE RES TGREO fghl'IMC P)PK L EXHAUST g I~TIGH VALVES LPCS,LPCI MATER COO.IHG FOR HVAC AT NEGo PRESSUKo A VALVES ROPE ofPCS BREAK IH DRYlKLL A IQLATES GN I/3 INSERT BACKLP RPR Y I SCLAT ION S I GHALS h 8 8 ROOS Pl RHR CLOSE I QLATI GH SUPPRESSION tS IV'S LEAKI)X)2 VAI VES OOL COOI.IH HAIHTAIHS POOL RfR RKOUCES CLOSE MATER INVENT REACTCN H lEHPERAI'URE I CCHTA II4%HT TGMo A)X) E URGE CY SPRAY ) REACTOR I QLATI GN HAKEUP h)X) GEPRESQS I ZATIGH PRE SQhtE YES SHJTDOA VAI.YES A SPRAY COCX ING H FGLLGMING CLOSE LOCA HSI V I SOLA T ION HIGH PRESSURE LCS DIRECTS LEAKING VALVES DRYlCLL PRKSSNIE CLOSE ECCS AOEOUATE ALTERHATE PATH FIà SWTIXXN OIFFKRKHT FRQC H STEM TO SGTS I QLATIGH VAI.VES YES SHITOSH IXXX.IIX) QXLI)X) IUSEO IF Sth)~ COGL ING IXIAVAILARE Qt METkKLL CLOSE AVA)LADLE Q4XS I RA)LE ) I QLATIGH TES VAI VES 8 f5R CVB EXCESS SOTS A QiJ TOST( PRE SSIhtK HA INTAINS REACTIXt. CLOSE RELIEF VIA ILOGo NEGo PRESR/IEo IRLATI GN CGXX.ING H CIXtE COCX. I NG ~CIC VAI.VKS F ILTERS CGHTAHINANTS VAI VES HA INTAIN Vlh l5R hX A 8 PRIOR TO RELEASE h 8 CGHTA I No CORE SPRAY I SCLAT ICN AOS A PROVIOES FLOMPATH FROH PREVENT IGH OF A 8 HIGH VOLUIIE RCS TO SUPPRESSIOH POOL PROV I GES SIG RELEASE CF CLOSE CORE FLOOOIH RAOo HAT'L TO I QLATIGH H2 H)XIHG KNVIRCttKNT VALVES A 8 C CAS HA I NTA IH A 8 CGNTA IN o Rltt )SLAT ICH MATER PUHPEO FROH POOL TO RHR hX TO RPV H A 8 CLOSE REACl'OR A 8 STARTS I QLATIIXI CORE COOLING COLO SHUTOOMH ACHIEVEO CRA VALVES FAHS HA INTAIN CATA)No A 8 IQKATIGH CORE RESIOUAL CLOSE hEAT REHOVAL I QLATIGH VALVES CAC CSP CIA HA INTAIN H2 RKCOHBIHE. PISGE QPPLY CGHTA INo H H CLOSE I QLATION I ECLAT I GH A 8 VAI VES A 8 CEP A 8 CONTAIN%M AIR TO SGTS HA INTAIN SPTH I CGNTA INo HCHI TOR RRC I SLAT I GH PNX. TEtt . Il 1 llh INTA IHS REACTOR I QLATIGN A 8 4 A 8 BLOGo %Go PRESSo SOTS VALVES ( F ILTERS ATM INANTS SUPPRESSION POOL PRIOR TO RELEASE COOLING HEEOEO I I HA INTAIN A 8 CGHTAINo YES IRLATIGH HAIHTAIHS PGCX. TEIY o HA INTAIN A 8 I SATION CATA INo A 8 Also Available Ou CCNTA IttKHT HA INTAIN INTEGRI TY CGHTA INo IQLATICN A 8 PRI HART CQCA I ttENT I SGLAT I Ql LARGE BREAK LOCA (REACTOR RECIRCULATION SYSTEM) SAFETY SEQUENCE.OIAGRAM BBOVOVOO 1.8 -4l I* ~ 3 a 7 I la C,~ 0 c ACCIOEHT OEFIHITIW - THIS EVENT CONSISTS Ià A BREAK IN A HAIN STEAN LIW l28 l BE~EN DE REACTIN VESSEL hi@7 FLM LIKITERo SIMX THE BREAK IS IHSIOE PAGE 2-15 ,%'HE )%CARO CQiTAI)tCMT ISOLATION VALVE, RAOIATIQl JQ47 ~ATLHE LARGE BREA)t LGCQ ~ELEPHANTS IN THE HAIN STEAN TIPt43. AK) INXJTIM) AREA ltlLL ICT BE I HS) >TW PR)NARY FACTORS RESPCtOI)4) ~ BUT RAPER RPV LEVEL Al4) PRESQNEo AM) hlGH BtA43l PRESSIVE ~ thE PLANT PROCEEDS TO CILD SHU~o CST LEVEL LOM OR SLPPRESS ION PR I NARY REACTOR OESIQKO TO FOIL LEVEL HIGH CCNTA 1 IPKHT BUILDING HINIHIZE RCI C CLOSE A lRHRL RELEASE OF IN A At A2 81 82 I/2 SCRAH SIGNAL TMICE I SOLAT ION VALVES CLOSE-IQLATICN P CST ~ PRIHARY MATER SLUR Y SLPPRESS I OH INITIAL CORE VESSEL LEVEL RESTORED P SLPRIESS I QI PROVIOES &AT P RB It@ET L EXHAUST RAO IOACT I V I TY KEEP REACTOR ILDGo CLOSE COOLING PIXL SIW FOR PIPE HVAC AT HEGo PRESSLRE ~ I QLATION Vht VES LPCSoLPCI MATER BREAK IQLATES CN I/3 VALVES QlNoLY o)PCS P IN GRTlglJ A I QLATION SIGNALS INSERT A BACKtl'PRY 8 A RfR HSIV'S LEAKIEST RODS Pl CLOSE SLPPRESS ION I QLATI CN PIXL CIXLItX) HAIHTAIHS POOL ltEOUCE5 VALVES REACTOR H TENPERhtURE I CCNTA I ft&IT IEl&o AQ) EMERGENCY RHI CLOSE MATER INVENT CEPRESSLS f ZED SPRAY ) PRE SSLIIE REACtOR I SCLAT ION HAKELP AN) SPRAY CIXL IHG H FILLCMI)X) YES SHUTOOMN VALVES LICA HSIV CLOSE A 8 LCS I QLATIOH HIGH PRESSURE ALTERNATE PATH FC)t SHJTDQiki CRY~ PRESSLRE 0 IRECTS LEAKI NG STEAN TO SOTS VALVES ECCS AOEOUATE SHJTD(X4i CITING IUSED IF SHJTOQA DIFFERENT FlKtt CLOSE OILI HG CIXLING LNAVAILABLE OR &T&LL A 8 I QLATIQI YES AVAILAILE INKS IRABLE ) VAI.VES CLOSE YES A 8 I QLATI ON VALVES SWTDSA t EXCESS SGTS HA INTAIHS REACTNI A 8 CIXLIHG PRE SQNE CCRE CIXLIQI RELIEF VIA ILIX)~ %Go PRESSLREo CLOSE N Vlh J5R F II-TERS ~TAHIHANTS IQLATION HX A ~CK VALVES PRIOR TO RELEASE VAI VES HA INTAIN A 8 CIXITAIN o CORE SPRAY IQLATION PROVIDES FLOMPATH FROH RCS TO SUPPRESSION POOL PREVENTI(W OF A 8 HIGH VILUNE PROV I CES SIG. RELEASE OF RAOo HAT'L TO CLOSE CORE FLOOOI h2 HIXIHG ENV IRQ4%HT I QLATI VI A 8 C VALVES A 8 HA INTA IN MATER PUNPEO FROH POOL A 8 CCNTAINo TO RHR hX TO RPV H I QLATION A 8 A 8 COLO SHUTOOMH ACHIEYED CLOSE REACTOR STARtS I QLATION CORE COOL IHG CRA FANS VALVES HA INTAIH CCNTA INo CORE RESIDUAL A 8 I QLATION HEAT RENOVAL CLOSE A 8 I QLATI CN VALVES CIA NA INTAIN H2 RECONBIHER H CCNTAINo CLOSE I QLATION I QLATIM Yhl VES A 8 EXHAUSTS CEP CONTA I QKHT A 8 AIR TO SGTS tlA]NTAIN GENITOR CCNTA INo I QLATICN RXL HA INTAIHS REhCTCR CLOSE P BLOGo )oEGo PRESS. SGTS I QLATI IXI A 8 TBPo'UPPRESSION F ILTERS ~TAHINAHTS Vht YES PIML PRICR TO RELEASE CITING HEEDED ! A 8 HA INTAIN CINTAINo YES I QLATI QI A 8 HA INTA IN CATAIHo I QLATI ON A 8 CQITAIt4%NT NA INTAIH fNTEGRtTY CIXITAI No I QLATION A 8 PRI)IARY CONTAI&KHT I QLAT.IOI FIGURE 2 7 LARGE BREAK LOCA (HAIN STEAH) SAFETY SEQUENCE OIAGRAH 8307.ovop zs -gg $ + k'..fji$ $ C I'~ f: v~ g K 1~ ~.- + ~A ACCI SYSTEM ~ T DEFINITION LEAKAGE IS GREATER thAN DE CAPACIT'Y FCR ~ SNALL BREAK LOCA !S DEFI%X AS ANY BREAK !N AE REACTCR COOLANT HGRHAL REACtCR CIXLANT MAKEIl SYSTEHIS) ~ NO SPECIFIC BREAK IS CCNSIGEREO FERE>> BUT RhthER DK GVKRAL EFFECTS GF ALL MS IQLATABLE SHALL SHALL BREAK LOCA PAGE 2-16 BREAK LOCAS>>AE PLANT PROCEEDS TO CGLD SHJTDQ4I>> I CST LEVEL LlR OR SLPPRESS I GN PRIMARY REACTOR DESI&TO TO PIXL LEVEL HIGH CGHTA lhtKNT CCNTAIHS SIE/JC BUILDING HININIZE RCI C VOUl% RELEASE OF CLOSE IM Dtt)GJ. ISOLATION P P RADIOACTIVITY A VALVES SCRAN PCS PRIHARY AI A2 Bl 82 SIGHAL TIP MATER QPPLY IHIT IAL VESSEL LEVEL RB INLET CLOSE QPPRESS I ON PROV I GES ?Eht KEEP REACTOR aalu. ~ I/2 TMICE CLOSE IQLATIGN P QPPRESS I OH CORE COOLING RESTORED POOL SI?44 FCR PIPE 4 EXI4AUST HVAC AT NEG ~ PRESQRE>> IQLATICN VALVES LPCS>>LPC I MATER BREAK IQLATES CN I/3 A VALVES QPPLY>> tPCS p P IN GRTI43A A I QLAT ION SIGNALS IHSERT BACKS QPPLY A 8 h 8 RHR '3 RODS Pl CLOSE MS I V LEAKIEST I SGLAT ION SUPPRESSION EICRGEHCY CLOSE YAI VES MATER INVENT REACTOR GEPRESQR I ZATI CN H OOI. COOL IN HAINTAIHS POOI TENPERA TURE I CGNTA I~r J?H4 SPRAY ) I?EOUCES TE?P>> h?X) PIIESSIRE REACtOR I QLATICN NAKEIP h?4) YES H FOLLOW I HG Q4JTOQH VALVES SPRAY CXLING LOCA NS I V CLOSE LCS ISOLATION HIGH PRESQRE VESSEL DEPRESQR I ZED GRYICLL PRE SQRE DIRECTS LEAKING YAI-VES TO POINT AT ?44ICH DIFFERENT FRCH STEAN TO SOTS CLOSE I QLATICN ECCS ADECUATE YES FURTHER OEPRESQR I ZAT ION SWTDQA GXLIfQ IS iKT~ A 8 REIX)I RED INIT I ATEO VALVES CLOSE YES A 8 I QLATION RCS PRESSURE VALVES HIGH LOM EXCESS SOTS MA IHTAIHS REACTOR A 8 PRE SSIRE RELIEF YIA )LOG>> %G>> PRESQRE>> CLOSE F ILTERS CGN'IANIHA?4TS ISCLAT ICN DECK VALVES PRIOR TO RELEASE VALVES A 8 A 8 MAINTAIN AUTO-DEPRESS F(R A 8 CERTAIN>>
- UX4 PRESQRE ECCS IQLATION GEPRESSIR I ZED A 8 TO PR5 135 PSIG PREYENT I GN CF h 8 ALTERNATE PATH FOR SHUTDOWN PROV I GES SIG ~ RELEASE GF COOLING IUSEO IF ShUTOOWH RAO>> NAT'L 'IO CLOSE I QLATION SWUTOOMH COOLING COOLING UNAVAILABLE OR IRABLE)
H2 MIXING ENV I RCPORT VALVES AVAILABLE U?40ES HA IHTAIN A 8 CGNTA IH>> I QLAT ION A 8 LPCS CLOSE IXRE SPRAY CORE CmLING STARtS I QLATIGN Vlh RHt hX CRA VALVES A FANS MA INTAIN HIGH VOLUME CCNTA IN>> ADS A 8 CORE FLOOD IH I QLATION PROVIDES FLOWPATH FROM CLOSE A 8 C RCS tO SUPPRESSION POOL IQLATIGH VALVES CIA MAINTAIN H2 RECOMBINE CCt4TAIN>> REACTOR MATER PUMPED FROM POOL CLOSE I QLATION CORE COOLING TO RMR HX TO RPY I QLATION A 8 VALVES A 8 EXHAUSTS A 8 COLO SHUTDOWN ACHIEVED CXRTA I ~MT AIR TO SGtS MAINTAIN HONI TCRS CCNTA IN>> IQLAT ION CORE RESIDUAL POOL TEt% J MAINTAINS REACTGR CLOSE HEAT REMOVAL SOTS IQLATI ON BLDG>> PEG>> PRESS>> VAI VES A 8 F ILTERS DX4TANIHATES SUPPRESS I OH PCOL PRIIR TO RELEASE COOLING HEEOEG, I A 8 NA INTAIN CCY4TAIN>> YES I QLATION A 8 MAINTAIN CGNTAIN>> I QLAT ION A 8 CONTAIN Hh INTAIN INTEGRI TY Also Available 0)I CDNTA IN>> >> QLATI @4 G@J 'Ikpeeue 4%1MARY CONTA I ttKHT I QLAT ION FIGURE 2.8 SMALL BREAK LOCA SAFETY SEQUENCE OIAGRAM 83'0707 0 0 yg -39 'v ~~ la b'L" ls,~ (~ )1 I PAGE 2-17 ACCIGEHT GEF INITION a THIS EVENT CONSISTS OF A FAILURE OF THE CONTROL ROG-TO-ORIVE MECHANISM COUPLIHG AFTER THE CONTROL ROO BECOMES STUCK IN ITS FULLY IHSERTEO POSITION+ thE CONtRI. ROG ORIVE IS 'thEH FULLY MITHORAMN BEFORE THE STUCK ROO FALLS OUT OF THE CORE INItIALLY AFTER THE hlGH FLUX SCRANiCORE COOLING IS PERFORMEG BY RCIC OR hPCS thE PLAHT THEN PROCEEOS tO COLO SHUTGOMH CONTROL ROD DROP ACCIDENT REACTCR MINIMIZE RELEASE < RB ) I BU ILO le OF REACTIVITY START CCHTA INHENT HIGH MAIN'TEAM 'NC IGENT 'S/RCI C NEUTRON AM) REACTOR HOH I TOR I HG ~TRCtt FLUX LINE RAG I AT ION HCtt I TCR IHG OETECTIGH GH UXI Qa VESM. ISOLA I'IM SYSTEM CIRCUITRY MATER LEVEL CQITRlL SYSTEM RB IMMIT KEEP REACTIXI S I GHAL SYSTEM EXHAUST BUILOIM At hVAC IKGo PRESQjK I RhRS HEAT REACTOR PROtECT ION SCRAM S I GNAL GN tKUTRCH CATA I~ AM) REACTOR VESSEL I SKAT ION 0+) REPRESSION PIL CIXLIHG t%XK I I EXCHANGER RHPS PUMP STANBY SERVICE MATER PUKP CSITA INtKNT (PASSIVE) HSI V 'S SYSTEM HEI TGR IHG CCHTIKL SYSTEH HA INTA!N LEAKING Y SYSTEM RCIC CORE TRIP COOLING HS I V SUPPRESSION POOL LCS TEMPERATURE LINII'TART OEPRESSURIZATIOH INSERT HAIN STEAH Cij4TA IHHENT CONTROL ROO C CHIRK LINE I SKAT ICN 'INTEGRITY ORI VE SYS'IEM RX)S FILTERS VALVES CCHTANINATES SGTS PRICR TO TRANSFER GECAY hEAT TO RPPRESSIGH POCL RELEASE REACTOR CGHE PREVENT I Ctt GF
- COOL IHG SIGo RELEASE IF EMERGENCY REACTOR SHUTOGMN CCHTA I ~T PRIMARY I SILAT I OH RAG+ HAT'L TO ENV I RQ+KHT MAINTAIN MATER CKG SHITMLtl ACH IEVEG IHITIA'tE CLOSURE OF ALL CONTAINMENT ISGLAT IOH LEVEL IN REACTOR VAI.VES EXCEP'I MAINS'TEAN LINES VESSEL OH HIGH COHTAIHHEHT PRESSURE CORE RESIDUAL HEAT REHOVAL
'PERTURE gARO FIGURE 2 ~ 9 CONTROL ROO OROP SAFETY SEQUENCE OIAGRAH ~q ovovo o zs - I~ t )t <<2 c; ~'" Q )p I I 'h +I ~ = q ~ ) Page 3-1 3.0 SCOPE OF ANALYSIS This analysis identified a minimum complement of safety-related electrical and instrumentation and control equipment in the harsh environment at MNP-2 required for safe shutdown and accident mitigation. Class lE (ClE) equipment in the Environmental gualification Program which was required for safe plant shutdown was evaluated. ClE equipment located in a mild environment was assumed to remain operable due to the lack of significant environmental stress on this equipment following the accident. The scope of the analysis is defined by four elements: 1) the accidents creating a harsh environment, 2) post-accident environmental conditions,
- 3) Regulatory Guide 1.97 requirements, and 4) non-safety equipment impact on safety-related equipment. The non-safety equipment evaluated is that associated with the safety-related equipment being reviewed.
3.1 Accidents Creatin a Harsh Environment The accidents considered in this analysis are those that potentially cause harsh environments that may adversely affect the functioning and/or integrity of safety-related electrical equipment. These accidents are Loss-of-Coolant Accidents (LOCAs) inside primary containment, High Energy Line Breaks (HELBs ) inside the reactor building, and the Control Rod Orop Accident. These accidents are defined in Section 5.2.2 3.2 Post-Accident Environmental Conditions The temperature, pressure, and radiation environments in which the equipment will be required to function are defined for: 0 Page 3-2
- 1. LOCAs inside the primary containment.
- 2. HELBs inside the reactor building.
- 3. The reactor building environment caused by LOCAs inside the primary containnent.
Post-accident environmental conditions specifically for the Control Rod Orop Accident (CROA) are not calculated. LOCA radiation profiles are used for equipment in the CROA preferred safe shutdown path that is the same as equipment needed to mitigate a LOCA (e.g. HPCS). Since the LOCA conditions are more severe than those resulting from a CROA, using LOCA radiation profiles to envelop both accidents provides a conservative treatment. I These considerations define the post-accident environments for which the analysis is performed. 3.3 Re ulator Guide 1.97 Re uirements Regulatory Guide 1.97 describes a method acceptable to the NRC for complying with the requirement to provide instrumentation to monitor plant variables and systems during and following an accident. This guide defines the minimum nuttier of variables to be monitored by the control room operating personnel under these conditions to perform their responsibilities in the evaluation, assessment, monitor ing and execution of control room functions. Sufficient variables are also defined for control room operating personnel to perform their role in the emergency plan when the other eaergency response facilities are not sufficiently manned. The application of the criteria for the instrumentation is limited to that part of the instrumentation system and its vital supporting features or power sources that provide the direct display of the variables. Page 3-3 The instrumentation for WNP-2 to satisfy Regulatory Guide 1.97 requirements has been identified and submitted to the NRC in Oecember, 1982. In the context of the BWR variables listed in Table 1 of Regulatory Guide 1.97, the WNP-2 instrumentation include:
- l. All Category 1 items on Table l.
- 2. The following Category 2 items on Table l.
A. Radiation exposure monitors in the reactor building for sources within the primary contaianent. B. All Type 0, Category 2 variables.
- 3. No Type E variables (other than those which are Category 1).
As with the Environmental gualification Program jn general, it is not expected that all of the instrumentation described in the Regulatory Guide 1.97 submittal will have complete qualification documentation prior to fuel load. Some instrumentation, upgraded to more stringent qualification requirements, are still being procured. For those instrument types not shown to meet the environmental qualification requirements, component-specific justifications have been performed. Many of the instrument types in the WNP-2 Class lE list for accident monitoring per Regulatory Guide 1.97 have different use codes for different required operating. times. For example, a HPCS flow transmitter is required to be operable when HPCS is required, but is only required to maintain system integrity when HPCS is not required. The HPCS system is required to be operable for 24 hours following certain accidents. Therefore, the HPCS flow transmitter is required to be operable for 24 hours and then maintain system integrity for the remainder of the time. gualification of this type of instrument would include operability for 24 hours and integrity for 4320 hours (6 months). jlj t y 1 ~ ' Page 3-4 Other instruments, although required to be operable per Regulatory Guide 'l.97 for the duration of accidents, provide indication of the completion of its related safety function within 10 minutes after an accident. The component-specific justification for interim operation of this type of instrument indicates that the instrument would be qualified, in the short term, to provide its required short>>term, active safety function. Diverse, qualified instrumentation or qualified safety function components are then available to provide long-term, indirect monitoring or assurance of maintaining the safety function. To pr event misleading the operator with unreliable information from unqualifed instrumentation, WNP-2 has established a program to install an identification scheme for control room instrumentation and control. The identification scheme will identify to the operators'whi ch indications and controls are qualified . 3.4 Impact of Non-Safet f ui ent on Safet -Related E ui ment Assessments of the Three Nile Island-2 incident and other recent events, such as those at Browns Ferry-3 and Crystal River-3, have identified the need for reducing unexpected reactor incidents caused by hidden system dependencies. These hidden dependencies, also known as Systems Interactions (SIs), have often resulted from non-safety equipment impact on safety-related equipment. To prevent adverse SIs from occurring, the NRC has established an on-going program to define and subsequently implement SI regulatory requirements for light water reactors. Page 3-5 WNP-2 has assessed the non-ClE components within ClE systems. This non-safety equipment impact on safety-related equipment is evaluated in terms of functional SIs. These are interactions resulting from either the sharing of components between systems or through physical connections between systems such as electrical,'hydraulic, pneumatic and mechanical. These SIs were evaluated as part of the WNP-2 Electrical Separation Program and also in the Failure Nodes and Effects Analysis task of the JIO effort. The WNP-2 Electrical Separation Program has assessed the impact of electrical non-C1E equipment on ClE busses. The non-ClE loads identified were then addressed in two ways:
- 1. The non-ClE loads were treated as prime or associated circuits; or
- 2. Class lE isolation devices (e.g., fuses, circuit breakers, etc.) have been installed.
Functional systems interactions were also evaluated by a Failure Nodes and Effects Analysis (FNEA) in the JIO analysis. The FNEA identified the correct Use Code (2 or 3) for equipment on the WNP-2 ClE list which did not have an active safety function. Use Code 2 describes equipment which need not perform an active function for mitigation of a design basis accident but must'not fail in a manner detr imental to safe shutdown. Use Code 3 describes equipment which need not function for accident mitigation and whose failure is deemed not detrimental to plant safety. Section 5.2.4 provides more details on the FNEA. Thus, the non-safety equipment impact on safety-related equipment has been evaluated within the boundaries of the JIO analysis. t ll Page 4-1 4.0 SAFETY -RELATED SYSTEMS 4.1 A~roach The identification of equipment for environmental qualification is consistent with the project design documents, including flow diagrams, electrical one-line diagrams, logic diagrams, elementary diagrams, and instrument loop diagrams. Examination of these documents entails the following:
- 1. Review of electrical one-line diagrams to identify electrical equipment which distributes safety-related power to safety-related electrical systems.
- 2. Review of system flow diagrams, related system descriptions, and FSAR system function sections (including Chapter 15
- accident analysis) to A. Identify the electrical equipment within each system, B. Analyze and tabulate emergency events for each equipment. item and the corresponding safety function required to mitigate the emergency event, C. Define operating use codes, and D.'etermine the amount of time required for completion of the safety function.
- 3. Review of elementary and loop diagrams for each Class lE electrical component to identify safety-related equipment required to control or monitor the equipment.
Equipment identified in this manner is listed on the Class lE Equi pment Li st (Appendi x A) . Page 4-2 4.2 ~5R 4" 4 Six safety functions were identified in performing this analysis. The following definitions of safety functions were adopted: 4~f4" 4: Rf 2 RR 24 2 4 maintain the nuclear plant in a safe and stable condition. Safety functions noted below will assure the safe shutdown of WNP-2. The six safety functions are:
- l. Emergency Reactor Shutdown Reactivity Control >> establish and maintain core reactivity to assure reactor shutdown.
- 2. Primar Containment Isolation Seal all potential paths out of primary containment following an accident in order to prevent a significant release of radioactive materials.
- 3. Reactor Core Coolin A. Initial Core Cooling - provide core heat removal imoediately following an accident to prevent damage to the fuel. This function involves the injection of emergency cooling water by various systems.
B. RCS Pressure Control - maintain reactor coolant system pressure in accordance with the thermal-hydraulic limits for any given reactor operating mode. C. RCS Level Control - maintain reactor vessel water level to ensure adequate cor e cooling. 4 4 2 ~ 4 Page 4-3
- 4. Containment Inte rit A. Primary Containment Hydrogen Control - maintain the containment H2 and 02 concentration within acceptable limits in order to prevent the possibility of the uncontrolled burning of hydrogen.
B. Primary Containment Pressure and Temperature Control-maintain the containment environment within acceptable limits in order to prevent damage to the containment structure and its contents'.
- 5. Core Residual Heat Removal Long-Term Core Cooling - provide core heat removal for extended periods of time via recirculation of reactor coolant.
- 6. Prevention of Si nificant Release of Radioactive Material to the Environment Reactor Building Isolation - seal and/or control potential release paths out of 'the reactor building in order to prevent the release of radioactive materials to the environment in excess of 10CFR100 limits.
The systems required to accomplish the six safety functions are presented below. Only a portion of each system may be needed to support a particular safety function. The equipment in all required systems and subsystems is included in the Class lE Equipment List (Appendix A). Instrumentation required by Regulatory Guide 1.97 to follow the course of an accident has been identified with respect to a particular system and is also included in the Class lE Equipment List. 0 ~ ' Page 4-4 The following systems support the six safety functions through the operation of individual components or operation of the system as requir ed. ~Sstem Abbreviation Containment Atmospher e Control CAC Control Air System CAS Containment Purge Exhaust CEP Containment Instrument Air CIA Containment Monitoring System CMS Containment Return Air CRA Control Rod Drive CRO Containment Purge Supply CSP Containment Vacuum Breaker CVB Electrical Distribution E Equipment Drains Radioactive EDR Floor Drains Radioactive FOR Fuel Pool Cooling FPC High Pressure Core Spray HPCS RRC Hydraulic Control HY Intermediate Range Monitors IRM Leak Detection LD Low Pressure Core Spray LPCS Local Power Range Monitor LPRM Main Steam MS Main Steam Isolation Valve Leakage Control MSLC Process Instrumentation Process Sampling Radioactive PSR Reactor Building Closed Cooling RCC Reactor Core Isolation Cooling RCI C Reactor Building Exhaust Air (HVAC) REA Reactor Feed water RFM Residual Heat Rermval (Includes Containment Spray) RN Page 4-5 ~sssem Abbreviation Reactor Building Outside Air (HVAC) ROA Reactor Protection System RPS Reactor Building Return Air (HVAC) RRA Reactor Recirculation RRC Reactor Mater Cleanup RMQl Standby Gas Treatment SGT Standby Liquid Control SLC Suppression Pool Temperature Monitoring SPAM Source Range Monitor SRN Standby Service Mater SM Traversing In-core Probe TIP Page 4-6 4.3 Correlation Hetween FSN Table 3.2-1 and Systems Reviewed The safety-related systems listed in the WNP-2 FSAR Table 3.2-1 were addressed. A detailed correlation between the systems listed in Section 4.2 and FSN Table 3.2-1 is given below. FSN Table 3.2-1 JIO S stem Review
- 1. Reactor System No electrical equipment to consider
- 2. Nuclear 8oiler System MS system r eview
- 3. Reactor Recirculation System RRC system review
- 4. QO Hydraulic System CRO system revie~
- 5. Standby Liquid Control System SLC system review
- 6. Neutron Monitoring System SRM, IRM ANO LPRM system reviews
- 7. Reactor Protection RPS system review
- 8. Leak Oetection System RCIC, RWQJ, MS, RHR and LO system reviews
- 9. Process Radiation Monitors MS and REA system reviews
- 10. RN System RHR system review ll. Low Pressure Core Spray LPCS system review
- 12. High Pressure Core Spray HPCS system review
- 13. RCIC System RCIC system review
- 14. Fuel Service Equipment No equipment required for accident mitigation
- 15. Reactor Vessel Service No equipment required for Equipment accident mitigation
- 16. In-Vessel Service Equipment No equipment required for accident mitigation
Page 4-7 FSAR Table 3.2-1 JIO S stem Review
- 17. Refuel ing Equipment No equipment required for accident mitigation
- 18. Storage Equi pment No equipment required for accident mitigation
- 19. Radwaste System EDR and FDR systems review
- 20. Reactor Water Cleanup System RWCU system review
- 21. Fuel Pool Cooling and Cleanup FPC, RHR and SW system System reviews
- 22. Control Room Panels Outside of harsh environment
- 23. Local Panels and Racks Equipment included with other system reviews
- 24. Off-Gas System No equipment required for accident mitigation
- 25. Standby Service Water System SW system review
- 26. Turbine Plant Service Mater No equipment required for accident mitigation
- 27. Reactor 8uilding Closed Cooling RCC system review Mater System
- 28. Primary Containment Cooling CEP, CSP, (RA, RCC system System review
- 29. Standby Gas Treatment System SGT system review
- 30. Primary Containment Atmospheric CAC system review Control System
- 31. Other HVAC REA, ROA and RRA system reviews
- 32. Condensate Storage and Transfer No equipment required for accident mitigation
- 33. Instrument and Sample Lines PI and PSR system reviews
- 34. Fuel Storage Faci1 ities No equipment required for accident mitigation
Page 4-8 FSN Table 3.2-1 JIO S stem Review
- 35. Building Cranes No equipment required for accident mitigation
- 36. Instrument and Service Air CAS system review
- 37. Containment Instruaent Air CIA system review
- 38. Diesel Generator Systems Outside of harsh environment
- 39. Standby AC Power Systems E system review (other systems outside of harsh environment)
- 40. Auxiliary 125/250 Volt DC E system review Power Systems
- 41. 24 Volt DC Power System E system review
- 42. 120 Volt Critical Power Supply E system review
- 43. Power Conversion System No equipment required for accident mitigation
- 44. Cir culating Mater and Cooling No equipment required for Tower Makeup Mater System(s) accident miti gation
- 45. Main Steam Isolation Valves MSLC system review Leakage Control System
- 46. Containnant Vessel Passive - structural, no electrical equipment to consider
- 47. Huildings Structural, no electrical equipment to consider
- 48. Containment/Drywell Atmosphere CMS system review Monitoring System
- 49. Drywell Insulation No electrical equipment to consider
- 50. Instrumentation and Control Addressed in various system Equi pment reviews
Page 5-1 5.0 METHODOLOGY The JIO analysis was performed in three phases: 1) Safe shutdown analysis, 2) Determination of a minimum set of equipment, and
- 3) Development of component-specific justifications.
5.1 j~roach The approach used to justify interim operation was designed to identify all equipment essential to achieve and maintain safe shutdown following an accident. Extensive accident analyses provided the basis for selecting an optimum shutdown path to accomplish the six safety functions. Based on this preferred safe shutdown path, a minimum set of equipment which must be qualified prior to fuel load was then identified. The safety-related electrical equipment in the Primary Containment and Reactor Building was used as a basis for equipment selection. This equipment includes all components essential to accomplish the six safety functions. Instrumentation required by Regulatory Guide 1.97 to monitor the course of an accident was also included. In general, Use Code 1, 2, or 3 components which could be potentially exposed to a harsh environment were considered in this analysis. However, there is one "component" type which does not appear in Tables A and B but does appear in the ClE Equipment List: composites. Composites provide redundant information, because the safe shutdown ClE equipment on composites are individually identified for environmental qualification. The composite does not require additional environmental qualification documentation. Page 5-2 Accident definition narrowed the list of equipment considered to that potentially exposed to a harsh environment; that is, equipment inside the Primary Containment and Reactor Building. A Safety Sequence Analysis (SSA) and Failure Modes and Effects Analysis (FMEA) further reduced the list to those components required for LOCA and/or HELB Mitigation. The final reduction, selection of minimum required equipment for the preferred safe shutdown path for each safety function, was completed in two steps. The first step consisted of checking the status of the equipment qualification effort at HNP-2 to identify those components with incomplete qualification documentation. In the second step, the results of the SSA were reviewed to determine a single shutdown path that consists of the minimum number of additional components that require qualification documentation prior to fuel load. Lastly, the components on the preferred safe shutdown paths were identified as Table A equipment, and components on the alternate paths were listed as Table 8 equipment. For those Table A components whose qualification documentation were incomplete, component-specific justifications were developed. These justifications were developed in accordance to the criteria set forth in the final NRC rule 10CFR50.49 (i). Page 5-3 5.2 Safe Shutdown Anal sis The Safe Shutdown Analysis involved defining the harsh environment producing accidents, performing a Safety Sequence Analysis, and conducting Failure Modes and Effects Analyses. 5.5 1 The following assumptions were made during the performance of the Safe Shutdown Analysis:
- 1. For shutdown analysis, no credit was taken for non-saf ety-re ated el ectr ical equipment.
1
- 2. A Design Basis Earthquake (OBE) can occur, but not simultaneously with the Design Basis Accident (OBA).
- 3. Only one accident at a time is postulated to occur.
- 4. Offsite power is lost at the time of the accidents, except for the auxiliary steamline break accidents 0 and M.
- 5. Containment radioactive leakage within design limits will occur+
- 6. Accidents occurring inside the reactor building have no effect upon environmental conditions inside the primary contairment.
- 7. HPCS is used to mitigate a Control Rod Drop Accident in the event RCIC does not function adequately.
5 ~ Page 5-4 5.2.2 Accident Definition The primary containment and most areas of the reactor ouilding (with the exception of specially designed electrical equipment rooms) can be exposed to a harsh environment from postulated LOCAs/HELBs. This task identified the line breaks potentially causing a harsh environment and the associated environmental conditions. From the environmental service conditions of Appendix B, a total of 18 line breaks in six systems were identified and categorized into seven events, including three LOCAs and four HELBs. The areas of the reactor building affected by each break were identified and tabulated. The environmental conditions associated with each break wer e also defined, including the calculated post-accident radiation, temperature, pressure, and humidity levels expected. The Control Rod Drop Accident (CRDA) was also assessed. Specific environmental conditions associated with this accident were not determined. For the equipment used to mitigate this accident that is the same as the equipment required for LOCA mitigation, LOCA radiation conditions were used to envelop the CRDA conditions. This approach is conservative since the LOCA environmental conditions will be more severe than those resulting from a CRDA. The eight postulated accident types are included in Table 2.1, and the areas of the plant affected by each postulated break location are identified in Table 2.2. Page 5-5 5.2.3 Safet Se uence Analysis A Safety Sequence Analysis was performed to identify equipment required for safe shutdoWn following any of the accidents considered. The cooeined results of this analysis and the FiilEA (see Section 5. 2. 4) reduce the set of equipment requiring qualification documentation prior to fuel load to that equipment essential for the mitigation of the postulated accidents, as shown in Figure 5.1. During the first stage of this analysis, Safety Function Path Diagrams (SFPOs) were developed for each safety system. These diagrams identified all auxiliary support systems associated with each safety system. The final stage involved assembling the appropr iate portions of each SFPD into Safety Sequence Diagrams (SSOs) for each accident. The SSOs for the accidents have been presented in Section 2.0. 5.2. 3. 1 Safet Function Path Dig rams Based on a review of the plant design, and the accident descriptions in Chapter 15 of the FSN, systems were selected that could achieve, or help to achieve, a given safety function. System descriptions, flow diagrams, and logic diagrams were then reviewed to determine system operation and to identify major components and their role in the completion of the safety function. When all the design paths that achieve the safety function were identified, a Safety Function Path Diagram (SFPO) was developed. Page 5-6 This diagram: 1) flowcharts all possible methods of achieving the safety functions, 2) depicts each required safety system's response to the accident, and 3) shows chronological and functional relationships, initiating input variables, and required operator actions. The entire accident duration is represented, including the activities necessary to achieve cold shutdown. A sample SFPO is provided as Figure 5.2. Safety Function Equipment Lists were developed upon completion of each diagram. The oasis for the list is the MNP-2 Class lE Equipment List (Appendix A). For each safety system on the SFPD, the corresponding set of equipment from the WNP-2 Class lE Equipment List is included in the Safety Function Equipment List. This assures that all equipment needed for the operation of the safety system and completion of the safety function is considered. 5.2.3.2 Safet S stem Auxiliar Diagrams Safety System Auxiliary Diagrams (SSAOs) were developed for each system identified on a Safety Function Path Oiagr am. The purpose of these diagrams is to identify all auxiliary systems that are necessary to support a given safety system. Specific equipment in those auxiliary systems that are required to operate to provide that support are also identified. Page 5-7 Prior to SSAO development, the references documenting the operation of the chosen safety system (PEIOs, FCDs, FSAR sections) were reviewed and all auxiliary systems which support the safety system identified. A block diagram was subsequently developed that presented the safety system support requirements. It included the support systems, the presence of r edundant trains, outputs of the support systems, initiating signals and trip conditions, and any operator actions. A sample SSAO is pr ovided as Figure 5.3. After completion of the SSAO, an Auxiliary Equipment List was prepared for the safety system. The basis for this list is the MNP-2 Class lE Equipment List (Appendix A). For each auxiliary system on the SSAOs, the corresponding set of equipment from this list is included in the Auxiliary Equipment List . This assures that all equipment required to support the operation of the safety system and completion of the safety function is considered. 5.2.3.3 Safet Se uence Oia rams Safety Sequence Diagrams (SSOs) were developed for each of the accidents postulated (see Section 5.2.2). First, an accident description was developed for each accident that includes a di scuss ion of the pl ant 's post-acci dent stable condition. Then the plant initial conditions were defined, and the portions of each generic SFPD page 5-8 applicable to the accident were assembled to form the Safety Sequence Diagram. Each path was modified to ref 1 ect acci dent-speci fi c parameters, actions, and inputs. The final SSD is a flowchart representation of the plant's response to the postulated accident via the operation of essential safety systems. A sample SSO is provided as Figure 5.4. 5.2.4 Failure Modes and Effects Analysis A Failure Nodes and Effects Analysis (FNEA) investigated the propagation of the consequences of a single component failure on its composite equipment, its system, and the safety function for which the system is required. Performed in conjunction with the SSA tasks, this analysis helped define the minimum set of equipment essential to safety system operation, thereby assuring safety function completion {see Figure 5.1). The FNEA was performed for each of the components on the Safety Function Equipment Lists or on the Auxiliary Equipment Lists that was not required to function to achieve the six safety functions (i.e., those components with accident Use Codes 2 or 3). Failure modes were then postulated for the equipment. Failure modes were postulated conservatively, without consideration of the scenario resulting in the failure. The effect of each credible failure on the equipment, its system, its safety system {for auxiliary equipment), and the safety function was assessed. Based on this assessment, the failure of some equipment was determined not to be detrimental to Page 5>>9 plant safety or accident mitigation and, therefore, need not be qualified for the accident environment. Equipment which can fail in a manner detrimental to plant safety must be qualified. Based on the effects of the failure, the appropriate use code was assigned to the equipment. When the use code resulting from the FMEA did not concur with the initial accident use codes, changes to the use codes in the C1E list were initiated. An example of the FMEA for three of the components considered is included in Figure 5.5. 5.2.5 Use Code Definition The Safe Shutdown Analysis verified equipment use as specified in the Class lE list (Appendix A.) Equipment use during accident conditions were described by five "Use Codes": 0, 1, 2, 3, and 4. The definitions of these codes for equipment use during a Design Basis Accident are provided below: Use Code Definition Equipment that will not experience the environmental conditions of design basis accidents and is not r equired before, during, or after an accident. Page 5-10 Equipment that will experience the environmental conditions of a Oesign Basis Accident for which it must function to miti gate said accident, and that will be qualified to demonstrate operability in the accident environment for the time required for accident mi ti gati on with a safety margin to fail ur e. Equipment that will experience the environmental conditions of a Oesign Basis Accident through which it need not provide an active function for mitigation of said accident, but through which it must not fail in a manner detrimental to plant safety or accident mitigation, and that will be qualified to demonstrate the capaoility to withstand any accident environment for the time during which it must not fail with a safety margin to failure. Equipment that wil experience environmental 1 conditions of a Oesign Basis Accident through which it need not function for mitigation of said accident, and whose failure (in any mode) is deemed not detrimental to plant safety or accident mitigation, and need not be qualified for any accident environment, but will be qualified for its non-accident service environment. Page 5-ll Safety-related equipment that will not experience environmental conditions of a Design Basis Accident for which it must function to mitigate said accident and that will be qualified to demonstrate operability under the expected extremes of its accident service environment. This equipment would be located outside the Reactor Building. 5.3 Minimum Re uired Set of Equipment In defining the minimum set of equipment requiring qualification documentation prior to fuel load, the current qualification status for the essential equipment, as identified during the SSA and FMEA, was reviewed. Items already documented as qualified were determined. The remaining items then comprised a preliminary set of essential components requiring additional qualification effor t (see Figure 5.6). This set of safety-related electrical equipment was then evaluated to determine a minimum set of equipment to be qualified prior to fuel load. 5.1.1 Two assumptions were made in selecting the minimum set of equipment needed for safe shutdown.
- l. Accomplishing the six identified safety functions will ensure plant safe shutdown.
A. Emergency Reactor Shutdown Page 5-12 B. Primary Containment Isolation C. Reactor Core Cooling O. Containment Integrity E. Core Residual Heat Removal F. Prevention of Significant Release of Radioactive Material to the Environment
- 2. Single failures (active or passive) were not assumed.
5.3.2 Preferred Safe Shutdown Paths The SSOs shown in Figures 2.1 - 2.9 were reviewed and a single path to accomplish each safety function for each accident type was chosen. The path with the least number of components yet to be documented as qualified was chosen where possible. Train A components were generally selected in order to assure that most items were powered from the same electrical division. Figure 5.7 shows the preferred path for each of the identified safety functions. This preferred path is the composite of the selected path for accomplishing the safety function for each accident type. The following systems in the preferred path support each safety function through the operation of individual components or operation of the system as required. Page 5-13 Emer ency Reactor Shutdown Reactor Protection System Control Rod Drive System Containment Isolation 'rimary RRC Hydraulic Control Hain Steam System Reactor Feedwater System Reactor Recirculation System High Pressure Core Spray System Low Pressure Core Spray System Standby Liquid Control System Residual Heat Removal System Reactor Core Isolation Cooling System Containment Atmosphere Control Containoent Supply Purge System Reactor Closed Cooling System Reactor Water Cleanup System Equipment Drains Radioactive System Floor Drains Radioactive System Containment Instrument Air System Process Instrumentation System Control Air System Fuel Pool Cooling System Traversing In-Core Probe System Page 5-14 Reactor Core Coolin High Pressure Core Spray System Low Pressure Core Spray System Main Steam System - Automatic Oepressurization System Residual Heat Removal System Containment Instr unent Air System (Support System) Standby Service Mater System (Support System) Containoent Inte it Containment Atmosphere Control System Containment Monitoring System Containment Return Air System Containoant Vacuum Breaker System Residual Heat Removal System Standby Service Mater System (Support System) Suppression Pool Temperature Monitoring Core Residual Heat Reooval Automatic Oepressurization System Residual Heat Removal System Standby Service Mater System (Support System) Prevention of Release of Radioactive Material to the Environment Standby Gas Treatment System Main Steam Leakage Control System Standby Service Water System (Support System) >} Page 5-15 Leak Detection System (Input) Reactor Building Exhaust Air System Reactor Building Outside Air System The alternate paths for accomplishing each of the six safety functions are shown in Figure 5.8. Three basic types of systems do not directly perform a safety function and are not shown in the SSDs in Figures 5.7 and 5.8:
- l. Auxiliary Systems which provide support to safety systems A. Service Water (SW) System B. Reactor Building Return Air (RRA) System
- 2. Systems which provide inputs to safety systems A. Electrical {E) System B. Leakage Detection (LO) System
- 3. Instrumentation to monitor the course of'he accident (per Reg. Guide 1.97)
A. Process Sampling Radioactive (PSR) System However, all of these equipment have been evaluated as part of the Safety Sequence Analysis. Page 5-16 The sum of the preferred paths for each safety function (six) comprise the preferred safe shutdown path for JIO analysis. 5.3.3 Selection of Ninimum Set of E ui ment The equipment required to assure the operation of each safety system in the chosen path represents the minimum set of required equipment to be provided with qualification documentation prior to fuel load (see Figure 5e9).. Table A identifies the equipment in the preferred safe shutdown path that will have qualification documentation or component-specifi c justifications prior to fuel load. Table B identifies the remaining safety-related electrical equipment in the alternate shutdown paths to have qualification documentation prior to the end of the first refueling outage. 5.3.4 Results The results of this analysis identify the necessary and sufficient equipment to ensure safe shutdown for HHP-2. All viable safe shutdown paths which can accomplish the required safety functions under accident conditions have been determined. A single preferred path is then selected such that the equipment on that path comprise the minimum set requiring documentation of qualification or justification prior to fuel load. gualification documentation or component-specific justifications ar e completed for the preferred safe shutdown path equipment. Table A identifies this minimum set of equipment. The remaining safety-related Page 5-17 electrical equipment in the alternate shutdown paths are listed in Table B and will have complete qualification documentation prior to the end of the first refueling outage. Interim operation is justified since a fully qualified or justified preferred safe shutdown path has been i denti fed for each saf ety function. An accident legend is included as the first page of each table. 5.4 Com onent-S ecific Justification 5.4.1 ~roach A small number of safety-related equipment in the preferred safe shutdown paths which have incomplete qualification documentation have been identified. To meet the current qualification requirements in 10CFR50.49(i), an analysis of this equipment has been completed to ensure that the plant can be safely operated pending compIetion of the envir onmental qual ification program. This analysis includes, where appropriate, consideration of:
- 1. Accomplishing the safety function by some designated alternative equipment if the principal equipment has not been demonstrated to be fully qualified.
- 2. The validity of partial test data in support of the original qualification.
Page 5-18
- 3. Limited use of administrative controls over equipment that has not been dermnstrated to be fully qualified.
- 4. Completion of the safety function prior to exposure to the accident environment resulting fr om a design basis event and ensuring that the subsequent failure of the equipment does not degrade any safety function or mislead the oper ator.
- 5. No significant degradation of any safety function or misleading information to the operator as a result of failure of equipment under the accident environment resulting from a design basis event.
Criteria, as presented below, were established to implement these areas in the development of the individual equipment justi ficati ons. 5.4.2 Criteria Criteria, with due consideration to the 10CFR50.49(i) items listed above, were established to minimize the impact on operation of the plant during accident conditions. This approach resulted in a prioritized set of criteria for justifying plant operation with equipment that has incomplete qualification documentation. Three criteria were utilized, with the highest priority placed on the first one. The other two criteria were applied only when the first criteria could not be used. Page 5-19 These three criteria, listed in descending priority, used in the development of the component-specific justifications include:
- l. Accomplishing the safety function by some designated alternate equipment. The alternate equipment is environmentally qualified to the accident for which it must operate to perform the safety function.
- 2. Completion of the safety function prior to exposure to the ensuing accident environment. The time required for completion of the safety function may result in subjecting the equipment to the harsh accident environment. If completion of the safety function can be verified before the equipment exposure exceeds the mild environment radiation limit (104 Rads), then the equipment is operable for that time period. This criterion is used only when radiation is the restr icting par ameter for qualification.
- 3. Administrative controls over equipment that has not been demonstrated to be fully qualified. In order to minimize the impact on operations, this criteria is applied only in cases where specific alternative operator action is required to achieve safety function.
In addition to these three major criteria, the component-specific justifications also considered the degradation of any safety function or providing misleading information to Page 5-20 the operator as a result of equipment failure. Failure analyses have demonstrated that the consequences of the electrical failure of most of the equipment could be isolated by Class lE fuses and/or circuit breakers or does not degrade the function of the safety. If the electrical failure could provide misleading information to the operator, then the equipment was identified to be included in a control room instruoent and control identification scheme. This identification scheme will prevent the operator s from being misled by potentially unreliable information. 5.4.3 Results The component-specific justifications for equipment with incomplete qualification documentation were developed based on the constraints imposed by the component or environment. The justification ar guments were primarily systems oriented. A failure consequence analysis is discussed regardless of the 'evidence provided in the justification that such a failure of the equipment is not anticipated. In some cases, the required radiation dose may be less than the zone dose, depending on operability time and component location within the zone. If this is so, details are provided within the body of the justification. Component-specific justifications developed in this analysis are included in this appendix. Each justification presents the following information: Page 5-21
- 1. I COMPONENT IDENT IF CAT ION EPN - Equipment Piece Nuober Description of equipment Component Type Manufacturer/Model
- 2. ACCIDENT CONDITIONS Accident Profile - LOCA or HELB accident conditions under which equipment must function to mitigate the accident Use Code Oper ability Time Radiation Zone Zone Dose - The sum of the accident dose (180 days) for the accident environment plus normal operation dose (40 years) for the worst target in zone.
- 3. DESCRIPTION OF CG4)PONENT SAFETY FUNCTION Description of the safety function provided by the component, including an operational description of the component within its system. If applicable, details of the component's mechanical/electrical operation are included.
- 4. AL IF I CAT ION STATUS Surreary of qualification status, including parameters requiring justification.
Page 5-22
- 5. JUSTIFICATION FOR INTERIM OPERATION Component-specific justification developed for LOCA and based on the criteria presented in Section 5.4.2. 'ELB
- 6. GONCLUS ION Main points of the justification statement.
Page 5-23 FlGURE S.'i Safety-Related Electrical Eciuipment Exposed to a Harsh Environment W '0 I C Safety-Related Electrical Equipment te 0l Q Equipment required 0 for LOCA mitigation a ~ ip LOCA 8 e CL W HELB 4l 0 ~ aa 0 Ci ~ ia C Equipment reaulred 4e for HELB mitlgetlon C 0 o CL 4 4a hl ~ ' Page 5-24 INITIATING ACCIDENT INDICATES THE SAFETY SYSTEMS INDICATES ALTERNATIVE SYSTEMS THAT ARE REQUIRED TO FUNCTION THAT ARE NOT REQUIRED TO IN ORDER TO ATTAIN A SAFETY ATTAIN A SAFETY FUNCTIONS BUT FUNCTION ARE EXPECTED TO OPERATE SAFETY SYSTEN SAFETY SYSTEYi A B NDICATES SYSTEMS HAVING TWO INDEPENDENT "TRAINS"e A AND B SAFETY SYSTEM (( T (FSAR SECTION) THE SAFETY SYSTEM) EITHER SENSED t VARIABLE TH IDENTIFIES SPECIFIC FSAR SECTION OR T WILL IN-ITIAfE SYSTEM S THAT DESCRIBES THE FUNCTION OF THE SAFETY SYSTEM SAFE'N SYSTEN PP CH ENSED (FSAR SECTION) VARIABLE TH AND P MUST EXCEES THEIR LIMITS TO INITIATE SYSTEM T SAFETY FUNCTION A DESIRED SAFETY FUNCTION ACHIEVEDI STABLE CONDITION l1AI NTA INED, a FIGVRE 6 2 Sample Safety Function Path Diagram SSAD No. (SSAD No.) AUXILIARY AUXILIARY SAFETY SYSTEM Supply Cooling SAFETY SYSTEM Supply Cooling Water for Pump X A Air for System K A B in System K A B Pump Motor (SAR Reference) (SAR Reference) AUXILIARY SAFETY SYSTEM SAFETY SYSTEM Supply Makeup K Water to Tank A B (SAR Reference) Indicates system (s) whose failure does not affect operation of the safety system SSAD No. (SSAD No. AUXILIARY AUXILIARY SAFETY SYSTEM Breaker Control SAFETY SYSTEM 4.16 KV SWGR Power for for System K Pumps A B System K A B (SAR Reference) (SAR Reference) SYSTEM DESCRIPTION 0 Page 5-26 CAa Accident Title Different Plant Condition Different Plant Condition Safety System Safety System Safety System Safety System j R Sianal II II A B A 9 C D AB B I 2/4 Safety System U Generates Signal "AB" which initiates system N Safety System Safety System Safety System A B SySTEM OPERATION IS AUTOMATIC Safety System Safety Safety System Safety System A T A J A Function A B U System K is Passive Safety Manual Action Safety System Function Required for A System Y B Safety Safety Function Function C ElGURE 5.4 Sample Safety Sequence Diagram f SPPD Sam le SSAD Sam le EQUIPMENT INITIAL MORST CASE FAILURE EFFECTS ON LOCAL 6 ASSOCIATED SySTEMS FAILURE PART NO USE MODE (INCLUDING COMPENSATING FEATURES) USE CODE CODE RRC-MO-2 3 A 2 Motor operator fails No effect - valve position is to operate valve irrelevent and has no post-accident c function 0 CL FDR-POS-V/4 2 Limit switch fails Must operate - provides operator Q with false indication of containment in false position Q isolation CL Ill Isa SGT- TS-1A11 2 Hot-short or open or No effect - switch controls heaters the integrity fails that are not required for safety n and the switch body function. Does not deterimentally Q leaks effect SGT function if heaters are on or off. Leakage would be from the Reactor Building into SGT, Q which would not adversely affect operation since SGT normally draws EO air from the Reactor Building. Q Page 5-28 FIGURE 5.6 Unqualified Safety-Related Electrical Equipment Exposed to a Harsh Environment 44 I Safety"Relatecf Eiectrica/ Equipment C w 0la Equipment required 0 for LOCA mitigation 0 0 LOCA e 0' HELB 4I C '0 w W 0 0 0 Equipment required ~0 for HFLB mltlgrtlon Ovrllfled Equipment O I 0 Hon-Ourllfled Equipment CL w 0' 4l W w
- 4 PAGE 5-2S INITIATING EVENT PR IHARY REACTOR RC IC CONTA IHtKHT PASSIVE PASSIVE VCLAP%
SUILOING CST LEVEL UXt OR QPPRESS I OH POOL LEVEL HIGH 2 Pl SLPP)tESSI OH PASS IVE POOL EHERGENCY HY REAC'TOR CST ORY))ELL SHL) 'TOO)tN PRE SSLNE YES LESS THAN NETHELL 2 I JECHAHICAl CVB OH.Y REACTIXt RHt tDCCK VALYES) M) HSIV'S IXPRESSLR I ZG) QRPRESS I OH lEAK 2 PIXL CtXLING HIGH PRES SHJTQQA CIXLING ECCS NXQJATE AVAILARX 2 l5R AOSlt)S) S)A)TOO4l CIA CIXLING PREVENT ION CF CAC SIGo RELEASE CF R4t RAOe HAT'L TO HIGH AlTo LOOLING EHV I RCI4%NT RCS PRESSURE AOSIHS) CORE RES IOUAL HEAT REHOVAL %&PRESS)OH YES POOL COIX.INO TIP LPCI P R C '"'~ REACTIXt PR I NARY COH TA INtKNT ISOLATION CORE CIXLItQ CONTA I ftKNT INTEGRITY ,'.APERTURE
- k. CARD A4O AVeQILblq OII
.>Rperture Caga FIGURE 5+7 PREFERREO SAFE SHUTOOHN PATHS (TABLE A) SAFETY SEQUENCE OIAGRAM .8@OVOV OO)S - 0'i I, ~ < 4~ a, e e* ~ 'l pWb W ~,'* g%~ CG PAGE 5-30 IttlT I A TING EVENT PR I NARY REAC tOR CQITAINHENT PASSI YE PASSIVE GUILOIHG CST LEVEL LQi CR YOUJC QPPRESSICN PM LEVEL HIGH 2 SLPPRESSION PASSIVE POQ EHERGEHCY REACTOR CST ORYNELL ShUTOOIN PRE SQRE CAC LESS 'tHAN &TNELL 2 MECHANICAL FOR CVB ONLY REACTOR QR I QCCK VALVES) Kl HSIY'S RC I C CAS OEPRESQR I ZED QPPAESS POQ. I (W COCL IHG LEAK 2 CIA SHUTQXA HIGH PRES CXXLING ECCS ACXCUATE AVAIlARZ 2 SGTS 2 RHt AI6(HS) SHUTOCLH COOLING PREYENTIOtt CF SIGo RELEASE OF LM HIGH flft RADAR NATL TO ENV IRLRPKNT ALTa COOLING TIP RCS PRESSURE HY CORE RESIOUAL ADSI HS) hEAT REHOVAL PR I NARY CONTAINHENT SLPPftESS I ON ma. I SOLAT ION ~O2 COOLIE RCS PRESSURE 100 PSIG 2 PRC CONTA I ttKNT P,t ERVuRE INTEGRITY CARD REACTOR ~0 AYaQttbjL OII COIE GLLIM3 'kpeztare Cara FIGURE 5 8 ALTERNATE SAFE SHUTOOHN PATHS (TABLE B) SAFETY SEQUENCE OIAGRAM SSOVOVOOiS -92 V le Ci C 1 Ly .* S w I. '( ~ Page 5-31 FlGURE 6.9 Safety-Related Electrical Equipment to Have Qualification Documentation Prior to Fuel Load or End of First Refueling 0 ~ ae Safety-Relatecf E/ectrical Equipment I I:0 C ~ Equipment regulred for LOCA mltlgatlon 0 ~0 C LOCA p HELB 4I @ac o e Osa O Q e 0'a Equipment required 4v for HELB mltlgatlon h Ouallfled Equipment 0 8 Equipment to be qualified prior to tuel loecl CL 4 Equipment to be qualified prior to end of 0'u ~ flrat refueling TABLE A ACCIDENT LEGEND A = HELB - RCIC H = HELB = RWCU 0 = HELB - RFW B "- HELB - RCIC I = HELB - RWCU P = LOCA - RRC C = HELB - RCIC J = HELB - RWCU Q = LOCA - MSL D~ HELB -AS K = HELB - RWCU R = LOCA - SMALL E = HELB - RCIC L ~ HELB - RWCV S = Control Rod Drop F = HELB - RWCU M= HELB - AS + = Submerged in Suppression Pool 6 -"HELB - RWWU N = HELB - MS *: This component should be qualified to the conditions inside the reactor building due to LOCA breaks inside the primary containment. 1: P,Q,R 7: N, P,Q,R 2: N,O,P,Q,R 8: O,P,Q,R 3: A through R 9: A,B,C,E 4: F through L 10: N 5: A, B, C, E through L, N, 0 11: N,O 6: A,B,C,E,P,Q,R 12: S TABLE A STATUS/COMMENT LEGEND Q = Qual ification Documentation Complete Q'quipment in test or being replaced, or review of final documentation being completed. Qualification documentation to be completed prior to fuel load. J = Justification for Interim Operation provided . Page A- 1 CONTAINMENT ATMOSPHERE CONTROL SYSTEM TABLE A E ui Ment on Preferred Safe Shutdown Paths Accident EPN Rad Use ln forrnat i on St at us Cornrnent s Zone Code Exod Read Qual CAC-CNTR-1A R572F 1 N 1 Q't CAC-EHC-1 A R572F 1 N 1 Q CAC-EHO-FCV/1A R572B 1 N 1 9 CAC-EHO-FCV/2A R548K 1 LN 1 Qt CAC-EHO-FCV/3A R471B 1 D 1 Qt CAC-EHO-FCV/4A R471E 1 NONE 1 Qt CAC-EHO-FCV/5A R572F 1 N 1 Qt CAC-EHO-FCV/6A R572F 1 N 1 Q't CAC-EHO-TCV/4A R572F 1 N 1 Qt CAC-EHO-V/1A R572F 1 N 1 Qt CAC-EHO-V/2A R572F 1 M 1 Qt CAC-EHO-V/3A R572F 1 M 1 Q' CAC-~T-5A R572F 2 M 1 CAC-FT-6A R572F 1 N 1 Q CAC-FT-7A R572F N 1 Q CAC-LT-1A R572F 1 Q CAC-N-FN/1A R572F 1 1 Q CAC-NO-1 1 R548H 1 L 1 Q CAC-NO-13 R471B D 1 Q CAC-NO-15 R548Q 1 K 1 Q CAC-NO-17 R488N 1 NONE 1 9 CAC-NO-2 R548K 1 LM 1 Q CAC-NO-4 R471E 1 NONE 1 Q CAC-MO-6 R572B 1 M 1 Q CAC-NO-8 R471B 1 D 1 Q CAC-PT-1A R572F 2 Q CAC-PT-68A R572F 1 N 1 Q CAC-RLY-4A/CR1 R471D 1 D Q CAC-RLY-4A/CR2 R471D 1 D 1 Q CAC-RLY-FCViA88 R471D 2 D 1 Q CAC-RLY-FCV2A88 R471D 2 D Q CAC-RLY-FCV3A88 R471D 2 D 1 Q CAC-RLY-FCV4ABS R471D 2 D 1 Q CAC-TE-2A R572F 1 N 1 Q CAC-TE-3A R572F 1 M 1 Q CAC-TE-4A R572F 1 N Q CAC-TE-5A R572F 1 N 1 Q CAC-TE-6A R572F N 1 Q CAC-TE-7A R572F 1 M 1 Q Page A- 2 CONTAINMENT PURGE EXHAUST SYSTEM TABLE A Equipment on PreFevred SaFe Shutdown Paths Accident Rad Use InFor mat ion St at us Cornrnent s Zone Code Exnd Read Qual CEP-POS-V/1A R548Q 1 K 1 CEP POS V/ 1 B R548Q 1 CEP-POS-V/3A R471 J 1 D 1 C P-POS-V/3B R471 J D 1 CEP-SPV-1A R5488 1 LM 2 CEP-SPV-1B R548$ 1 LM 1 CEP-SPV-3A R471 J 1 D 1 CEP-SPV-3B R471 J D 1 Page A- 3 CONTAINMENT INSTRUMENT AER SYSTEM TABLE A Ecui ment on Preferred Safe Shutdown Paths Accident EPN Rad Use lnfor ma ion St at us Cornrnent s Z o'e Code Exud Read Qual C E A-MO-28 R522K 1 D 3 D+ Q CIA-MO-38A R522P 1 D 3 D+ Q CIA-PROS-1 A R548G 1 LM 3 LM% J JIO ¹1 CIA-PS-21A R548G 1 LM 3 LM+ Q CIA-PS-22A R548S 1 LM 3 LM+ J JIO ¹2 CI A-PT-21A R548G 1 LM 3 LM+ Q CIA-RLY-21A R548S 1 LM 3 LM~ J JEO ¹3 CIA-RLY-22A R548$ 1 LM 3 LM+ J JIO ¹4 CIA-SPV-18A R441D 1 NONE 3 Q CI A-SPV-1 1A R441D 1 NONE 3 Q CIA-SPV-12A R441D 1 NONE 3 Q CIA-SPV-13A R44 1D 1 NONE 3 Q CIA-SPV-14A R441D 1 NONE 3 Q CIA-SPV-15A R441D 1 NONE 3 Q CIA-SPV-1A R441D 1 NONE 3 Q CIA-SPV-2A R441D 1. NONE 3 Q CIA-SPV-3A R441D 1 NONE 3 Q CIA-SPV-4A R441D 1 NONE 3 Q CEA-SPV-5A R441D 1 NONE 3 Q CIA-SPV-6A R441D 1 NONE 3 Q CEA-SPV-7A R441D 1 NONE 3 Q CIA-SPV-8A R441D NONE 3 Q CIA-SPV-'3A R441D 1 NONE 3 Q CIA-TDS-1A R548G 1 LM 3 LM~ J JIO ¹5 Page A- 4 CONTAINMENT MONITORING SYSTEM TABLE A Eoui ment on Preferred Sa+e Shutdown Paths Accident EPN Rad Use Inforrnat ion St at us Cornrnent s Zone Code Exzd Read Qua1 CMS-AY-1 R548E a NONE 1 J JIO 06 CMS-AY-3 R548E NONE J JIO 86 CMiS-H-SR13/T5 R548E NONE Q CMS-H-SR13/T6 R548E NONE 1 Q CMS-H-SR13/T7 R548E 1 NONE Q CMS-H-SR13/T8 R548r-" NONE 1 Q CMS-LE-3A C435 1 PQR+ PQR+ J JIO 47 CiMS-LE-3B C588 PQR PQR J JIO 07 CMS-LS-1 C583 1 PQR 1 PQR Q CMS-LS-2 C518 1 PQR 1 PQR Q CMS-LB-3 C551 PQR 1 PQR Q CMS-LS-4 C578 1 PQR 1 PQR Q CMS-LT-2 R441F NONE 3 Q C'AS-PT-1 R548G 1 LM Q CMS-PT-3 R581F 1 D 1 Q C'nB-PT-5 R548G LM a Q CMS-PT-7 R548G 1 LM 1 Qt CMB-RE-27E C516 PQR PQR Q> CMS-RMS-HTP71 R548E NONE 1 J JIO 08 CMS-TE-1 C517 1 PQR 1 PQR J JIO 09 CMS-TE-18 c5a2 1 PQR 1 PQR Q% CAS-TE-11 C528 PQR 1 PQR Q'1 CMS-TE-13 C549 PQR a PQR CMS-TE-2 C517 1 PQR 1 PQR J JIO 09 CMS-TE-26 C596 1 PQR 1 PQR J JIQ 09 CAS-TE-27 C553 PQR 1 PQR J JIO 09 CMS-TE-29 C568 1 PQR PQR J JIO 59 CAS-TE-38 C568 PQR 1 PQR J J IO 0'9 CMS-TE-5 C584 1 PQR 1 PQR Qt CAS-TE-6 C584 PQR 1 PQR Q'% CMS-TE-8 C568 1 PQR 1 PQR CiMS-TS-4A R548E 1 NONE 1 J JIO 018 CMS-TS-4B R548E NONE 1 J JIQ 018 CiMB-TS-4C R548E 1 NONE 1 J JIO 018 CMS-TS-4D R548E NONE 1 J JIO 018 CMS-TS-5A R548E 1 NONE 1 J JIO 418 CMS-TB-5B R548E 1 NONE 1 J JIO 018 CMS-TB-5C R 548." 1 NONE 1 J JIO 018 CMS-TS-5D R548F NONE 1 J JIO I18 Page A- 5 CONTA I NUGENT RECIRCULATION A I R SYSTEM TABLE A Ecui ment on Pref'err ed Safe Shutdown Paths Accident Rad Use I n for mat i on St at us Comment s Zone Code Expd Read Qual CRA-8-FN/3A C536 1 PQR 1 PQR CRA-N-FN/4A C572 1 PQR PQR CRA-N-FN/5A C564 1 PQR 1 PQR CRA-N-FN/5C C564 1 PQR 1 PQR H Page A- 6 CONTROL ROD DRIVE SYSTEN TABLE A Equi ment on Preferred Safe Shutdown Paths Accident Rad Use Informat ion St at us Cornrnent s Zone Code Exnd Read Qual CRD-POS-126X X X X R522J 1 D 3 D+ Q CRD-POS-126X X X X R522B 1 DI 3 D!+ Q CRD-POS-127XXXX R522B 1 DI 3 DI+ Q CRD-POS-127 X X X X R522J 1 D 3 D% Q CRD-SPV-1 18A R522C 2 DI 1 Q CRD-SPV-1 17XX X X R522J 1 D 3 D+ Q CRD-SPV-117XXXX R522B 1 DI 3 DI+ Q CRD-SPV-118XXXX R522J 1 D 3 D+ Q CRD-SPV-118XXXX R522B 1 DI 3 DI+ Q CRD-SPV-182 R522P 1 D 1 Q'l CRD-SPV-9 R522P 1 D 1 Qt CRD-SV-128/XXXX R522J 1 D 3 D+ Q CRD-SV-128/X X X X R522B 1 DI 3 DI+ Q CRD-SV-121/X X X X R522J 1 D 3 D+ Q CRD-SV-121/X X X X R522B 1 DI 3 DI+ Q CRD-SV-122/XXXX R522J D 3 D+ Q CRD-SV-122/ X X X X R522B 1 DI 3 DI+ Q CRD-SV-123/XXXX R522J D 3 D+ Q CRD-SV-123/X X X X R522B 1 DI 3 D I+ Q i Page A- 7 CONTAINMENT PURSE SUPPLY SYSTEM TABLE A Equi ment on Pr eferr ed Sa e Shutdown ~ Paths Accident Rad Use Informat ion St at us Cornrnent s Zone Code Exnd Read Qual CSP-POS-V/1 R581 I 1 NONE 1 Q CSP-POS-V/18/Pi R471B 1 D Q CSP-POS-V/18/P2 R471B 1 D 1 Q CSP-POS-V/18/P3 R471B D Q CSP-POS-V/18/P4 R471B 1 D 1 Q CSP-POS-V/18/P9 R471B 1 D 1 Q CSP-POS-V/3 R471D 1 D 1 Q CSP-POS-V/7/P1 R471D 1 D 1 Q CSP-POS-V/7/P2 R471D 1 D 1 Q CSP-POS-V/7/P3 R471D D 1 Q CSP-POS-V/7/P4 R471D 1 D 1 Q CSP-POS-V/7/P9 R471D 1 D 1 Q CSP-POS-V/8/Pi R471J 1 D Q CSP-POS-V/8/P2 R471 J 1 D 1 Q CSP-POS-V/8/P3 R471 J 1 D 1 Q CSP-POS-V/8/P4 R471J 1 D 1 Q CSP-POS-V/8/P9 R471 J 1 D 1 Q CSP-SPV-1 R581F D 1 Q CSP-SPV-18A R471B 1 D 1 Q CSP-SPV-18B R471B 1 D 1 Q CSP-SPV-3 R471B D 1 Q CSP-SPV-7A R471B 1 D 1 Q CSP-SPV-7B R471B 1 D 1 Q CSP-SPV-8A R471J 1 D Q CSP-SPV-8B R471J 1 D 1 Q CSP-V-97 R581I 1 NONE 1 Q CSP-V-98 R471D 1 D Q Page A- 8 ELECTRICAL SYSTEM TABLE A Ec u 1 @sent f f on Pr e erred Sa e Shut down Paths Accident Rad Use Informat ion St at us Cornrnent s Zone Code Exod Reed Qual E-CONN-X 188D/81 C511 2 PQR 1 PQR E-CONN-X 188D/82 C511 2 PQR 1 PQR E-CONN-X 182A/81 CS34 2 PQR 3 PQR E-CONN-X 182A/82 C534 2 PQR 3 PQR E-PP-7AE+ R471D D 1 E-PP-8AE+ R471D 1 D 1 E-X-188A C587 2 PQR 3 PQR E-X-188B C587 2 PQR 3 PQR E-X-188C C51 1 2 PQR 3 PQR E-X-188D C511 PQR 3 PQR E-X-181 A C511 2 PQR 3 PQR E-X-181B C511 2 PQR 3 PQR E-X-181 C C511 2 PQR 3 PQR E-X-181D C511 PQR 3 PQR E-X-182A C534 2 PQR 3 PQR E-X-182B C534 2 PQR 3 PQR E-X-183A C534 PQR 3 PQR E-X-183B C534 PQR 3 PQR E-X-183C C534 PQR 3 PQR E-X-183D C534 2 PQR 3 PQR E-X-184A C511 PQR 3 PQR E-X-184B C511 2 PQR 3 PQR E-X-184C C534 2 PQR 3 PQR E-X-184D C534 2 PQR 3 PQR E-X-185A CS87 2 PQR 3 PQR E-X-185B C511 2 PQR 3 PQR E-X-185C C534 2 PQR 3 PQR E-X-185D C534 2 PQR 3 PQR E-X-187A C475 2 PQR 3 PQR E-X-187B C475 2 PQR 3 PQR 1 Paoe A- 9 EQUIPMENT DRAINS RADIOACTIVE SYSTEM TABLE A E uioment on Preferred Sa.e ShutCown Paths Accident Rad Use Informat ion Status Comments Zone Code Exod Reed Qual EDR-POS-V/28 R441C 1 NONE 1 JIO ¹11 EDR-SPV-28 R471B D 1 Page 4-18 FLOOR DRAIN RADlOACTIVE SYSTEM TABLE A Ecui ment on Prefet r ed Sa~e Shutdown Paths Accident EPN Rad Use I n forrnat i on St at us Cornrnent s Zone Code Exvd Reed Qual FD R-LS-4 1 R422J 1 NONE Q FDR-LS-45 R4a~ZC 1 NONE 1 Q FDR-LS-46 R42BD 1 NONE 1 Q FDR-POS-V/4 R441C 1 NONE 1 J JIO 012 FDR-SPV-4 R471B 1 D 1 Q Page A-11 FUEL POOL COOLING SYSTEM TABLE A Eoui ment an Preferred Safe Shutdown Paths Accident EPN Rad Use Infor mat ion Status Comments Zone Code Eood Read Qual FPC-MO-153 R441G NONE 3 FPC-MO-154 R441G 1 NONE 3 Page A-12 HIGH PRESSURE CORE SPRAY SYSTEM TABLE A Equi ment on Preferred Sa.e Shutdown Paths Accident Rad Use Inforrnat ion St at us Cornrnent s Zone Code Exod Reed Qual HPCS-F IS-6 R471B D 3 D+ Q HPCS-FT-5 R471B 1 D 3 D+ J JIO 013 HPCS-LS-1A R441 J 1 C 3 C+ Q HPCS-LS-1B R441 J 1 C 3 C+ Q HPCS-LS-2A R441 J C 3 C+ Q HPCS-LS-2B R441F 1 NONE 3 Q HP CS-M-P / 1 R422D NONE 3 Q HPCS-M-P/3 R422D 1 NONE 3 Q HPCS-MO-1 R422D 1 NONE 3 Q HP CS-MO-18 R441C 1 NONE 3 Q HPCS-MO-11 R441C 1 NONE 3 Q HP CS-MO-12 R422D NONE 3 Q HPCS-MO-15 R441C 1 NONE 3 Q HPCS-MO-23 R441C 1 NONE 3 Q HPCS-MO-4 R522H 1 D 3 D+ Q HPCS-PS-12 R471B D 3 D+ Q I Page A-13 ,. RRC HYDRAULIC CONTROL TABLE A Eoui ment on Preferred Safe Shutdown Paths Accident EPN Rad Use Informat ion Status Comments Zone Code Exod Read Qual HY-V-17A R522C '1 DI 1 HY-V-17B R522J 1 D HY-V-18A R522C 1 DI 1 HY-V-18B R522J 1 D 1 HY-V-19A R522C 1 DI 1 HY-V-19B R522J D 1 HY-V-28A R522C 1 DI 1 HY-V-28B R522J 1 D 1 Page A-14 LEAK DETECTION SYSTEM TABLL A E ui ment on Prefers ed Safe Shutdown Pat'hs Accident Rad Use Inforrnat ion St at us Cornrnent s Zone Code Ex ad Read Qual LD-TE-1A R522F HI 4 HI LD-TE-ic R522F HI 4 HI LD-TE-1E R548B J 4 J LD-TE-29A R5810 NO 11 NO LD-TE-29C R5810 NO 11 NO LD-TE-2A R522F HI 4 HI LD-TE-2C R522F HI 4 HI LD-TE-2E R548B J 4 J LD-TE-38A R5220 G 11 NONE LD-TE-38C R5220 G 11 NONE LD-TE-31 A R5810 NO 11 NO LD-TE-31 C R5810 NO 11 NO LD-TE-3A R522F HI 4 HI LD-TE-3C R522F HI 4 HI LD-TE-3E R548B J 4 J" LD-TE-4A R441 I AB 9 AB LD-TE-5A R422L AB 9 AB LD-TE-GA R441I AB 9 AB Page A-15 LOW PRESSURE CORE SPRAY SYSTEM TABLE A Ecui ment on Preferred Safe Shutdown Paths Accident EPN Rad Use In for mat i on St at us Comment s Zone Code Exod Read Qual LPCS-M-P/2 R422C 1 NONE 1 5 LPCS-MO-1 R441B 1 NONE 1, 5 LPCS-MO-1 1 R422C 1 NONE 1 5 LPCS-MO-12 R441B 1 NONE 1~ 5 LPCS-MO-5 R522B 1 DI 1,5 'I+ Page A-16 LOCAL POWER RANGE MONITOR SYSTEM TABLE A Eaui ment on Pr eferred Bat e Shutdown Paths Accident Rad Use In for mat i on Status Cornrnent s Zone Code Exod Read Qual LPRM-CONN-12BCD C581 2 PQRS 8 J JIQ ¹15 LPRM-CONN-13CDA C581 PQRS 8 J JIO ¹15 LPRM-CONN-14ABD C581 2 PQRS 8 J JIO ¹15 LPRM-CONN-15ABC C581 2 PQRS 12 8 J JIO ¹15 LP RM-CONN-16BCD C581 2 PQRB 12 8 J JIO ¹15 LP RM-CONN-21 ABC C581 2 PQRS 12 8 J JIO ¹15 LPRM-CONN-22DAB C581 2 PQRB 12 8 J JIO ¹15 LPRM-CONN-23CDA C581 2 PQRS 12 8 J JIO ¹15 LPRM-CONN-24BCD C581 2 PQRS 8 J JIO ¹15 LPRM-CONN-25ABC C581 2 PQRS 12 8 J JIO ¹15 LPRM-CONN-26DAB C581 2 PQRS 12 8 J JEO ¹15 LPRM-CONN-27CDA C581 2 PQRS 12 8 J JIO ¹15 LPRM-CONN-31CDA C581 PQRS 12 8 J JIQ ¹15 LPRM-CONN-32DAB C581 2 PQRS 12 8 J JIO ¹15 LPRM-CONN-33ABC C581 2 PQRB 12 8 J JIQ ¹aS LPRM-CONN-84BCD C581 2 PQRS 12 8 J JIO ¹15 LPRM-CONN-35CDA C581 2 PQRS 12 8 J JIO ¹15 LPRM-CONN-36DAB C581 2 PQRS 12 8 J JIO ¹15 LPRM-CONN-37ABC C581 2 PQRB 12 8 J JIQ ¹15 LP RM-CONN-41CDA C581 2 PQRS 12 8 J JIO ¹15 LPRM-CONN-42BCD C581 2 PQRS 12 8 J JIQ ¹as LPRM-CONN-43ABC C581 2 PQRS 12 8 J JIO ¹15 LPRM-CONN-44DAB C581 2 PQRS 12 8 J JIO ¹15 LPRM-CONN-45CDA CS81 2 PQRB 12 8 J JIO ¹15 LPRM-CONN-46BCD C581 2 PQRS 12 8 J JIO ¹15 LPRM-CONN-47ABC C581 2 PQRS 12 8 J JIQ ¹15 LPRM-CONN-S1ABC C581 2 PQRS 12 8 J JIO ¹15 LPRM-CONN-52BCD C581 2 PQRS 12 8 J JIO ¹15 LPRM-CONN-53CDA C581 2 PQRS 12 8 J JIO ¹15 LPRM-CONN-54DAB C581 2 PQRS 12 8 J Jlo ¹15 LPRM-CONN-55ABC C581 2 PQRS 12 8 J JIO ¹15 LPRM-CONN-56BCD C581 2 PQRS 12 8 J JIQ ¹15 LPRM-CONN-S7CDA C581 2 PQRS 12 8 J JEO ¹15 LP RM-CONN-61 ABC CS81 2 PQRS 12 8 J JIQ ¹15 LPRM-CONN-62DAB C581 PQRS 12 8 ~ J JIQ ¹as LPRM-CONN-63CDA C581 2 PQRS 12 8 J JIO ¹15 LPRM-CONN-64BCD C581 2 PQRS 12 8 J JIO ¹15 LPRM-CONN-65ABC C581 PQRS 12 8 J JIQ ¹15 LPRM-CONN-66DAB C581 PQRS 12 8 J JIO ¹15 LPRM-CONN-72DAB C581 J 1 PQRS 12 8 J JIQ ¹15 LPRM-COND-73ABC C581 2 PQRS a2 8 J JIO ¹1S LPRN-CONN-74BCD C581 2 PQRB 12 8 J JIQ ¹15 LPRM-CONN-75CDA C581 2 PQRS a2 8 J J 0 ¹15 LP RM-DET-12BCD C581 1 PQRS 12 8 J JIO ¹14 LPRM-DET-13CDA C581 1 PQRS 12 8 J JEO ¹14 Page A-17 LOCAL POWER RANGE MONITOR SYSTEM TABLE A E ui ment on Pr eferred Safe Shutdown Paths Accident Rad Use Infor mat ion Status Comment s Zone Code Excd Reed Qual LP RM-DET-14ABD C581 1 PQRS 12 S J JIO ¹14 LP RN-DET-15ABC C581 1 PQRB 12 S J JIO ¹14 LPRN-DET-16BCD C581 1 PQRS 12 S J JIO ¹14 LP RN-DET-21 ABC C581 1 PQRS 12 S J JIO ¹14 LPRM-DET-22DAB C581 1 PQRS 12 S J JIO ¹14 LPRM-DET-23CDA C581 1 PQRS 12 S J JIO ¹14 LPRN-DET-24BCD C581 1 PQRB 12 S J JIO ¹14 LPRM-DET-25ABC C581 1 PQRS 12 B J JIO ¹14 LPRM-DET- 6DAB C581 1 PQRS 12 S J JIO ¹14 LPRN-DET-27CDA C581 1 PQRS 12 S J JIO ¹14 LPRM-DET-31CDA C581 1 PQRS 12 S J JIO ¹14 LPRM-DET-32DAB C581 1 PQRS 12 S J JIO ¹14 LPRN-DET-33ABC C581 1 PQRS 12 S J JIO ¹14 LPRN-DET-34BCD CS81 1 PQRS 12 S J JIO ¹14 LPRM-DET-35CDA C581 1 PQRB 12 S J JIO ¹14 LPRM-DET-36DAB C581 PQRS 12 S J JIO ¹14 LPRM-DET-37ABC C581 1 PQRS 12 S J JIO ¹14 LPRM-DET 41CD C581 1 PQRS 12 S J JIO ¹14 LPRN-DET-42BCD C581 1 PQRS 12 S J JIO ¹14 LPRM-DET-43ABC C581 1 PQRB 12 S J JIO ¹14 LPRM-DET-44DAB C581 1 PQRS 12 S J JIO ¹14 LPRN-DET-45CDA C581 1 PQRS 12 S J JIO ¹14 LPRN-DET-46BCD C581 1 PQRS 12 S J JIO ¹14 LPRM-DET-47ABC C581 1 PQRS 12 S J JIO ¹14 LP RN-DET-51 ABC C581 1 PQRS 12 S J JIO ¹14 LPRM-DET-52BCD C581 1 PQRS 12 S J JIQ ¹14 LPRM-DET-S3CDA C581 P.QRS 12 S J JIO ¹14 LPRN-DET-54DAB C581 1 1 PQRS 12 S J' JIO ¹14 LPRM-DET-5SABC C581 1 PQRB 12 S JIQ ¹14 LPRN-DET-56BCD C581 1 PQRS 12 S J JIO ¹14 LPRM-DET-57CDA C581 PQRS 12 B J JIO ¹14 LPRN-DET-61ABC C581 1 PQRS 12 S J JIO ¹'4 LPRM-DET-62DAB C581 1 PQRB 12 S J JIO ¹14 LPRM-DET-63CDA C581 1 PQRB 12 S J JIO ¹14 LPRN-DET-64BCD C581 1 PQRB 12 S J JIQ ¹14 LPRM-DET-65ABC C581 1 PQRS 12 S J JIO ¹14 LPRM-DET-66DAB C581 PQRS 12 S J JIO ¹14 LPRM-D"T-72DAB C581 1 PQRS 12 S J JIO ¹14 LPRM-DET-73ABC C581 1 PQRB 12 S J JIO ¹14 LPRN-DET-74BCD C581 PQRS 12 B J JIO ¹14 LPRN-DET-75CDA C581 1 PQRS 12 S J JIQ ¹14 Page A-18 MAIN STEAM SYSTEM TABLE A Equi ment on Preferred Safe Shutdown Paths Accident Rad Use Information Status Comments Zone Code Expd Read Qual MS-CONN-V22A/Ji C513 2 PQR 1,5 PQR MS-CONN-V22A/J2 C513 2 PQR 1,5 PQR MS-CONN-V22A/J3 C513 2 PQR 1,5 PQR MS-CONN-V22A/J4 C513 PQR 1,5 PQR MS-CONN-V22B/Ji C513 PQR ip5 PQR MS-CONN-V22B/J2 C513 2 PQR 1,5 PQR MS-CONN-V22B/J3 C513 2 PQR 1,5 PQR MS-CONN-V22B/J4 C513 2 PQR 1~5 PQR MS-CONN-V22C/Jl C513 2 PQR 1,5 PQR MS-CONN-V22C/J2 C513 2 PQR 1,5 PQR MS-CONN-V22C/J3 C513 2 PQR 1,5 PQR MS-CONN-V22C/J4 C513 2 PQR PQR MS-CONN-V22D/Ji C513 2 PQR 1,5 PQR MS-CONN-V22D/J2 C513 2 PQR 1,5 PQR MS-CONN-V22D/J3 C513 2 PQR 1,5 PQR MS-CONN-V22D/J4 C513 2 PQR 1,5 PQR MS-DP I S-1 1 A R581K 1 D 18 NONE MS-DPIS-1 iB R471D 1 D 18 NONE MS-DPIS-818A R581K 1 D 18 NONE MS-DP IS-818B R471D D 18 NONE MS-DP I S-8A R581K 1 D 18 NONE MS-DP I S-8B R471D 1 D 18 NONE MS-DP I S-SA R581K D 18 NONE MS-DP I S-9B R471 D D 18 NONE MS-L I S-1 88A R522K 1 D 3 D+ Q 9 MS-L I S-188B R522C 1 DI 3 DI+ Q9 MS-L I S-24A R522K 1 D Q MS-L I S-24B R522H 1 D 1 Q MS-L I S-31 A R522K 1 D 3 D+ Q MS-LIS-31B R522C 1 DI 3 DI~ Q MS-L IS-31 C R522K 1 D 3 Dm Q MS-L IS-31D R522C D 3 D+ MS-LIS-37A R522P D D+ MS-LIS-37C R522P 1 D 3 D+ MS-L I S-38A R522P 1 D 3 D+ MS-L ITS-26A R522K D D+ MS-LITS-26B R522P 1 D 1 MS-LITS-44A R471B 1 D MS-MO-19 R5810 1 NO 2 NO+ MS-MO-67A R5810 1 NO 1,5 NO+ MS-MO-67B R5810 1 NO 1,5 NO+ MS-MO-67C R5810 NO 1,5 NO+ MS-MO-67D R5810 1 NO 1,5 ~% NO+ MS-POE-1A C547 PQR PQR MS-POE-lB C547 PQR 3 PQR Page A-19 MAIN STEAN SYSTEM TABLE A Eoul ment on Preferred Safe Shutdown Paths Accident EPN Rad Use Informat ion St at us Comment s Zone Code Expd Read Qual NS-POE-1 C C547 1 PQR 3 PQR NS-POE-1D C547, 1 PQR 3 PQR NB-POE-2A C547 1 PQR 3 PQR NS-POE-2B C547 1 PQR 3 PQR NS-POE-2C C547 1 PQR 3 PQR NS-POE-2D C547 1 PQR 3 PQR NS-POE-3A C547 1 PQR 3 PQR NS-POE-3B C547 1 PQR 3 PQR MS-PQE-3C C547 1 PQR 3 PQR NS-POE-3D C547 1 PQR 3 PQR NS-POE-4A C547 1 PQR 3 PQR Q NS-POE-4B C547 1 PQR 3 PQR Q MB-POE-4C C547 PQR 3 PQR Q NS-POE-4D C547 1 PQR 3 PQR Q NB-POE-5B C547 PQR 3 PQR Q NS-POE-5C C547 1 PQR 3 PQR Q MS-POS-V/22A/1 C513 1 PQR 1%5 PQR Qf NS-POS-V/22A/2 C513 PQR 1,5 PQR QT NS-POS-V/22R/3 C513 1 PQR 1,5 PQR Q1 NS-POS-V/22A/4 C513 1 PQR 1,5 PQR MS-POS-V/22B/ 1 C513 1 PQR 1,5 PQR NS-PQS-V/22B/2 C513 1 PQR 1,5 PQR Q1 NS-POS-V/22B/3 C513 1 PQR 1,5 PQR Q'I NS-PQS-V/22B/4 C513 1 PQR 1,5 PQR Qf MB-POS-V/22C/1 C513 1 PQR 1,S PQR Qt NB-PQS-V/22C/2 C513 .1 PQR 1,5 PQR QS MS-POS-V/22C/3 C513 1 PQR 1,5 PQR Qf NS-POS-V/22C/4 C513 1 PQR 1~5 PQR Qf NS-POS-V/22D/1 C513 1 PQR 1,5 PQR NS-POS-V/22D/2 C513 1 PQR 1,5 PQR Q1 MS-POS-V/22D/3 C513 1 PQR 1~5 PQR Q'i NS-PQS-V/22D/4 C513 PQR 1,5 PQR Qt NS-POT-1A C547 1 PQR 3 PQR Q NS-POT-1B C547 PQR 3 PQR Q MS-POT-1C C547 1 PQR 3 PQR Q NS-POT-1D C547 1 PQR 3 PQR Q MS-POT-2A C547 1 PQR 3 PQR Q NS-POT-2B C547 PQR 3 PQR Q NS-POT-2C C547 1 PQR 3 PQR Q NS-POT-2D C547 PQR 3 PQR Q NB-POT-3A C547 PQR PQR Q tiiS-PQT-3B C547 1 PQR PQR Q MS-POT-3C C547 1 PQR 3 PQR Q NS-POT-3D C547 1 PQR PQR Q MS-POT-4A C547 PQR 3 PQR Q Page A-28 MAIN STEAM SYSTEM TABLE A Eoui ment or Preferred Sa+e Shutdown Paths Accident Rad Use Ir for mat ion St at us Comment s Zone Code Excd Read Qual NB-POT-4B C547 1 PQR 3 PQR Q MS-POT-4C C547 PQR 3 PQR Q NS-POT-4D .C547 1 PQR 3 PQR Q MiS-POT-5B C547 1 PQR PQR Q NS-POT-SC C547 1 PQR PQR Q MS-PS-28A R 522!C D 1 Q 1 MS-PS-28B R522H 1 D 1 Qt MS-PB-23A R522K D 1 Q t MS-PS-23B R522P D 1 Qt MS-PS-39A R522P 2 D 3 D+ Q MB-PS-39B R522P 2 D 3 D+ Q MS-PS-39C , R522P D 3 D+ Q MS-PS-39D R522P 2 D 3 D+ Q NS-PS-39E R522P 2 D 3 D+ Q NB-PS-39F R522P 2 D 3 D+ Q MS-PS-39G R522P D D+ Q NS-PS-39H R522P D D+ Q MS-PS-39J R522P 2 D D+ Q NS-PS-39K R522P 2 D 3 D+ Q MS-PS-39L R522P D 3 D+ Q MS-PS-39M R522P 2 D 3 D+ Q MS-PS-39N RS22P 2 D 3 D+ Q NS-PS-39P R522P 2 D 3 D+ Q MS-PS-39R R522P 2 D D+ Q NS-PS-39S R522P 2 D D+ Q iNS-PS-39U R522P 2 D D+ Q NS-PS-39V R522P 2 D 3 D+ Q NS-PB-47A R522, D 3 D+ Q NS-PS-47B R522C 1 DI 3 DI+ Q NS-PS-47C R522K 1 D 3 D+ Q MiS-PS-47D R522C 1 DI 3 DI+ Q MS-PB-48A R522P 1 D D+ Q MS-PS-4SC R522P 1 D 3 D+ Q MS-PT-51A R522P 1 D 3 D+ gf NiB-RE-3A R5810 1 NOS 12 S J 116 NS-RE-3B R5810 NOS 12 S J JIO 516 iNS-SPV-22A2 C51S 1 PQR is 5 PQR Qt MB-SPV-22R3 C515 1 PQR 1,5 PQR Q% MS-SPV-22B2 C515 1 PQR 1,5 PQR Qt lB SPV 22B3 MS-SPV-22C2 C515 C515 PQR PQR 1,5 1,5 PQR PQR Q'IO Qf Q5 MS-BPV-22C3 C515 PQR 1,5 PQR NS-SPV-22D2 CS15 J PQR 1.5 PQR NB-BPV-22D C515 1,5 PQR NS-SPV-3DA C547 1 PQR J ~ ~ '>> '- Page A-R1 NAIN STEAN SYSTEN TABLE A Accident Rad Use Infor mat i os St at us Cornrnent s Zone Code Egad Read Qual NS-SPV-4AA C547 1 PQR 3 PQR Q) NS-SPV-4BA C547 1 PQR 3 PQR Q') NS-SPV-4CA C547 1 PQR 3 PQR NS-SPV-4DA C547 PQR 3 PQR Q) NS-SPV-5BA C547 1 PQR 3 PQR Qt NS-SPV-5CA C547 1 PQR 3 PQR Q') Page A-22 MAIN STEAN LEAKAGE CONTROL SYSTEM TABLE A ~Ecui rrrent cn unstarved Safe Bhutdcwn Paths Accident Rad Use Xnfor mat ion St at us Cornrnent s Zone Code Expd Read Qual NSLC-M-FN/2 R581K D 1,5. + Q NSLC-MO-18 R5810 NO 1,5 NO+ Q NSLC-NO-3A R5810 NO 1,5 NO+ Q MSLC-NO-3B R5810 NO 1,5 NO~ Q NSLC-MO-3C R5810 NO 1,5 NO+ Q MSLC-MO-3D R5810 NO 1, 5 NO+ Q NSLC-MO-4 R5810 NQ 1,5 NO+ Q NSLC-NO-5 R5810 NO 1,5 NO+ Q NSLC-NO-9 R581Q NO 1,5 NO+ Q iNSLC-PS-28 R522K D f~5 + Q NSLC-PS-24 R522K D 1~5 Q NSLC-PS-25 R522K D 1 5 Q NSLC-PS-68 R522K D Q NrSLC-PT-1 1 R522K D f~5 + Q t NSLC-PT-13 R52' D f~5 + Q 0 NSLC-RLY-CR/1 R522K D 1,5 Q NSLC-RLY-CR/3 R522K D fq5 Q +SLC-RLY-CR/4 R522K D ft5 - Q NSLC-RLY-CR/5 R522K D Q NSLC-TD-TK/2 R522K D f~ 5 Q ~ 'i i Page A-23 PROCESS INSTRUMENTATION SYSTEM TABLE A E ui ment on Preferred Safe Shutdown Paths Accident Rad Use Inforrnat ion St at us Comment s 2 o'e Code Ex ad Read Qual PI-V-X258 R522H D Q P -V-X251 R522H 1 D Q P I-V-X253 R522H 1 D Q PI-V-X256 R522P 1 D Q P I-V-X257 R522P D Q P I -V-X259 R522P 1 D Q P I-V-X262 R522H 1 D Q PI -V-X263 R522H 1 D Q PI-V-X264 R522H 1 D Q ¹-V-X265 R471D 1 D Q P I -V-X266 R522H 1 D Q PI-V-X267 R522H 1 D Q PI-V-X268 R522P 1 D Q PI-V-X269 R471A 1 D Q Page A-24 PROCESS INSTRUMENTATION SYSTEM TABLE A Ecuioment on Preferred Safe Shutdown Paths Accident EPN Rad Use In for mat ion St at us Comment s Zone Code Ex ad Read Qual PSR-V-883/A R422J NONE 1 Q PSR-V-883/8 R422I NONE Q PSR-V-X73/2 R522P D 1 Q PSR-V-X77A/2 R581K D 1 Q PSR-V-X77A/4 R581K D 1 Q PSR-V-X88/2 R522H D 1 Q PSR-V-X82/2 R471D D 1 Q PSR-V-X82/8 R471D D 1 Q PSR-V-X83/2 R471D D 1 Q PSR-V-X84/2 R471A D 1 Q PSR-V-X88/2 R441D NONE 1 Q Page A-25 REACTOR BUILDING CLOSED COOLING WATER SYSTEM TABLE A Equi ment on Preferred Safe Shutdown Paths Accident EPN Rad Use Inforrnat ion Status ComMents Zone Code Exod Reed Qual RCC-MO-138 R548L 1 NONE 1 Q RCC-MO-21 R518S 1 EF 1 Q RCC-MO-5 R5188 1 EF 1 Q Page A-26 REACTOR CORE ISOLATION CODLING SYSTEM TABLE A Equi nugent on Prefer v ed Sa e Shutdown Paths Accident Rad Uee Infovrnat ion Statue Comments Zone Code Exod Reed Qual RCEC-DP ES-13A R471D D 9 NONE Q RC IC-DP E S-7A R471D D 9 NONE Q RC I C-NO-1 3 R548H L 6 Q RCIC-VO-19 R441I AB 6 AB+ Q RC I C-MO-3 1 R441I AB 6 AB~ Q RCIC-NO-64 R548B J 6 Q RC I C-NO-68 R471I NONE 6 Q RCEC-MO-69 R441 I AB 6 AB+ Q RC I C-NlO-8 R518S EF 6 E+ Q RCIC-NO-88 R471I NONE 6 Q RC I C-PS-28A R471D D 9 NONE Q RC I C-PS-RPC R471D D 9 NONE Q 1 I I Page A-27 REACTOR BUILDING EXHAUST AIR(HVAC) SYSTEN TABLE A Eoui Ment on Preferred Sa~e Shutdown Paths Accident Rad Use Inforrnat i on St at us Cornrnent s Z o'e Code Ex pd Read Qual REA-DPT-1A 1 R572N 1 N 1~5 Q REA-DPT-1 A2 R572C 1 N 1,5 Q REA-DPT-1A3 R572C 1 N 1,5 Q REA-DPT-1 A4 R572L 1. NONE 1, 5 Q REA-POS-V/ 1 R572N 1 N 1 + Q REA-RE-9A R572C 1 N 1,5 Q REA-RE-9B R572C 1 N 1,5 + Q 0 REA-RLY-CR1 R522K 2 D 1 + Q REA-SPV-V/ 1 R522K D 1 + Q Page A-28 REACTOR FEEDWATER SYSTEM TABLE A E ui ment on Prefer red Safe Shutdown Paths Accident Rad Use Inforrnat ion Stat ua Comment s Zone Code Exod Reed Qual RF4J-NO-65A R58i0 1 NO 8 RFN-KO-65B R5810 i NO 8 Page A-29 RESIDUAL HEAT REMOVAL SYSTEM TABLE A Eoui ment on Prefer red Safe Shutdown Paths Accident Rad Use In forrnat i on St at us Cornrnent s Zone Code E>< ad Read Qual RHR-DP I S-12A R581B D 3 D+ Q RHR-F I S- 1 8A R581B D 3 D+ Q RHR-FT-15A R581 B 1 D 3 D+ Q '7 RHR-LS-1 1A R471F NONE Q RHR-LS-1 f B R471F 2 NONE 3 Q RHR-LS-11C R471F 2 NONE 3 + Q RHR-LS-11D R471F 2 NONE 3 Q RHR-M-P/2A R422J 1 NONE + Q RHR-MO-11A R471F NONE 3 Q RHR-MO-11B R471E 1 NONE 3 Q RHR-MO- 124A R471F 1 NONE 3 Q RHR-MO-124B R471 F 1 NONE 3 Q RHR-MO-125A R471E 1 NONE 3 Q RHR-MO-125B R471E 1 NONE 3 Q RHR-MO-134 A R548M M 3 Q RHR-MO-134B R548L NONE 3 Q RHR-MO-16A R548B 1 J f Q RHR-MO-1 6B R581M 1 NONE 1 Q RHR-MO-17A R548B 1 J 1 Q RHR-MO-17B R581 M 1 NONE 1 Q RHR-MO-21 R441J 1 C 3 C+ Q RHR-MO-24 A R471F 1 NONE Q RHR-MO-24B R471E f NONE 1 Q RHR-MO-26A R471F NONE 3 Q RHR-MO-26B R471E 1 NONE Q RHR-MO-27A R471A 1 D 3 D+ Q RHR-MO-27B R471E 1 NONE 3 Q RHR-MO-3A R548N 1 NONE 3 Q RHR-MO-48 R548J 1 NONE 3 Q RHR-MO-42A R5220 1 G 3 G+ Q RHR-MO-42B R522G 1 HI 3 H1+ Q RHR-MO-42C R5220 1 G 3 G+ Q RHR-MO-47A R572L 1 NONE 3 RHR-MO-48A R548N 1 NONE RHR-MO-4A R441G 1 NONE 3 Q RHR-MO-4B R441F 1 NONE 3 Q RHR-MO-4C R44' 1 C 3 C+ Q RHR-MO-52A R572L NONE 3 Q RHR-MO-53A R518S 1 EF 3 EF+ Q RHR-MO-53B R581M 1 NONE Q RHR-MO-64A R441 G 1 NONE 3 Q RHR-MO-64B R441F NONE Q RHR-MO-64C R441J C C+ RHR-MO-68A R548N 1 NONE 3 Q RHR-MO-6A R422J NONE 3 Q Page A-38 RESIDUAL HEAT REMOVAL SYSTEM TABLE A Equi nugent on Preferred Safe Shutdown Paths Accident EPN Rad Use Inforrnat ion St at us Cornrnent s Zone Code Expd Read Qual RHR-MO-73A R572L NONE 3 RHR-NO-73B R572I NONE 3 RHR-NO-74 A R572L NONE 3 RHR-MO-8 R581 I NONE 3 Q RHR-NO-87A R572L NONE 3 RHR-NO-93 R548J NONE 3 RHR-NO-99A C514 PQR 1 PQR Q RHR-PS-16A R581B D 3 D+ RHR-PS-19A R581B D 3 D+ RHR-V-75A R548N NONE 3 RHR-V-75B R548J NONE 3 Pane A-31 REACTOR BUILDING OUTSIDE AIR(HVAC) SYSTEM TABLE A Ecui ment on Preferred Safe Shutdown Paths Aaaident EPN Rad Use Info'at ion St at us Comment s Zone Code Exod Rend Qual ROA-POS-V/ 1 R572F 1 M J JIO ¹17 ROA- RLY-CR 1 A R548$ 2 L,M 1 Q ROA-SPV-188 R5488 1 L,M 1 Q ROA-SPV-11 R522K 2 D 3 Q ROA-SPV-12 R471H NONE 3 Q ROA-SPV-13 R572F 2 NONE 3 Q ROA-SP V-1 5 R548C 2 M 1 Q REACTOR PROTECTION SYSTEM TABLE A E ui n>ent on Preferred Safe Shutdown Paths Accident EPN Rad Use Infor mat ion St at us Carnrnent s Zone Code Exod Read Qual RPS-PS-2A R522K 1 D RPS-PS-2B R522H 1 D Page A-33 REACTOR BUILDING RETURN AIR (HVAC) SYSTEM TABLE A Ecui ment on Pr eferred SaFe Shutdown Paths Accident EPN Rad Use In for mat i on St at us Cornrnent s Zone Code Excd Reed Qual RRA-M-FN/1 1 R522N NONE 3 RRA-M-FN/12 R471H 1 NONE 3 RRA-M-FN/13 R572D NONE 3 RRA-M-FN/15 R548E 1 NONE 1 RRA-M-FN/2 R441G 1 NONE 3 RRA-8-FN/4 R441C 1 NONE 3 RRA-M-FN/5 R441B 1 NONE 3 RRA-RMS-FN/2 R441G 2 NONE 3 RRA-RMS-FN/4 R441C 2 NONE 3 BRA-RMS-FN/5 R441B 2 NONE 3 Page A-34 REACTOR REC I RCULATION SYSTEM TABLE A Eoui ment on Preferred Safe Shutdown Paths Accident EPN Rad Use Inforrrrat ion St at us Corrrrnent s Zone Code Exnd Read Qual RRC-MO-16A R581F 1 D RRC-MO-1KB R581K 1 D RRC-PS-18A R471B 1 D RRC-V-28 R581K 1 D Page A-35 REACTOR WATER CLEANUP SYSTEM TABLE A Equi ment on Pr efer red Sare Shutdown Paths Accident EPN Rad Use Ir tor mat ion St at us Cornrnent s Zone Code Exzd Read Qual RWCU-FT-15 R522C DI 4 I RWCU-FT-36 R522C 1 DI 4 I RWCU-FT-4 1 R522C 1 DI 4 I RWCU-MO-4 R522F 1 HI 1, 4 HI+ RWCU-NO-48 R518S 1 EF 1,4 F~ h' Page A-36 STANDBY GAS TREATMENT SYSTEM TABLE A E ui ment on Preferred Safe Shutdown Paths Accident Rad Use Informat ion Stat us Comments Zone Code Exud Read Qual SST-EHC-1A 1 R572N 5 Q SST-EHC- 1A2 R572N 5 Q SST-EHO-1A 1 R572N3 5 Qs SGT-EHO-1 A2 R572N3 5 Q% SST-FS-2A2 R572N4 Q SGT-FT-1A1 rR572N2 5 Q SST-F T-1A2 R572N2 5 Q SGT-M-FN/ 1 Ai R572N3 5 Q SGT-M-FN/1A2 R572N3 5 Q SGT-ME-6A1 . R572N 5 Q SGT-ME-6A2 R572N 5 Q SST-ME-6A3 R572N 5 Q SGT-ME-7A 1 R572N 5 Q SBT-ME-7A2 R572N 5 Q SGT-ME-7A3 R572N 5 Q A 'ST-MO-1 R572N4 5 Q SBT-MO-3A 1 R572N4 Q SST-MO-3A2 R572N4 5 Q SGT-MO-4A 1 R572N4 5 Q SGT-MO-4A2 R572N4 5 Q SGT-MO-5A1 R572N4 Q SBT-MO-5A2 R57-N4 5 Q SGT-MO-5Bi R57PN4 5 Q SST-MO-582 R572N4 5 Q SGT-POS-V/2A R572N 5 Q SGT-SPV-2A R572N 5 Q SST-SPV-F 1 R572N1 2 M 5 Q SST-SPV-F2 R57PN1 2 5 Q SGT-SPV-F3 R57PN1 2 M 5 Q SGT-TS-EH1A11 R572N6 1 M 5 Q SGT-TS-EH1A1 18 R572N6 1 M Q SGT-TS-EH1A111 R572N6 1 M 5 Q SST-TS-EH1A112 R57P.N6 1 5 Q SST-TS-EHl Ai 13 R572N6 1 M 5 Q SST- TS-EH I A 1 14 R572N6 1 5 Q SGT-TS-EH1A115 R572N6 1 M 5 Q SGT- TS-EH1A 1 16 R572N6 1 M Q SGT-TS-EH1A1 17 R572N6 M 5 Q SBT-TS-EH1A118 R572N6 1 M 5 Q SGT-TS-EH1A12 R572N6 1 M 5 Q SGT-TS-EH1A'3 R572N6 1 M 5 Q SST-TS-EH1A14 R572N6 1 5 Q SBT-TS-EH1 A15 R572N6 1 M 5 Q SGT-TS-EH'16 R572N6 1 M 5 Q SGT-TS-EH'17 R572N6 1 M Q Pane A-37 STANDBY GAS TREATMENT SYSTEM TABLE A Equi ment on Preferred Safe Shutdown Paths Accident EPN Rad Use Infor mat ion Status Comments Zone Code Expd Read Qual SGT-TS-EH1A18 R572N6 1 M 5 Q SGT- TS-EH 1 A 19 R572N6 1 M 5 Q SGT-TS-EH 1A21 R572N6 1 M 5 Q SGT-TS-EH1A218 R572N6 1 M 5 Q SGT-TS-EH1A21 1 R572N6 1 M 5 Q SGT-TS-EH1A212 R572N6 1 M 5 Q SGT-TS-EH1A213 R572N6 1 M 5 Q SGT-TS-EH1A214 R572N6 1 M 5 Q SGT-TS-EH1A215 R572N6 1 M 5 Q SGT-TS-EH1A216 R572N6 1 M 5 Q SGT-TS-EH1AP 17 R57PN6 1 M 5 Q SGT-TS-EHI A218 R572N6 1 M 5 Q SGT-TS-EH1A2P R572N6 1 M 5 Q SGT-TS-EH1A23 R572N6 1 M 5 Q SGT-TS-EH1A24 R572N6 1 M 5 Q SGT-TS-EH1A25 R572N6 1 M 5 Q SGT-TS-EH1A26 R572N6 1 M 5 Q SGT-TS-EH1A27 R572N6 1 M 5 Q SGT-TS-EH1A28 R572N6 1 M 5 Q SGT-TS-EHiA29 R572N6 1 M 5 Q Page A-38 STANDBY LIQUID CONTROL SYSTEM TABLE A E ui ment on Preferred Safe Shutdown Paths Accident Rad Use Informat ion Status Comments Zone Code Ex ad Read Qual SLC-V-4A R548C M SLC-V-4B R548C M Page A-39 SUPPRESSSION POOL TEMP MONITORING SYSTEM TABLE A Ecui ment on Preferred Safe Shutdown Paths Accident Rad Use Info'at ion St at us Comment s Zone Code Expd Reed Qual SPTM- TE-1 1 C448 1 PQR 3 PQR SPTN-TE-13 C448 1 PQR 3 PQR SPTM-TE-15 C448 1 PQR 3 PQR SPTN-TE-1A C466 1 PQR 3 PQR SPTM-TE-RA C466 1 PQR 3 PQR SPTM-TE-3A C466 1 PQR 3 PQR SPTM-TE-4A C466 1 PQR 3 PQR S>TN-TE-5A C466 1 PQR 3 PQR SPTM-TE-6A C466 1 PQR 3 PQR SPTN-TE-7A C466 1 PQR 3 PQR SPTN-TE-8A C466 1 PQR 3 PQR SPTN-TE-9 C448 1 PQR 3 PQR Page A-48 SOURCE RANGE MONITOR SYSTEM TABLE A Equi ment on Preferred Safe Shutdown Paths Accident EPN Rad Use Is<format ion Status Comments Zone Code Exvd Read Qual SRA-CONN-84 C 2 PQR 1 PQR Qt SRN-DET-iD C 1 PQR 1 PQR JIO 018 S RN-EANP-1 D R581K 1 D 1 Qf Page A-41 STANDBY SERVICE WATER SYSTEN TABLE A Equi ment on Preferred Safe Shutdown Paths Accident EPN Rad Use Informat ion St at us Comment s Zone Code Exnd Reed Qual SW-MO-187A R548L 1 NONE 3 J JIO ¹19 SW-MO-188A R5'48L 1 NONE 3 J JIO ¹19 SW-MO-24A R441G 1 NONE 3 Q SW-NO-44 R44f B 1 NONE 3 Q SW-MO-54 R441C 1 NONE 3 Q SW-NO-75A R522K 1 D 3 D+ Q SW-PS-1814 R548E 1 NONE 3 J JIO ¹28 SW-V-281 R548E 1 NONE 3 J JIO ¹21 SW-V-284 R548E 1 NONE 3 J JIO ¹21 SW-V-212 R548E 1 NONE 3 J JIO ¹21 SW-V-213 R548E 1 NONE 3 J JIO ¹21 SW-V-848 R471D D 3 D+ Q SW-V-842 R471D 1 D 3 D+ Q SW-V-844 R471 D 1 D 3 D+ Q SW-V-846 R471D 1 D 3 D+ Q Page A-42 TRAVERSING IN-CORE PROBE SYSTEM TABLE A E ui rient on Preferred Safe Shutdown Paths Accident EPN Rad Use Inforrnat ion St at us Cornrnent s Zone Code Exnd Reod Qual TIP-V-1 R581P EF 1 J JIO ¹22 TIP-V-2 R581P 1 EF 1 J JIO ¹22 TIP-V-3 R581P 1 EF J Jl'0 ¹22 TIP-V-4 R581P 1 EF 1 J JIO ¹22 T IP-V-5 R581P 1 EF 1 J JIO ¹22 TABLE B ACCIDENT LEGEND A = HELB - RCIC H ~ HELB = RWCU 0 = HELB - RFW B -"HELB - RCIC I = HELB - RWCU P = LOCA - RRC C = HELB - RCIC J ~ HELB - RWCU Q = LOCA - MSL 0= HELB - AS K = HELB - RWCU R LOCA - SMALL E = HELB - RCIC L = HELB - RWCU S Control Rod Drop F = HELB - RWCU M ~ HELB - AS + = Submerged in Suppression Pool G = HELB - RWWU N = HELB -MS *: This component should be qualified to the conditions inside the reactor building due to LOCA breaks inside the primary containment. i 1: P,Q,R 7: N~ P~ Q~R 2: N,O,P,Q,R 8: O,P,Q,R 3: A through R 9: A,B,C,E 4: F through L 10: N 5: A, B,C, EthroughL, N,O ll: N, 0 6:A,B,C,E,P,Q,R 12: S TABLE B STATUS/COMMENT LEGEND Q = Qualification Documentation Complete Q' Equipment in test or being replaced, or review of final documentation being comp'leted. Qualification documentation to be completed prior to fuel load. J = Justification for Interim Operation provided. Page 8- 1 CONTAINMENT ATMOSPHERE CONTROL SYSTEM TABLE 8 Equi ment on Alternate Safe Shutdown Paths Accident Rad Use Inforraat ion Status Comments Zone Code Exnd Read Qual CAC-CNTR-18 R572F 1 M Q~ CAC-EHC-18 R572F 1 Q CAC-EHO-FCV/18 R548Q K Qf CAC-EHO-FCV/28 R548H 1 L Qf CAC-EHO-FCV/38 R471M NO Qt CAC-EHO-FCV/48 R471 8 1 D QS CAC-EHO-FCV/58 R572F 1 M Q1 CAC-EHO-FCV/68 R572F M Ql CAC-EHO-TCV/48 R572F 1 M Q~ CAC-EHO-V/18 R572F 1 M Ql CAC-EHO-V/28 R572F 1 M Qt CAC-EHO-V/38 R572F M Qt CAC-FT-58 R572F 2 M Q CAC-FT-68 R572F 1 M Q CAC-FT-78 R572F 1 M Q CAC-LT-18 R572F Q CAC-M-FN/18 R572F 1 Q CAC-PT-18 R572F 2 M Q CAC-PT-688 R572F 1 M Q CAC-RLY-48/CR1 R471D 1 D Q CAC-RLY-48/CR2 R471D 1 D Q CAC-RLY-FCV1888 R471D 2 D Q CAC-RLY-FCV2888 R471D 2 D Q CAC-RLY-FCVZB88 R471D D Q CAC-RLY-FCV4888 R471D 2 D Q CAC-TE-28 R572F 1 M Q CAC-TE-38 R572F M Q CAC-TE-48 R572F Q CAC-TE-58 R572F 1 Q CAC-TE-68 R572F 1 Q CAC-TE-78 R572F 1 Q Page B- P CONTROL AIR SYSTEM TABLE B E ui ment on Alternate Safe Shutdown Paths Accident EPN Rad Use Inforrnat ion St at us Cornrnent s Zone Code Exod Read Qual CAS-V-453 R471D 1 D Page B- 3 CONTAINMENT PURSE EXHAUST SYSTEM TABLE B Eoui ment on Al ernate Safe Shutdown Paths Accident Rad Use Infor mat ion Status Comments Zo7 ie Code Exnd Read Qua'. CEP-POS-V/2A R548Q K CEP-POS-V/RB R548Q K CEP-POS-V/4A R471 J D C~P-POS-V/4B R471 J D CEP-SPV-RA R548P JM CEP-SPV-RB R548P JM CEP-SPY-4A R581K D CEP-SPV-4B R581K D Pane B- 4 CONTAINMENT INSTRUMENT AIR SYSTEM TABLE B E ui ment on Alternate Baf-'e ShutCown Paths Accident Rad Use Inl'or mat ion St at us Comment s Zone Code Expd Rend Qual C I A-NlO-38B R522H D 3 D+ CIA-PROS-iB R548P JM 3 JM+ C I A-PS-21 B R548P JM 3 JM% CIA-PS-22B R548P JM 3 JM+ CIA-PT-21B R548P JM 3 JM+ C I A-RLY-21B R548P JM 3 JM% CIA-RLY-22B R548P JM 3 JM+ CIA-SPV-18B R441 D NONE 3 9 CIA-SPV-1 1B R441D NONE 3 Q CI A-BPV-12B R441D NONE 3 9 CIA-SPV-13B R441D NONE 3 9 CI A-SPV-14B R441D NONE 3 9 CIA-SPV-15B R441D NONCE 3 9 CI A-SPV-16B R441D NONE 3 9 CIA-SPV-17B R44' NONE 3 Q CIA-BPV-18B R441D NONE 3 9 CIA-SPV-19B R441D NONE 3 9 C A-SPV-1B R441 D NONE 3 9 CIA-SPV-2B R441D NONE 3 Q CIA-SPV-3B R441D NONE 3 9 CIA-SPV-4B R441D NONE 3 Q CIA-SPV-5B R441D NONE 3 Q C I A-SPV-6B R44' NONE 3 9 CI A-SPV-7B R441 D NONE 3 9 C I A-SPV-8B R441D NONE 3 Q C I A-SPV-9B R441 D NONE 3 9 CIA-TDS-1B R548P JM 3 Jg+ e.- Page B- 5 CONTAINMENT MQNITQRINB SYSTEM TABLE B Equi ment Gn Alternate Sa e Shutdaxn Paths Accident Rad Use Inl. Grrnat i Gn St at us Cornrnent s ZGne Cade Exvd Reed Qual CMS-AY-2 R548F 1 NONE CMS-AY-4 q R548F 1 NONE 1 CMS-H-SR'4/TS R548F NONE 1 CMS-H-SR14/T6 R548F 1 NONE CMS-H-SR'4/T7 R548F NONE 1 CMS-H-SR14/T8 R548c 1 NONE 1 CMS-LE-4A C435 1 PQR+ 3 PQR+ CMS-LE-4B C588 1 PQR 3 PQR CMS-LE-5A C435 a PQR+ 3 PQR+ CMS-LE-5B C588 1 PQR 3 PQR CMS-LT-1 R441J 1 C 3 C+ CMS-PT-2 R548P JM Q CMS-PT-4 R581K 1 D 1 Q CMS-PT-6 R548P 1 JM Q CMS-PT-8 R548P 1 JM Qf CMS-RE-27F C516 1 PQR PQR Q~ CMS-RMS-HTP88 R548F NONE 1 CMS-TE-12 C528 PQR PQR CMS-TE-14 C544 1 PQR 1 PQR CMS-TE-24 C596 1 PQR 1 PQR CMS-TE-2S C596 PQR PQR CMS-TE-28 C568 1 PQR 1 PQR CMS-TE-3 C517 PQR PQR CMS-TE-31 C568 1 PQR PQR CMS-TE-4 CS58 PQR 1 PQR CMS-TE-7 C584 1 PQR 1 PQR CMS-TE-9 C547 1 PQR 1 PQR CMS-TS-8A R548F 1 NONE 1 CMS-TS-8B R548F 1 NONE CMS-TS-8C R548F 1 NONE 1 CMS-TS-8D R548F 1 NONE 1 CMS-TS-9A R548F NONE 1 CMS-TS-9B R548F a NONE a CMS-TS-9C R548F 1 NONE 1 CMS-TS-9D R548F 1 NONE 1 Page B- 6 CONTAINMENT RECIRCULATION A I R SYSTEM TABLE B ~cCui ment on Alternate Sa e Shutdown Paths Accident Rad Use In forrnat i on St at us Cornrnent s Zone Code Exnd Read Qual CRA-M-FN/3B C5.6 PQR i PQR CRA-M-FN/3C C536 PQR PQR CRA-M-FN/4B C572 PQR i PQR CRA-M-FN/5B C564 PQR i PQR CRA-M-FN/5D C564 PQR i PQR 0 Pane B- 7 CONTROL ROD DRIVE SYSTEM TABLE B Equi ment on Alternate Sa e Shutdown Paths Accident Rad Use Infor mat ion St at us Cornrnent s Zone Code Exod Reed Qual CRD-SPV-liQB R522C 2 DI e Page B- 8 CONTAINMENT PURGE SUPPLY SYSTEM TABLE B Ecui ment on Alternate Sa~e Shutdown Paths Accident Rad Use Inforrnat i on St at us Cornrnent s Zone Code Exnd Reod Qual CSP-DPT-4 R581K 1 D 1 CSP-DPT-5 R581F 1 D CSP-DPT-6 R581F 1 D 1 CBP-POS-U/2 R581 I 1 NONE 1 CSP-POS-V/4 R471D 1 D 1 CSP-POS-U/5 R471D 1 D 1 CSP-POS-V/6 R471J 1 D 1 CSP POS V/9 R471B 1 D 1 CBP-RLY-ARCSPV5 R581F 1 D 1 CBP-RLY-ARCSPV6 R581K 1 D 1 CSP-RLY-ARCBPV9 R581F 1 D 1 CSP-RLY-CR4 R581K D 1 CSP-RLY-CR5 R581F D 1 CBP-RLY-CR6 R581F 1 D 1 CSP-SPV-2 R581K 1 D 1 CSP-SPV-4 R581F 1 D CSP-SPV-5 R581F 1 D 1 CSP-SPV-6 R581K 1 D 1 CSP-SPV-9 R581F D 1 CBP-V-93 R471D 1 D 1 CBP-V-96 R581I 1 NONE Page B- 9 ELECTRICAL SYSTEM TABLE B Eoui ment on Alternate SaÃe Shutdown Paths Accident Rad Use Inforrrration St at us Cornrnent s Zone . Code Ex md Read Qual E-CONN-X 1 88A/81 C587 2 PQR 1 PQR E-CONN-X 188A/82 C587 2 PQR 1 PQR E-CONN-X 188B/81 C587 2 PQR PQR E-CONN-X 188B/82 C587 2 PQR 1 PQR E-CONN-X 188C/81 C511 2 PQR 1 PQR E-CONN-X 188C/82 C511 2 PQR 1 PQR E-CONN-X 182B/81 C534 2 PQR 3 PQR E-CONN-X 182B/82 C534 2 PQR 3 PQR Page B-18 EQUIPMENT DRAINS RADIOACTIVE SYSTEM TABLE B Eoui ment an Alternate Sa, e Shutdawn Paths Accident Informat ion Rad Zone Use Coce E><nd Read Qua't at us Comment s EDR-POB-V/19 R441 C 1 NONE 1 ED R-BPV-19 R422E 1 NONE Page B-ii FLOOR DRAIN RADIOACTIVE SYSTEM TABLE B E ui ment on Alternate Safe Shutdown Paths Accident Rad Use In for mat i on St at us Corrrrnent s, Zone Code Expd Read Qual FDR-LS-42 R422I 1 NONE 1 FDR-LS-43 R422M 1 C 1 FDR-LS-44 R422L 1 AB 1 FDR-POS-V/3 R441C 1 NONE 1 FDR-SPV-3 R422E 1 NONE 1 Page B-12 FUEL POOL COOLING SYSTEM TABLE B Equi ment on Alternate Sa. e Shutdown Paths Accident EPN Rad Use Infor rrrat ion St at us Comment s Zone Code Exvd Read Qual F>C-MO-156 R4418 1 NONE 3 Q FPC-TE-7 R572K 1 NONE 3 .Page B-13 RRC HYDRAULIC CONTROL TABLE B ~Eaui went on Alter na e Safe Shutdown patne Accident EPN Rad Use Inforrnat i on St at us Comment s Zone Code Exvd Reed Qual HY-V-33A R522C 1 DI Q HY-V-33B R522J 1 D Q HY-V-34A R522C 1 DI Q HY-V-34B R522J D Q HY-V-35A R522C 1 DI Q HY-V-35B R522J 1 D Q HY-V-36A R522C 1 DI Q HY-V-36B R522J D Q Page B-14 LEAK DETECTION SYSTEN TABLE B Ecui ment on Alternate Safe Shutdown Paths Accident Rad Use Inforrnat i on St at us Cornrnent s Zone Code Expd Reed Qual I LD-TE-18A R441G 1 NONE 3 9 LD-TE-18B R441F NONE 3 9 LD-TE-18C R441G 1 NONE 3 9 LD-TE-18D R441F 1 NONE 3 Q LD-TE-1 B R522F 1 MI 4 HI 9 LD-TE-1D R522F 1 HI 4 HI 9 LD-TE-1F R548B 1 J J 9 LD-TE-27A R422I 1 NONE 3 Q LD-TE-P7B R422J 1 NONE 3 9 LD-TE-27C R422I NONE 3 9 LD-TE-27D R422J 1 NONE 3 9 LD-TE-28A R441F 1 NONE 3 9 LD-TE-28B R441G 1 NONE 3 9 LD-TE-28C R441F 1 NONE 3 Q LD-TE-28D R441G 1 NONE 3 Q LD-TE-29B R5810 NO 11 NO 9 LD-TE-29D R5810 1 NO 11 NO Q LD-TE-2B RSP2F 1 HI 4 HI 9 LD-TE-PD R522F 1 HI 4 HI Q LD-TE-2F R548B 1 J 4 J 9 LD-TE-38B R5220 11 NONE 9 LD-TE-38D LD-TE-31B R52P.O R5810 1 1 G G NO ii 11 NONE NO 9 9 LD-TE-31D R5810 1 NO 11 NO 9 LD-TE- B R522F 1 HI 4 HI Q LD-TE-3D R522F 1 HI 4 HI 9 LD-TE-3F R548B 1 J 4 J 9 LD-TE-4B R441 I 1 AB 9 AB 9 LD-TE-5B R422L 1 AB 9 AB 9 LD-TE-6B R441 I 1 AB 9 AB Q Page B-15 LOW PRESSURE CORE SPRAY SYSTEM TABLE B E ui ment on Alternate Safe Shutdown Paths Accident Rad Use Info'at ion St at us Cornrnent s Zone Code Ex ad Reed Qual LPCS-F I S-4 R471A D is 5 LPCS-FT-3 R471A D 1,5 LPCS-8-P/1 R422C NONE 1, 5 LPCS-P IS-1 R471A D 1,5 LPCS-PS-9 R471A D 1,5 Page B-16 MAIN STEAM SYSTEM TABLE B Ea ui ment on Alternate Safe Shutdown Paths Accident Rad Use In forrnat i on St at us Cornrnent s Zone Code Exnd Read Qual MS-CONN-V28A/Jl R58' 2 NO 2 NO>> Q MS-CONN-V28A/J2 R5810 2 NO NO>> Q MS-CONN-V28A/J3 R58' 2 NO 2 NO>> Q MS-CONN-V28B/Jl R5810 2 NO 2 NO>> Q MS-CONN-V28B/J2 R5810 2 NO 2 NO>> Q YS-CONN-V28B/J3 R5810 2 NO 2 NO>> Q MS-CONN-V28C/Jl R5810 2 NO P NO>> Q MS-CONN-V28C/J2 R5810 2 NO 2 NO>> Q MS-CONN-V28C/J3 R5810 2 NO 2 NO>> Q YS-CONN-V28D/Jl R5810 2 NO 2 NO>> Q MS-CONN-V28D/J2 R5810 2 NO 2 NO>> Q YS-CONN-V28D/J3 R5810 2 NO 2 NO>> Q MS-DPIS-11C R471B D 18 NONE Q MS-DP IS-1 1D R581B 1 D 18 NONE Q MS-DPIS-818C R471B 1 D 18 NONE Q MS-DP I S-818D R581B D 18 NONE Q MS-DPIS-8C R471B D 18 NONE Q MS-'DP I S-BD R581B D 18 NONE Q MS-DPIS-'9C R471B D 18 NONE Q YS-DP IS-9D R581B 1 D 18 NONE Q MS-L I S-24C R522C 1 DI 1 Q YS-L IS-24D R522P 1 D Q MS-LIS-37B R522H 1 D 3 D>> Q MS-L I S-37D R522H 1 D 3 D>> Q MS-L I S-38B R522H 1 D 3 D>> Q MS-L ITS-26C R522C 1 DI 1 MS-L I TS-26D RS22H 1 D 3 D>> MS-L I TS-44B R471D 1 D 1 MS-MO-16 C584 1 PQR 2 PQR Q MS-POS-V/28A/1 R5810 1 NO 2 NO>> Qt MS-POS-V/28A/2 R5810 1 NO 2 NO>> MS-POS-V/28A/3 R5810 1 NO 2 NO>> 2 NO>> Q'l MS-POS-V/28B/ 1 R5810 1 NO Q'l MS-POS-V/28B/2 R5810 NO 2 NO>> MS-POS-V/28B/3 RS810 2 NO>> Q1 MS-POS-V/28C/1 R5810 1 NO 2 NO>> Qf YS-POS-V/28C/2 R5810 1 NO 2 NO>> Q1 MS-POS-V/28C/3 R5810 1 NO 2 NO>> QT MS-POS-V/28D/1 R5810 NO NO>> Qf 2 Q'l MS-POS-V/28D/2 R5810 1 NO NO>> MS-POS-V/28D/3 R5810 1 NO 2 NO>> Qs MS-PS-28C R5220 1 DI 1 Q1 MS-PS-28D R522P D Q% MS-PS-23C R522C DI 1 Qt MS-PS-23D R522P 1 D 1 Q~ Page B-17 ,MAIN STEAM SYSTEM TABLE B Equi ment on A3.ternate Safe Shutdown Paths Accident Rad Use In for mat i on St at us Comment s Zone Code Exod Reed Qual MB-PS-48B R522H D 3 D+ MS-PS-48D R522H 1 D 3 D+ NS-PT-51B R522H D 3 D+ MS-RE-3C R5810 1 NOS 12 S NB-RE-3D R5810 1 NOS 12 S MS-SPV-28A2 R5810 1 NQ 2 NQ+ Q'1 MS-SPV-28A3 R5810 1 NO 2 NQ+ Qt MS-SPV-2812 R5810 1 NO 2 NO+ 'Q1 NS-SPV-28B3 R5810 1 NO 2 NO+ Q'1 MS-SPV-28C2 R5810 1 NO 2 NO+ Qt NS-SPV-28C3 R5810 1 NQ 2 NQ+ Qt MS-SPV-28D2 R5810 1 NO 2 NO+ gt NS-SPV-28D3 R5810 1 NO 2 NQ+ Qt MS-SPV-3DB C547 1 PQR 3 PQR Q'1 MS-SPV-4AB C547 1 PQR 3 PQR Q't MS-SPV-4BB C547 1 PQR 3 PQR MS-SPV-4CB C547 PQR 3 PQR Qt MS-SPV-4DB C547 1 PQR 3 PQR Q'1 MS-SPV-5BB C547 1 PQR 3 PQR Qt MS-SPV-5CB C547 1 PQR 3 PQR Q1 Page B-18 MAZN STEAM LEAKAGE CONTROL SYSTEM TABLE B Eeui ment on Alternate Safe Shutdown Accident Rad Use T.nf orrnat ion St at us Cornrnent s Zone Code ExDd Read Qual NSLC-FT-3A R471J D 1,5 NSLC-FT-39 R471J D ',5 Ã St C F'T 3C R471J D 1,5 MSLC-FT-3D R471J D 1,5 MSLC-H-A R471J D 1,5 MSLC-H-B R471 J D 1,~ NSLC-H-C R471 J D 1,5 MSLC-H-D R471J D 1,5 MSLC-N-FN/ 1 R471J D 1,5 MSLC-MQ- 1 A R471J D 1,5 NSLC-MQ-1B R471J D 1,5 MSLC-MQ-1 C R471J D 1,5 MSLC-MO-1D R471J D 1,5 NSLC-MQ-2A R5810 NO 1,5 NQ+ MSLC-NO-2B R5810 NO igS NO~ NSLC-NO-2C R5810 NO 1,5 NQ+ MBLC-MO-2D R5810 NO 1,5 MSLC-PB-78A R522P D 1,5 MSLC-PB-788 R522P D 1,5 MSLC-PB-78C R522P D 1,5 NSLC-PB-78D R522P D 1,S MSLC-PS-7A R522P D 1 5 NSLC-PS-7B RSP2P D 1,5 MSLC-PS-7C R522P D 1,5 MSLC-PS-7D R522P D 1,5 MSLC-PB-8A R522P D 1,5 NSLC-PB-8B R522P D 1,5 MSLC-PS-8C R522P D 1,5 NSLC-PS-8D R522P D 1,5 MSLC-PT-18A R522P D 1,5 NSLC-PT-18B R522P D 1,5 MSLC-PT-18C R522P D 1,5 MSLC-PT-18D R522P D MSLC-PT-12A R522P D MSLC-PT-12B R522P D 1,5 MSLC-PT-12C R522P D MS'-PT-1 c.D R522P D 1,5 MSLC-RLY-CR/18 R522P D 1,5 NSLC-RLY-CR/11 R522P D 1,5 MSLC-RLY-CR/12 R522P D 1,5 MS'-RLY-CR/13 R5c.2P D 1,5 MS'-RLY-CR/lA R5~2> D l,~ MBLC-RLY-CR/1B RSc.2P D 1,5 MSLC-R'-CR/lC R522P D 1,5 >S'-RLY-CR/' RS22P D 1,5 l Page B-19 NAEN STEAN LEAKAGE CONTROL SYSTEN TABLE B Eoui ment on Alternate Sa~e Shutdown Paths Accident Rad Use Inf'or mat ion St at us Comment s Zone Code Ex ad Read Qual NSLC-RLY-CR/5A1 R522P 1 D 1,5 Q NSLC-RLY-CR/5A2 R522P 1 D 1,5 Q NSLC-RLY-CR/5Bi R522P 1 D 1,5 Q NSLC-R'-CR/5B2 R522P D 1~5 Q NSLC-RLY-CR/5C1 R522P 1 D 1,5 Q NSLC-RLY-CR/5C2 R522P 1 D 1,5 Q NSLC-RLY-CR/5D1 R522P 1 D 1,5 Q NSLC-RLY-CR/5D2 R522P D 1,5 Q NSLC-RLY-CR/6A1 R522P 1 D 1,5 Q NSLC-RLY-CR/6A2 R522P 1 D 1,5 Q NSLC-RLY-CR/6B1 R522P 1 D 1,5 Q NSLC-RLY-CR/6B2 R522P D 1,5 Q NSLC-RLY-CR/6C1 R522P 1 D 1,5 Q NSLC-RLY-CR/6C2 R522P 1 D 1,5 Q NSLC-RLY-CR/6D1 R522P 1 D 1,5 Q NSLC-RLY-CR/6D2 R522P D 1,5 Q NSLC-RLY-CR/8 R522P 1 D 1,5 Q NSLC-RLY-CR/9 R522P 1 D 1 5 Q NSLC-TD-TK/2A R522P D 1,5 NSLC-TD-TK/2B R522P D 1,5 NSLC-TD-TK/2C R522P 1 D 1,5 NSLC-TD-TK/2D R522P D 1,5 NSLC-TD-TK/3A R522P 1 D 1,5 NSLC-TD-TK/3B R522P i. D 1%5 NSLC-TD-TK/3C R522P 1 D NSLC-TD-TK/3D R522P J D NSLC-TD-TK/4A R522P 1 D 1,5 NSLC-TD-TK/4B R522P 1 D 1 f5 NSLC-TD-TK/4C R522P 1 D 1,5 NSLC-TD-TK/4D RS22P 1 D 1,5 NSLC-TE-18A R471J 1 D 1,5 NSLC-TE-18B R471J 1 D 1,5 NSLC-TE-18C R471J D iq5 NSLC-TE-18D R471J D 1,5 Page B-88 PROCESS INSTRUMENTATION SYSTEM TABLE 8 Eoui ment on Alternate Sa. e Shutdown Paths Accident EPN Rad Use Inforrnat ion St at us Cornrnent s Zone Code Exod Read Qual PSR-V-X73/ 1 CSP2 1 PQR 1 PQR Q PSR-V-X771/1 C581 1 PQR 1 PQR Q 't PSR-V-X77A/3 C581 1 PQR 1 PQR Q1 PS R-V-X88/1 CSRR 1 PQR 1 PQR Q PSR-V-X82/ 1 R471D 1 D 1 Q PS R-V-X82/7 R471D 1 D 1 Q PSR-V-X83/ 1 R471D 1 D 1 Q PS R-V-X84/ 1 R471A 1 D 1 Q PSR-V-X88/1 R441D 1 NONE Q Page B-Ri REACTOR BUILDING CLOSED COOLING MATER SYSTEM TABLE B E ui ment on Alternate Safe Shutdown Paths Accident EPN Rad Use Inforrnat ion St at us Cornrnent s Zone Code Exod Read Qual RCC-MO-184 R518S 1 EF 1 RCC-MO-1~~9 R548L 1 NONE 1 RCC-MO-131 R548L 1 NONE 1 RCC-MO-48 C517 1 PQR 1 PQR Page B-22 REACTOR CORE ISOLATION CQOLINB SYSTEM TABLE S Equi Ment on Alternate Safe Shutdown Paths Accident EPN Rad Use Inforrnat ion St at us Cornrrrent s Zone Code Exnd Reed Qual RC I C-CNTR-C882 R422L 1 AB 12 RCIC-DPIS-13B R471A D 9 NONE RC I C-DP S-7B R471A 1 D 9 NONE RC I C-EHO-C882 R422L AB 12 RCIC-FIS-2 R471D 2 D 1P. RC I C-FT-3 R47 1D 1 D 1P. RC I C-LS-18 R422L 1 AB 12 RC I C-LS-1 1 R422L 2 AB 12 RC I C-LS-15A R441J 1 C 12 RC I C-LS-15B R441J 1 C 1P. RCI C-LS-3 R422L 2 AB 12 RC I C-LS-4 R548N 2 NONE 12 RC I C-LS-5 R548J 2 NONE 12 RC I C-LS-6 R548B 2 J f2 RCIC-M-P/3 R422L 1 C 12 RC I C-NQ-1 R422L 1 AB 12 RC I C-NO-18 R422L 1 AB 12 RC I C-NO-22 R441I 1 AB 12 RCIC-MO-45 R422L 1 AB 12 RC I C-MO-46 R4P.2L 1 AB 12 RCIC-NO-59 R441I 1 AB 12 RC I C-NO-63 C556 1 PQR 6 PQR RCIC-MO-76 C556 1 PQR 6 PQR RCIC-MO-86 R471 I NONE 6 RCI C-POS-V/1 R422L 1 AB 12 RC I C-POS-V/25 R422L 2 AB 12 RC IC-POS-V/26 R422L AB 12 RC IC-POS-V/4 R422F P NONE 12 RCIC-POS-V/5 R422L 2 AB RC I C-PQS-V/54 R422L 2 AB 12 RCIC-PS-1 R422L 2 AB 12 RC I C-PS-12A R471D D 12 RCIC-PS-12B R471A 1 D 12 RC I C-PS-12C R471D 1 D 12 RC 1 C-PS-12D R471A 1 D 1P. RC I C-PS-28 R471D D 12 RC I C-PS-2 1 R471D D 12 RC I C-PS-22B R471A f D 9 NONE RCIC-PS-22D R471 A 1 D 9 NONE RC I C-PS-34 R422L AS 12 RC I C-PS-6 R471D 1 D 12 RC I C-PS-7 R422L 2 AS 12 RC I C-PS-SA R422K 1 NONr= 12 RCIC-PS-9B R422K 1 NONE 1P. RC I C-PT-4 R471D D 12 Page B-23 REACTOR CORE ISOLATION COOLINB SYSTEN TABLE B Equi ment on Alternate Sa e Shutdown Paths Accident Rad Use Inforrnat i on St at us Cornrnent s Zone Code Exod Read Qual RCIC-PT-5 R471D D 12. RC I C-PT-7 R471D D 12 RC I C-PT-8 R471D 2 D 12 RC IC-RLY-CR1 R422L 1 AB 12 RCIC-RLY-CR2 R422L AB 12 RCIC-SE-C882 R422L 1 AB 12 RC I C-SP V-25 R471 J 1 D 12 RCI C-SPU-26 R581K 1 D 12 I RC C-SPV-4 R471J 1 D 12 I RC C-SPV-5 R581H D 12 RC IC-SPV-54 R471 J D 12 RC IC-SS-1 R422L AB 12 I RC C-SV-C882 R422L 1 AB 12 I I RC C-T S-18A R422L 2 AB 12 RC C-TI S-18B I R422L 2 AB 12 0 Page B-24 REACTOR BUI'ING EXHAUST AIR(HVAC) SYSTEM TABLE B E ui ment on A'ternate Safe Shutdown Paths Accident EPN Rad Use Informat ion Status Comments Zone Code Exvd Reed Qual REA-DPT-1B 1 R572N 1 M iq5 REA-DPT-1 B2 R572C 1 N 1,5 REA-DPT-1B3 R572F 1 M 1,5 REA-DPT-1 B4 R572I 1 NONE 1, 5 REA-M-AD/8 R548K 1 LN 1 REA-POS-AD/8 R548K 1 LN REA-POS-V/2 R572N 1 N Q REA-RE-9C R572C 1 N 1,5 Qt REA-RE-9D R572C 1 M 1,5 Q1 REA-RLY-CR2 R548P 2 J,N 1 Q REA-SPV-V/2 R548P 1 J,M 1 Q Page B-25 REACTOR FEEDWATER SYSTENt TABLE B Ecui ment on Alternate Safe Shutdown Paths Accident EPN Rad Use nonfor mat ion Status Cornrnent s Zone Coce Exnd Read Qual RFW-SPV-32Ai R581F 1 D 8 RFW-SPV- 2A2 R581F 1 D 8 RFW-SPV-32B1 R581F 1 D 8 RFW-SPV-32B2 R581F 1 D 8 Page B-26 RESIDUAL HEAT REMOVAL SYSTEM TABLE B Equipment on Alter nate Safe Shutdown Patha Accident Rad Use Informat ion Statue Comment@ Zone Code Exod Read Qual RHR-DP IS-12B R581K 1 D 3 D+ RHR-F I S-18B R581K D 3 D+ 'RHR-F IS-18C R581K D 3 D+ RHR-FT-15B R581K 1 D 3 D+ RHR-FT-15C R581K 1 D 3 D% RHR-LS-18A R471E NQNE Q RHR-LS-18B R47!E 2 NONE 3 RHR-LS-18C R471E 2 NONE 3 Q RHR-LS-18D R471E 2 NONE 3 Q RHR-M-P/2B R422I 1 NONE 3 Q RHR-M-P/2C R429M 1 C 3 C+ Q RHR-N-P/3 R422M 1 C 3 C+ Q RHR-MO-23 R548H 1 L 1 Q RH R-MO-3B R548J 1 NONE 3 Q RHR-MO-47B R572I 1 NONE 3 Q RHR-MO-48B R548J 1 NONE 3 Q RHR-MO-49 R548J 1 NONE 3 Q RHR-NO-52B R572I NONE 3 Q RHR-MO-68B R548J NONE 3 Q RHR-MO-6B R422I 1 NQNE 3 Q RHR-MO-74B R57P I NONE 3 Q RHR-MO-87B R572 I 1 NONE 3 Q RHR-MQ-9 C511 1 PQR 3 PQR Q RHR-NO-94 R548J 1 NONE 3 Q RHR-MO-99B C581 1 PQR 1 PQR Q RHR-PS-16B R581K 1 D 3 D+ Q RHR-PS-16C R581K 1 D 3 D+ Q RHR-PS-19B R581K 1 D 3 D+ Q RHR-PS-19C R581K 1 D 3 D+ Q RHR-V-68A R548N 1 NONE 3 Q RHR-U-68B R548J 1 NONE 3 Q Page B-27 REACTOR BUILDING OUTSIDE AIR<HVAC) SYSTEM TABLE B E ui ment on Alternate Safe Shutdown Paths Accident Rad Use Info@mat ion Statue Comments Zone Code Exzd Reed Qual ROA-N-AD/19 R548K 1 LN 1 ROA-POS-V/2 R572F 1 N 1 RQA-RLY-CR288 R522H 2 D 1 RQA-SPV-18 R522D 2 NONE 3 RQA-SPV-14 R572F 2 N 3 BOA-SPV-17 R548C 2 N 1 ROA-SPV-288 R522H 1 D 1 Page B-28 REACTOR PROTECTION SYSTEN TABLE B Equi n>ent on Alternate Safe Shutdown Paths Accident EPN Rad Use Infor mat ion St at us Cornrnent s Zone Code Exod Read Qual RPS-PS-2C R522C 1 Dl i RPS-PS-2D R522P i, D i Page B-29 REACTOR BUILDING RETURN AIR(HVAC) SYSTEM TABLE B Equi ment on Alter nate Sa, e Shutdown Paths Accident Rad Use In forrnat i on St at us Cornrnent s Zone Code Exod Reed Qual RRA-M-FN/ 1 R441 J 1 C 3 RRA-M-FN/18 R5220 1 NONE 3 RRA-M-FN/14 R572H 1 NONE 3 RRA-M-FN/17 R548F 1 NONE 1 RRA-M-FN/19 R548L. 1 NONE 3 BRA-M-FN/28 R548L 1 NONE 3 RRA-M-FN/3 R441F 1 NONE 3 RRA-M-FN/6 R441I 1 AB 12 RRA-RMS-FN/1 R441 J 2 C 3 RRA-RMS-FN/3 R441F 2 NONE 3 RRA-RMS-FN/6 R441 I 2 AB 12 l Page B-38 REACTOR RECIRCULATION SYSTEN TABLE B E ui ment on Alternate Safe Shutdown Paths Accident EPN Rad Use In format ion St at us Comment s Zone I Code Ex ad Reed Qual RRC-PS-18B R471D 1 D 1 RRC-V-19 C586 1 PQR 1 Page B-31 REACTOR WATER CLEANUP SYSTEM TABLE B Eeui ment on Alternate Safe Shutdown Paths Accident EPN Rad Use Knformat ion St at us Comment s Zone Code Exad Read QuaL RWCU-MO-1 C548 PQR 1~ 4 PQR l I I l j I STANDBY GAS TREATMENT SYSTEM TABLE 8 E ui ment on Alternate Safe Shutdown Paths Accident Rad Use Enformat ion Status Comments Zone Code Exud Read Qual SGT-EHC-fBf R572N 5 Q SGT-EHC-182 R572N 5 Q SBT-EHO-18 1 R572N3 5 Qf SST-EHO-182 R572N3 5 Q' SBT-FS-281 R572N4 5 SST-FT-181 R572N2 5 Q SST-FT-182 RS72N2 M 5 Q SGT-N-FN/181 R572N3 M 5 Q SBT-M-FN/182 R572N3 5 Q SBT-NE-681 R572N 5 Q SST-NE-682 R572N 5 Q SGT-ME-683 R572N 1 N 5 Q SGT-NE-78 1 R572N M 5 Q SST-NE-782 R572N 1 5 Q SST-NE-783 R572N 1 N 5 Q SBT-MO-18 R572N4 1 5 Q SGT-NQ-38 1 R572N4 1 N 5 Q SBT-NO-382 R572N4 1 N 5 Q SBT-MO-48 1 R572N4 N 5 Q SST-MQ-482 R572N4 1 M 5 Q SGT-POS-V/28 RS72N 1 N 5 Q SBT-SPV-28 R572N 1 N 5 Q SGT-SPV-F4. R572N1 2 M 5 Q SGT-SPV-F5 R572Nl 2 N 5 Q SST-SPV-F6 R572Nl 2 N 5 Q SGT-TS-EH181 1 R572N6 1 N 5 Q SST-TS-EH18118 R572N6 1 N 5 Q SST-TS-EH181 1 1 R57RN6 1 M 5 Q SBT-'-EH18112 R572N6 1 M 5 Q SBT-TS-EH18f 13 R572N6 1 N 5 Q SST- TS-EH1 8 1 1 4 R57RN6 N 5 Q SGT-TS-EH18115 R572N6 1 M 5 Q SGT- S-EH181 16 R572N6 1 N 5 Q SST-TS-EH181 17 R57RN6 1 5 Q SST-TS-EH18118 R57RN6 1 5 Q C' SGT-TS-EH1812 R572N6 N Q SGT-TS-EH1813 R572N6 1 M Q SST-TS-EH1814 R572N6 5 Q SGT-TS-EH1 815 R572N6 1 N 5 Q SGT-TS-EH1816 R572N6 1 'N 5 Q SST-TS-EH1817 R572N6 1 N 5 Q SST-TS-EH1818 R572N6 1 M 5 Q SGT-TS-EH1819 R572N6 1 N 5 Q SBT-TS-EHf 821 R57RN6 1 5 Q SST-TS-EH1821 8 R572N6 1 N 5 Q ~ I Page B-33 STANDBY GAS TREATMENT SYSTEN TABLE B Ecui ment on Alternate Sa. e Shutdown Paths Accident EPN Rad Use Informat ion St at us Cornrnent s Zone Code Ex ad Read Qual SGT-TS-EH1B21 1 R572N6 M 5 SGT-TS-EH1B212 R572N6 1, M 5 Q SST-TS-EH1B213 R572N6 M 5 Q SGT-TS-EH1B214 R572N6 M 5 Q SST-TS-EH1B215 R572N6 1 M 5 Q SGT-TS-EH1B216 R572N6 1 M 5 Q SGT-TS-EH1B217 R572N6 1 N 5 Q SGT-TS-EH18218 R572N6 1 M 5 Q SGT- TS-EH 1 B22 R572N6 M 5 Q SGT-TS-EH1 B23 R572N6 1 M 5 Q SST-TS-EH1B24 R572N6 1 M 5 Q SGT-TS-EH1B25 R572N6 N 5 Q SGT-TS-EH1B26 R572N6 '1 N 5 Q SST-TS-EH1B27 R572N6 M 5 Q SGT- TS-EH 1B28 R572N6 1 N Q SST-TS-EH1B29 R572N6 1 M 5 Q y y t Page B-34 SUPPRESSSION POOL TEMP NONITORINB SYSTEM TABLE B Equi ment on Alter nate SaFe Shutdown Paths Accident EPN Rad Use Inf'orraat i on St'at us Cornrnent s Zone Code Exod Read Qual SPTN-TE-18 C448 PQR 3 PQR SPTN-TE'-,12 C448 PQR 3 PQR SPTM- TE-1 4 C448 PQR 3 PQR SPTN-TE-16 C448 PQR 3 PQR SPTN-TE-1B C466 PQR 3 PQR SPTN-TE-2B C466 PQR 3 PQR SPTM- TE-3B C466 PQR 3 PQR SPTM-TE-4B C466 PQR 3 PQR SPTN-TE-5B C466 PQR 3 PQR SPTN-TE-68 C466 PQR 3 PQR SPTN-TE-7B C466 PQR 3 PQR SPTN-TE-8B '466 PQR 3 PQR Page B-35 SOURCE RANGE MONITOR SYSTEM TABLE B E ui ment on Alternate Safe Shutdown Paths Accident Rad Use In f orrnat ion St at us Corrrnient s Zone Code Ex ad Reed Qual S RM-CONN-8 1 C 2 PQR 1 PQR SRM-CONN-82 C PQR 1 PQR SRM-CONN-83 C 2 PQR 1 PQR SRM-DET-1A C 1 PQR 1 PQR SRM-DET-1B C PQR 1 PQR SRM-DET-1C C 1 PQR 1 PQR SRM-EAMP-' R581B 1 D 1 Q~ SRM-EAMP-1B R581K 1 D Q~ SRM-EAMP-1C R581B 2 D 1 Qr Page B-36 STANDBY SERVICE WATER SYSTEM TABLE B E ui ment on Alternate Saf'e Shutdown Paths Ac+ident EPN Rad Use I nforrnat i on St at us Corrrrnent s Zone Code Exud Reed Qual SM-NO- 1 87B R548L 1 NONE 3 + SW-NO-188B R548L NONE 3 SM-NO-248 R441F 1 NONE 3 SM-NO-84C R441J 1 C 3 C+ SM-NO-758 RSBBH 1 D 3 b+ SM-PS-1815 R548F 1 NONE 3 SW-V-286 R548F 1 NONE 3 SW-V-F89 R548F 1 NONE 3 SW-V-818 R548F 1 NONE 3 SW-V-Rii R548F 1 NONE 3 SW-V-34 R441I 1 NONE 3 Page B-37 TRAVERSING IN-CORE PROBE SYSTEM TABLE B E ui ment on Alternate Safe Shutdown Paths Accident EPN Rad Use Informat ion St at us Con]ment s Zone Code Expd Rend Qual TIP-SV-1 R581P 1 EF IP-SV-c R581P 1 EF TIP-SV-3 R581P 1 EF 1 TIP-SV-4 R581P 1 EF 1 TIP-SV-5 R581P 1 EF 1 TIP-SV-6 R581 9 1 D 1 0 E UIPMENT JUSTIFICATIONS E IDENT JUSTIF ICATION INDEX JIO lb. EPN s CIA-PROG-1 A CIA-PS-22A CIA< L Y-21 A CIA <LY-22A CIA-TDS-1A NS-AY-1) 3 QlS-LE-3A, 3B QlS-RNS-HTP71 AS-TE-l, 2, 26, 27, 29, 30 10 CNS-TS-4A, 4B, 4C, 4D, 5A, 5B, 5C, 5D EDR-POS-V/20 12 FDR-POS-V/4 13 HPCS-F T-5 LPRM-Detector s 15 LPRN-Connectors 16 NS-RE-3A, 3B 17 ROA-POS-V/1 18 SRN NET-1 D 19 SW-N0-187A, 188A 20 SW- PS>>1014 21 SW-V-201, 204, 212, 21 3 22 TIP-V-l, 2, 3, 4. 5 0 E IPNENT JUSTIF ICATION fl 1.0 COMPONENT IDENT IF I CAT ION EPN: CIA-PROG-1 A
Description:
Containment Instrument Air System (CIA) 16 STEP Prograraner to N2 bottle SPV's Component Type: STEP Programmer Manufacturer/Model: Automatic Time and Control/1820 BLg20XX 2.0 ACCIDENT CONDIT IONS LOCA Accident Profile: 830 and f32 Use Code Operability Time: 4320 Hours Radiation Zone: R548G Zone Dose: 1.2xl04 Rads 3.0 COMPONENT SAFETY FUNCT ION The programmer controller sequences the opening of the solenoid valves (CIA-SPV-lA through 15A) for the nitrogen bottles in the Containment Instrument Air System. These bottles provide nitrogen to the seven Automatic lhpr essurization System {ADS) safety/relief valves. 4.0 UALIFICATION STARS 4.1 Summar of uglification Status The environmental qualification test program is scheduled and test results are not expected prior to fuel load. Therefore, this component is assumed not to be qualified and the following justification is provided.
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- 4. 2 Parameters Re uirin Justification Temperature, pressure, humidity and radiation dose.
5.0 JUSTIFICATION FOR INTERIM OPERATION The ADS is required to reduce reactor vessel pressure i f the High Pressure Core Spray {HPCS) is not maintaining the proper reactor vessel water level. This allows the Low Pressure Coolant Injection (LPCI) mode of the Residual Heat Removal {RHR) System and the Low Pressure Core Spray (LPCS) System to provide make-up water to the reactor core. ADS safety/relief valves can be supplied either by the Nitrogen Supply System controlled by CIA-PROG-lA, or by the charged air accumulator tank provided for each safety/relief valve (SRV). In the event that the program controller fails, the accumulator tank is capable of providing at least one SRV actuation. Once actuated, these SRV's will reduce the reactor vessel pressure so that the LPCS and LPCI systems can provide core cooling. If subsequent depressurization is required, the backup nitrogen supply can oe manually initiated from the remote nitrogen station for additional ADS actuation. Upon receipt of a low pressure indication in the control room from the nitrogen supply header pressure sensor, CIA-PT-21A, the operator will manually initiate charging of the CIA system from the remote nitrogen bottle station (CIA-TK-20A). This station, located in the corridor between the reactor building and the diesel generator building, is accessible under post-LOCA conditions. 6.0 CO NCLUSIO N Interim operation is justified on the following basis:
- 1. The SRV accumulator tanks provide alternate means to initiate reactor vessel depressurization to allow the LPCS and LPCI to function.
- 2. The operator can manually charge the CIA system for an indefinite period of time, from the remote nitr ogen station, actuations are needed.
if subsequent ADS
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EQUIPMENT JUSTIFICATION ¹2 1.0 I COMPONENT IDENT IF CAT ION EPN: CIA-PS-22A
Description:
ADS Pneumatic Supply Pressure Switch Component Type: Pressure Switch Manufacturer/Model: ASCO/SBllAKR 2.0 ACCIDENT 03NOIT IONS LOCA HE LB Accident Profile ¹30 and ¹32 Use Code: Operability Time: 4320 Hours Radiation Zone: R548G Zone Dose: 1.2xl04 Rads 3.0 COMPONENT SAFETY FUNCTION CIA-PS-22A is used to close valve CIA-V-39A upon sensing low pressure in the CIA compressed air system. Closure of CIA-V-39A would allow the Automatic Depressurization System (ADS) to switch its pneumatic supply to a ni trogen bottl e backup. 4.0 gUAI IF ICATION STAIUS 4.1 Suamar of uglification Status The environmental qualification test progr am is currently in progress and test results are not expected prior to fuel load; therefore, this component is assumed not to be qualified. The following justification is provided.
- 4. 2 Parameters Re uirin Dustification Temperature, pressure, humidity and radiation dose.
5.0 JUSTIFICATION FOR INTERIM OPERATION CIA-PS-22A is not required for transfer of the AOS pneumatic supply. If sufficient pressure is not maintained in the compressed air system and the nitrogen bottle backup system is initiated, transfer will occur automatically by closure of check valve CIA-V-41A. Failure of the switch will not degrade any safety-related functions for the following reasons:
- 1. Whether the switch fails closed or open, backup valve CIA-V-41A will accomplish transfer between the two pneumatic systems.
- 2. If the switch shorts to ground, no safety-related electrical control systems are affected.
Furthermore, CIA-PS-22A's input to the transfer program controller (CIA-PROG-lA) is also not required (See JIO fl for CIA-PROG-lA). 6.0 CO NCLUSIO N Interim operation is justified on the following basis:
- 1. CIA-PS-22A is not required to perform any active safety-related function.
- 2. This pressure switch cannot degrade any safety-related system, since its signal to the program controller (CIA-PROG-lA) and the program controller itself are not required for accomplishment of a safety function.
- 3. Pressure indication i's available from the qualified pressure transmitter, CIA-PT-21A, for the operator to take manual action to charge the air system from the remote N2 bottle station.
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E UIPMENT JUSTIFICATION 83 1.0 COMPONENT IDENTIFICATION EPN: CIA-RLY-21A
Description:
Containment Instrument Air System (CIA) Control Relay Component Type: Control Relay Manufacturer/Model: Struthers Dunn, Inc./219BXP 2.0 ACCIDENT CONDITIONS LOCA HELB Accident Profile: 830 and 832 Use Code: Operability Time: 4320 Hours Radiation Zone: R548G Zone Dose: 1.2x104 Rads 3.0 COMPONENT SAFETY FUNCTION This normally energized control relay activates the program controller CIA-PROG-lA. This control relay is activated by the low pressure switch CIA-PS-21A. 4.0 VALI F ICATION STATUS 4.1 Suranar of uglification Status The environmental qualification test program is currently in progress and test r esults are not expected prior to fuel load. Therefore, this component is assumed not to be qualified, and the following justification is provided. 4.2 Parameters Re uirin Justification Temperature, pressure, humidity and radiation dose.
5.0 JUSTIFICATION FOR INTERIM OPERATION The ADS is required to reduce reactor vessel pressure if the High Pressure Core Spray (HPCS) is not maintaining the proper reactor vessel water level. This allows the Low Pressure Coolant Injection (LPCI) mode of the Residual Heat Removal (RHR) System and the Low Pr essure Core Spray (LPCS) System to provide make-up water to the reactor core. ADS safety/relief valves can be actuated by either the Nitrogen Supply System, controlled by CIA-PROB-lA, or by the charged air accumulator tank provided for each safety/relief valve (SRV). In the event that the program controller fails, the accumulator tank is capable of providing at least one SRV actuation. Once actuated, these SRV's will r educe the reactor vessel pressure so that the LPCS and LPCI systems can provide core cooling. If subsequent depressurization is required, the backup nitrogen supply can be manually initiated from the remote nitrogen station for additional ADS actuation. Upon receipt of a low pressure indication in the control room from the nitrogen supply header pressure sensor, CIA-PT-21A, the. operator will manually initiate charging of the CIA system from the remote nitrogen bottle station (CIA-TK-20A). This station, located in the corr idor between the reactor building and the diesel generator building, is accessible under post-LOCA conditions.
6.0 CONCLUSION
Interim operation is justified on the following basis:
- 1. The SRV accumulator tanks provide alternate means to initiate reactor vessel depressurization to allow the LPCS and LPCI to function.
- 2. The operator can manually charge the CIA system for an indefinite period of time from a remote nitrogen station, actuations are needed.
if subsequent ADS
E UIPMENT JUSTIFICATION 84 1.0 COMPONENT IOENTIF ICATION EPN: CIA-RLY-22A
Description:
Containment Instrument Air System (CIA) Control Relay Component Type: Control Relay Manufacturer/Model: Agastat Relay Co./EGPI-002 2.0 ACCIOENT CONOITIONS LOCA HELB Accident Profile: 830 and 032 Use Code: Operability Time: 4320 Hours Radiation Zone: R548G Zone Oose: 1.2x104 Rads 3.0 COMPONENT SAFETY FUNCTION This normally energized control relay activates the programmer controller CIA-PROG-lA. This control relay is activated by the low pressure switch CIA-PS-22A when pressure decreases below 135 psi. 4.0 UALIFICAT ION STATUS 4.1 Summar of uglification Status The environmental qualification test documentation indicates this relay is qualified to 2x105 rads and 100'C for 42 days. The manufacturer recommends not installing this relay where it will be exposed to steam and indicates maximum relative humidity (RH) should be limited to 95K. Since this relay may be exposed to 100K RH, it is not considered qualified. Therefore the following justification is provided. 4.2 Parameters Re uirin Justification Humidity.
'.0 JUSTIFICATION The ADS FOR INTERIM OPERATION is required to reduce reactor vessel pressure if the High Pressure Core Spr ay (HPCS ) is not maintaining the proper reactor vessel water level. This allows the Low Pressure Collant Injection (LPCI) mode of the Residual Heat Removal (RN ) System and the Low Pressure Core Spray (LPCS) system to provide make-up water to the reactor core. ADS safety/relief valves can be actuated by either the Nitrogen Supply System controlled by CIA-PROG-lA, or by the charged air accumulator tank provided for each safety/relief valve (SRV). In the event that the program controller fails, the accumulator tank is capable of providing at least one SRV actuation. Once actuated, these SRV's will reduce the reactor vessel pressure so that the LPCS and LPCI systems can provide core cooling. If subsequent depressuriaction is r equired, the backup nitrogen supply can be manually initiated from the remote nitrogen station for additional ADS actuation. Upon receipt of a low pressure indication in the control from the introgen supply header pressure sensor, CIA-PT-21A, the operator will manually initiate charging of the containment instrument air system from the remote nitrogen bottle station (CIA-TK-20A). This station is located in the corridor between the reactor building and the diesel generator building, and is accessible under post-LOCA conditions.
E IPMENT JUSTIFICATION 85 'I 1.0 COMPONENT IDBlTIF I CAT ION EPN: CIA-TDS-1 A
Description:
3 Second Time Delay for Prograraner CIA-PROG-lA Component Type: Timer Manufacturer/Model: Agastat Relay Co./7022AE 2.0 ACCIDENT CONDIT IONS LOCA Accident Profile: 830 and 832 Use Code: Operability Time: 4320 Hour s Radiation Zone: R5488 Zone Dose: 1.2x104 Rads 3.0 COMPONENT SAFETY FUNCT ION This timer provides a time delay of 3 seconds for the program controller C IA-PROG-1A. 4.0 UALIFICATION STA1lJS 4.1 Summar of uglification Status The timer is exposed to maximum accident conditions of 178'F, 14.7 psia and 100K relative humidity. The timer is designed for a maximum temperature of 165'F and data for other environmental parameters is not available. Therefore, this component is not qualified environmentally, and the following justification is provided. 4.2 Parameters re uirin Justification Temperature, pressure, humidity and radiation dose.
i( 5.0 JUSTIF ICATION FOR INTERIM OPERATION 3 The Automatic Depressurization System (ADS) is required to reduce reactor vessel pressure if the High Pressure Core Spray (HPCS) is not maintaining the proper reactor vessel water level. This allows the Low Pressure Coolant Injection (LPCI) mode of the Residual Heat Removal (RHR) System and the Low Pressur e Core Spray (LPCS) System to provide make-up water to the reactor core. ADS safety/relief valves can be actuated by either the Nitrogen Supply System, controlled by CIA-PROG-lA, or by the charged air accumulator tank provided for each safety/relief valve (SRV). In the event that the rogram controller fails, the accumulator tank is capable of providing at east one SRV actuation. Once actuated, these SRV's will reduce the reactor vessel pressure so that the LPCS and LPCI systems can provide core cooling. If subsequent depressurization is required, the backup nitrogen supply can be manually initiated from the remote nitrogen station for additional ADS actuation. Upon receipt of a low pressure indication in the control room from the nitrogen supply header pressure sensor, CIA-PT-21A, the operator will manually initiate charging of the containment instrument air system from the remote nitrogen bottle station (CIA-E-20A). This station, located in the corridor between'he reactor building and the diesel generator building, is accessible under post-LOCA conditions. 6.0 CO NCLUSIO N Interim operation is justified on the following basis:
- 1. The SRV accumulator tanks provide alternate means to initiate reactor
'essel depressurization to allow the LPCS and LPCI to function.
- 2. The operator can manually charge the CIA system for an indefinite period of time, from a remote nitrogen station, if subsequent ADS actuations are needed.
Egl IPMENT XSTIF ICATION 0'6 1.0 COMPONENT IDENT IF I CAT ION EPN: CMS-AY-l, 3
Description:
Containment H2-02 Analyzer Component Type: Gas Analyzer Manufacturer/Model: Kaman/Beckman 7C (hydrogen) and 755 (oxygen) 2.0 ACCIDENT G)NDITIONS LOCA HE LB Accident Profile: N/A Use Code: 1 Operability Time: 4320 Hours Radiation Zone: R548E Zone Dose: 1 xl 06 Rads(estimate)
*Tmax = 107'F and Pmax = Atm.
3.0 COMPONENT SAFETY FUNCT ION The containment H2-02 analyzer is part of the containment monitoring system. Instrumentation to monitor containment hydrogen and oxygen is required in accordance with Regulatory Guide 1.97 to provide information to indicate the potential for breach of the primary containment. The HZ-02 analyzer 's function is to continuously monitor, record, and display in the control room, the containment hydrogen and oxygen concentrations. Mhen oxygen concentration reaches 4.4X by volume, a visual and audible alarm sounds in the control room. Operators then initiate at least one of the two 100K capacity hydrogen-oxygen recombiners. 4.0 UALIFICATION STA'1US 4.1 Summar of uglification Status The H2-02 analyzer is located in an isolated room serviced by guality Class 1 HVAC. Thus it is in a mild environment for temperature, pressure, and relative humidity.
The H2-02 analyzer is required to be qualified for radiation conditions resulting from a LOCA inside primary containment. The radiation dose for its location is currently being reanalyzed. Previous radiation dose calculations show a dose of 9.0xl03 rads based on shine from primary containment and nearby piping. It is estimated that the radiation dose will increase to approximately lxl06 rads when dose contributions from the analyzer's process stream are taken into account. Therefore, based on the dose estimate of 106 rads, the equipment is not qualified and a justification is included for interim operation while the analyzer is being qualified. 4.2 Parameter Re uirin Justification Radiation dose. 5.0 JUSTIFICATION FOR INTERIM OPERATION accordance with Technical Specification 3.6.6.3, primary containment will be inerted with nitrogen at the 25M power level. Prior to inerting, combustible gas control depends on the control of primary'containment 'n hydrogen concentration. Approved show analytical models, described in Section 6.2.5.3 of the FSAR, that the drywell hydrogen concentration will exceed the control limit of C by volume approximately 4.0 hours after a postulated LOCA the hydrogen recombiner is not in operation. if Operation of one qualified 100K capacity hydrogen-oxygen recombiner (CAC-HR-lA or CAC-HR-1B) will be initiated when the hydrogen concentration reaches approximately 3.5X by volume (2.75 hours after the postulated LOCA). This manual initiation of the recombiner, from the control room, conservatively limits the hydrogen concentration in containment to less than the 4.0X control limit. After primary containment is inerted with nitrogen at the 25K power level, combustible gas control depends on the control of primary containment oxygen concentration. The H2-02 analyzer alarms at 4.4X by volume containment oxygen concentration to alert operators. The oper ators then initiate the qualified, 100% capacity hydrogen-oxygen recombiner (CAC-N-lA). Initiating recombiner operation at 4.4X provides adequate margin to meet the recombiner operational limit (4.8X) and the oxygen flammability limit of 5X.
0 Section 6.2.5.3 of the FSAR shows that the containment wetwell oxygen " concentration will reach 4.4X by volume approximately six hours after a LOCA if the recombiner is not operating. At 12.5 hours after a LOCA, if the recombiner is not operating, the wetwell oxygen concentration will reach 4.85 by volume. This is the maximum oxygen concentration for control of the recombiner to limit the catalytic bed exit temperature to 1150'F. Therefore, the operators will initiate operation of the recombiner within 2.75 hours of the declaration that a LOCA condition exists. This manual initiation of the recombiner, from the control room, conservatively limits the oxygen concentration in containment to less than the 4.8% recombiner operational limit, and less than the 5.0f flammability limit by volume. 6.0 CO NCLUSIO N Until qualification of the H2-02 analyzer is documented, interim operation is justified with provision that the hydrogen-oxygen recombiner operation be initiated as described above, since the recombiner operation is independent of the analyzer operation. In the unlikely event that the H2-02 analyzer fails due to lack of data on radiation qualification, the requirement to initiate one of the qualified hydrogen-oxygen recombiners within 2.75 hours after a postulated LOCA will provide conservative assurance that the containment hydrogen control limit, or'he containment oxygen flammability limit, will not be reached.
EQJ IPMENT JUSTIFICATION 87 1.0 COMPONENT IDENT IF I CAT ION EPN: QlS-LE-3A, 3B
Description:
Suppr ession Pool Wide Range Level Monitoring Component Type: Pressure Transducer Manufacturer/Model: ElectroSyn/962 2.0 ACCIDENT CONDITIONS LOCA HELB Accident Profile: $ 1 and P2>> N/A Use Code: Operability Time: 4320 Hours Radiation Zone: C500* Zone Dose: 9.0 x 107 Rads*
- The following exceptions apply to (NS-LE-3A only:
Temperature: 200'F max Pressure: Dependent on suppression pool level Humidity: Submerged Radiation Zone: C435 Zone Dose: 3-7 x 106 rads 3.0 COMPONENT SAFETY FUNCT ION CMS-LE-3A and CMS-LE-38 provide verification of suppression pool water level and long-term surveillance in accordance with the guidelines of Regulatory Guide 1.97. Water level indication provides verification of the availability of water for the ECCS and a diverse indication of breach of the primary system (LOCA).
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NS-LE-3A measures pressure at the bottom of the suppression pool, and (NS-LE-38 measures pressure in the suppression chamber air space. The signal fr om each of these pressure transducers is sent to an electronics ackage in the control room which determines the suppression pool water evel which corresponds to the indicated differential pressure. 4.0 AL IF I CAT ION STATUS 4.1 Summar of uglification Status The 1 evel moni toring system i s being custom buil t for WNP-2. The qualification plan is currently being completed, and testing to verify qualification is scheduled. This system is not scheduled for installation prior to fuel load, but it will be demonstrated to be qualified prior to installation. This level monitoring system has been designed to function in normal and accident environments. Radiation and temperature resistant materials have been specified for the transducer seals, lead wires, cable (and shrink tubing), junction box gasket, and connections. The flexible and rigid conduits containing the transducers and cables will be water tight. The junction box will be above the water level and is designed to protect the connectors from any postulated. water spray. Based on this design, the level monitoring system is expected to perform its function. 4.2 Parameters Re uirin Justification Radiation dose, temperature, pr essure, relative humidity and submergence. 5.0 JUSTIFICATION FOR INTERIN OPERATION The conditions for which suppression pool water level would require wide range level instrumentation involve the long-term passive failure of an ECCS inside the primary containment coincident with a LOCA. Since this assumes a passive failure in addition to the accident scenario, which is not considered in JIO, GRS-LE-3A and SS-LE-3B are not required. The suppression pool water level can therefore be continuously monitored with narrow range level instrumentation. gualified suppression pool narrow range instrumentation is available at this time. (MS-LT-1 and CMS-LT-2 are qualified and provide suppression pool level indication over the range of 31' 27". The suppression pool normal operating level is 31' 2".
6.0 CO NCLUSIO N Interim operation is justified on the basis that because no single active or additional long-term passive failure was assumed for the suppression pool, its water level will remain within the narrow range instrumentation. In other words, water is not lost due to a break in the ECCS. The qualified suppression pool narrow range instrumentation system will provide adequate level monitoring until environmental testing can be completed on the wide range instrumentation.
E IPMENT JUSTIF ICATION 88 1.0 COMPONENT IDENT IF I CAT ION EPN: CMS-RMS-HTP71
Description:
Containment Monitoring System Control Switch for Heat Tracing for H2-02 Analyzer s Component Type: Remote Manual Control Switch Manufacturer/Model: General El ectric/CR2940 2.0 ACCIDENT CONDITIONS LOCA HE LB Accident Profile: N/A Use Code: Operability Time: 4320 Hours Radiation Zone: R548E Zone Dose: l.'Oxl06 Rads (estimate) 3.0 COMPONENT SAFETY FUNCT ION CMS-RMS-HTP71 is part of the containment hydrogen-oxygen monitoring system (H2-02 analyzer). This instrumentation is required in accordance with Regulatory Guide 1.97 to provide indication of the potential for'reach of primary containment. The H2-Og analyzer's function is to continuously monitor, record, and display >n the control room the containment hydrogen and oxygen concentrations. When the hydrogen or oxygen concentrations rise above setpoints, a visual and audible alarm sounds in the control room. Heat tracing is pr ovided to maintain the HZ-02 mixture in the vapor state to ensure accurate readings at the analyzer. This ensures that the lines will remain open allowing continuous sampling. Failure of (NS-RMS-HTP71 could affect the automatic operation of the heat trace system.
4.0 UALIFICATION STAlljS 4.1 Summar of uglification Status CMS-RMS-HTP71 is located in a mild environment for temper'ature, pressure, and relative humidity. Surveillance and post-maintenance testing will assure operability of the component under these mild environmental conditions. The control switch is required to be qualified for radiation conditions resulting from a LOCA. The radiation dose for its location is currently being reanalyzed. Previous radiation dose calculations show a dose of 9.Ox)03 rads based on shine from primary containment and nearby piping. It is estimated that the radiation dose will increase to approximately lxl06 rads when dose contributions from the H2-02 analyzer's process stream are included. Therefore, based on the dose estimate of 106 rads, the equipment is not qualified and a justification is included for interim operations while the control switch is being qualified. 4.2 Par ameters Re uirin Justification Radiation dose. 5.0 JUSTIFICATION FOR INTERIM OPERATION As discussed in JIO f6, the H2-02 analyzer will not be required to initiate recombiner operation since operators will start the hydrogen-oxygen recombiner within 2.75 hours after declaration of a LOCA. This action ensures that containment hydrogen and oxygen concentrations remain at safe levels. Since CMS-RMS-HTP71 is only required for H2-02 analyzer heat trace normal operation, this control switch performs no active safety function post-LOCA. The heat tracing is normally on at all times, with temperature switches controlling on-off for temperature control. There are three possible failure modes of this control switch:
- 1. Switch fails open - heaters will not heat. Since the analyzers will not be required to initiate recombiner operation, the heaters are not required to be on.
- 2. Switch fails closed - heater s will stay on. This is a normal condition.
- 3. Switch shorts - a fuse will blow in the motor control center power panel and disable heater operation.
Other equipment will not be degraded.
6.0 CON CLUS ION CMSRMS-HTP71 is justified for interim operation since it does not perform an active safety function post-LOCA. Since operator action to initiate the hydrogen recombiner system precludes the need for the Hp-Op analyzer, the failure of (ÃS-RMS-HTP71 wi11 not degrade the safety of the plant..
I IPMENT JUSTIF ICATION 09 E 1.0 COMPONENT IDENTIFICATION EPN: CMS-TE-l, 2, 26, 27, 29, 30
Description:
Containment Monitoring System Temperature Elements Component Type: Temperature El ement Manufacturer/Model: Hy-Ca 1 Engineering Co. /TC2370CCA250TT68JXH7 2.0 ACCIDENT CONDITIONS LOCA HELB Accident Prof ile: fl and 82 N/A Use Code: Operability Time: 4320 Hours Radiation Zone: Zone Dose: 7.0xl07 Rads 3.0 COMPONENT SAFETY FUNCTION The above listed temperature elements monitor the air temperature at various locations throughout containment. If the temperature in the upper drywell area rises above 340'F, or the temperature of the air in the inlets and outlets of the recirculating fans rise above 160'F, control room alarms are annunciated. A temperature recorder constantly records the temperature in the drywell. Regulatory Guide 1.97 requires that the drywell atmosphere temperature be monitored during and following an accident to aid the operator in taking cor rective actions to mitigate accident consequences. 4.0 AL IF I CAT ION STATUS 4.1 Summar of al ification Status Environmental testing is incomplete at this time. Until environmental testing is completed, the following justification for interim operation is provided.
- 4. 2 Parameters Re uirin Justification Temperature, pressure, humidity and radiation dose.
5.0 JUSTIFICATION FOR INTERIM OPERATION (MS-TE-1,2,26,27,29 and 30 provide temperature monitoring of various points in the drywell. All of these TE's are input to recorders through selector switches. For the temperature monitoring of various points in the drywell, there are alternate qualified TE's ((NS-'K-5,6,8,10,ll and 13) that input to the same r ecorders through the same selector switches. These temperature inputs are sufficient for the operators to monitor the primary containment temperature after an accident. The failure of OS-TE-1,2,16,27,29 and 30 could cause inaccurate temperature signals. However, this should have no consequence for primary containment temperature monitoring because the qualified temperature elements, SS-TE-5,6,8,10,ll and 13, can provide the correct information. The selector switches allow only the monitoring of qualified temperature elements. Identification of the qualified temperature elements will prevent the operators from being misled by possible failures of CMS-TE-l, 2, 26, 27, 29 or 30. 6.0 SNCLUS ION Interim operation is justified on the basis that the primary containment temperature monitoring function can be sufficiently provided by qualified temperature elements. A more complete temperatur e survey of the primary containment will be available after (ÃS-TE-1,2,26,27,29 and 30 are shown to be qualified.
E IPMENT JUSTIF ICATION 810 1.0 COMPONENT IDENT IF I CAT ION EPN: CMS-TS-4A, 4B, 4C, 40, 5A, 5B, 5C, 5D
Description:
H2-02 Analyzer Sample Rack Heat Trace Temperature Switch Component Type: Temperature Switch Manufacturer/Model: ASCO/SC11AR/gTllA4R HE LB 2.0 ACCIDENT CONDITIONS LOCA Accident Profile: N/A Use Code: Operability Time: 4320 Hours Radiation Zone: R548E Zone Dose: lxl06 Rads (estimate) 3.0 COMPONENT SAFETY FUNCT ION The temperature switches listed above are part of the containment hydrogen-oxygen monitoring system (Hp-02 analyzer). This instrumentation is required in accordance with Regulatory Guide 1.97 to provide indication of the potential for breach of primary containment. The H2-02 analyzer's function is to continuously monitor, record, and display in the control room the containment hydrogen and oxygen concentrations. When the hydrogen or oxygen concentrations rise above setpoints, a visual and audible alarm sounds in the control room. Heat tracing is provided to maintain the HZ-02 mixtur e in the vapor state to ensure accurate readings at the analyzer This ensures that the lines will remain open allowing continuous sampling. CMS-TS-4A, 4B, 4C, 40, 5A, 5B, 5C, 50 regulate the heat tracing operation. 4.0 UALIFICATION STATUS 4.1 Summar of uglification Status CMS-TS-4A, 4B, 4C, 40, 5A, 5B, 5C, 5D are located in a mild environment for temperature, pressure, and relative humidity.
The temperature switches are required to be qualified for radiation conditions resulting from a LOCA inside primary containment. The radiation dose for its location is currently being reanalyzed. Previous radiation dose calculations show a dose of 9.0xl03 rads based on shine from primary containment and nearby piping. It is estimated that the radiation dose will increase to approximately lxl06 rads when dose contributions from the H2-02 analyzer's process stream are included. Therefore, based on the dose estimate of 106 rads, the equipment is not qualified.and a justification is included for interim operation while the temperature switches are being qualified. 4.2 Parameters Re uirin Justification Radiation dose. 5.0 JUSTIFICATION FOR INTERIM OPERATION As discussed in JIO 86, the HZ-02 analyzer wil] not be required to initiate recombiner operation since operators will start the hydrogen-oxygen recombiners within 2.75 hours after declaration of a LOCA. This action ensures that containment hydrogen and oxygen concentrations remain at safe levels. Since CMS-TS-4A, 4B, 4C, 40, 5A, 5B, 5C, 50 are only required for Hg-02 analyzer operation, these temperature switches perform no active safety function post,-LOCA. There are three failure modes for these switches:
- 1. Switch fails open - heaters will not heat. Since the analyzers will not be required to initiate recombiner operation, the heaters are not required to be on.
- 2. Switch fails closed - heaters will stay on. This is a normal condition.
- 3. Switch shorts - a fuse will blow in the motor control center power panel and disable heater operation.
Other equipment will not be degraded. 6.0 GNCLUS ION CMS-TS-4A, 4B, 4C, 40, 5A, 5B, 5C, 50 are justified for interim operation since they do not perform an active safety function post-LOCA. Since operator action to initiate the hydrogen recombiner system precludes the need for the H2-02 analyzer, the failure of these temperature switches will not degrade the safety of the plant.
E UIPMENT JUSTIFICATION all 1.0 COMPONENT IDENTIFICATION EPN: EDR-POS-V/20
Description:
Position Indication Switch for EDR-V-20 Component Type: Position Switch Manufacturer/Model: NNCo Controls/SAI-133 2.0 ACCIDENT CONDITIONS LOCA HELB Accident Prof ile: N/A Use Code: Operability Time: 4320 Hours Radiation Zone: R441C Zone Dose: 1.4xl06 Rads 3.0 COMPONENT SAFETY FUNCTION EDR-POS-V/20 is the position switch for air -operated valve EDR-V-20. This valve, EDR-V-20, is part of the primary containment isolation boundary. Regulatory Guide 1.97 requires position indication of containment isolation valves. 4.0 UALIFICATION STATUS 4.1 Summar of ual ification Status For LOCA conditions, the position switch is exposed to maximum conditions of 128 F, 14.7 psia and 64K relative humidity. The position switch is designed for a maximum temperature of 160'F and a maximum humidity of 95K indicating the switch should function proper ly under these LOCA conditions. However, due to lack of documentation, the switch is scheduled for replacement with a documented, qualified switch. In the interim, a justification for operation is provided.
4.2 Parameters Re uirin Justification Temperature, pressure, humidity and radiation dose. 5.0 JUSTIFICATION FOR INTERIM OPERATION EDR-POS-V/20 provides the position indication for valve EDR-V-20. EDR-V-20 is normally 'open and is automatically closed by the following isolation signals during LOCA conditions:
- a. Isolation Signal A - Reactor vessel low water level (Trip 2)
- b. Isolation Signal F - High drywell pressure EDR-V-20 closes in approximately 15 seconds after receipt of an isolation signal and thus will close prior to experiencing any harsh environment.
EDR-POS-V/20 will indicate the correct valve position to verify containment isolation for approximately 12 minutes after the accident until the mild environment radiation dose (104 rads) has been exceeded. It would take approximately 54 hours into the accident to achieve the maximum temperature of 128'F and 350 hours to achieve the maximum humidity level of 64%. Therefore, the correct valve position can be verified for at least 12 minutes by the control room operator. The containment isolation function is assured in any event by the qualified valve (EDR-V-20) and its redundant, qualified back-up valve, EDR-V-19. Also, both EDR-V-20 and EDR-V-19 fail in the closed position, which provides further assurance of accomplishing containment isolation. In the event that EDR-POS-V/20 fails after the first 12 minutes, there are three possible consequences. The position switch could indicate open, closed, or neither. In each case, the correct position indication was available for at least 12 minutes, during which time the operator can verify containment isolation. After this time no further operator action is required because no credible event can occur to cause both isolation valves to open.
6.0 CONCLUSION
Interim operation is justified on the following basis: EDR-POS-V/20 will be operable for at least the first 12 minutes, during which time the operator can verify containment isolation.
- 2. Valve EDR-V-20 and its associated solenoid valve EOR-SPV-20 are qualified and will be automatically closed by either reactor vessel low water level or high drywel.l pressure isolation signals.
- 3. The operating time of the valve is approximately 15 seconds and will close prior to experiencing any harsh environment.
- 4. There is a redundant valve (EOR-V-19) in series with EDR-V-20. Valve EDR-V-19 and its associated solenoid valve EDR-SPV-19 are qualified,
. and provide a redundant means of containment isolation.
E IPMENT JUSTIFICATION 812 1.0 COMPONENT IDENT IF I CAT ION EPN: F DR-POS-V/4 Oescription: Position Indication Switch for FOR-V-4 Component Type: Position Switch Manufacturer/Model: NNCO Controls/SAI-133 2.0 ACCIOENT (QNOIT IONS LOCA Ac ci dent Prof i 1 e: N/A Use Code: Operability Time: 4320 Hours Radiation Zone: R441 C Zone Oose: 1.4xl06 Rads 3.0 COMPONENT SAFETY FUNCT ION FOR-POS-V/4 is the position switch for air-operated valve FOR-V-4. This valve is part of the primary containment isolation boundary. Regulatory Guide 1.97 requires the position indication of the containment isolation valves. 4.0 UALIFICATION STAIUS 4.1 Summar of uglification Status For LOCA conditions, the position switch is exposed to maximum condi-tions of 128'F, 14.7 psia and 64% relative humidity. The position switch is designed for a maximum temperature of 160'F and a maximum humidity of 95%, indicating that the switch should function properly under these LOCA conditions. However, due to the lack of documentation, the switch is scheduled for replacement with a documented, qualified switch. In the interim, a justification for operation is provided.
- 4. 2 Parameters Re uirin Justification Temperature, humidity and radiation dose.
5-0 JUSTIF ICATION FOR INTERIM OPERATION FDR-POS-V/4 indicates the open or closed position of valve FDR-V-4. FDR-V-4 is normally open and is automatically closed by the following isolation signals during LOCA conditions: a- Isolation Signal A - Reactor vessel low water level (trip 2)
- b. Isolation Signal F - High drywell pressure FDR-V-4 closes approximately 15 seconds after receipt of an isolation signal and thus will close prior to experiencing any harsh environment consequences.
FDR-POS-V/4 will indicate the correct valve position to verify containment isolation for approximately 12 minutes after the accident until the mild environment radiation dose (104 rads) has been exceeded. It would take approximately 54 hours into the accident to achieve the maximum temperature of 128'F and 350 hours to achieve the maximum humidity level of 64%. Therefore, the correct valve position can be monitored for at least the first 12 minutes. Within the first 12 minutes, FDR-V-4's safety function of containment isolation can be verified by the operator. The containment isolation function is assured in any event by the qualified valve (FDR-V-4) and its redundant, qualified back-up valve, FDR-V-3. Also, both FDR-V-4 and FDR-V-3 fail in the close position, which provides further assurance of accomplishing containment isolation. In the event that FDR-POS-V/4 fails after the first 12 minutes, there are three possible consequences. The position switch could indicate open, closed, or neither. In each case, the correct position indication was available for at least 12 minutes, during which time the operator could verify containment isolation. After this time, no further operator action is required because no credible event can cause both isolation valves to open.
6.0 CO NCLUS IO N Interim operation is justified on the following basis:
- 1. FOR-POS-V/4 will be operable for at least the first 12 minutes following an LOCA, during which time the operator can verify containment isolation.
- 2. Valve FOR-V-4 and its associated solenoid valve FOR-SPY-4 are ualified and will be automatically closed by either reactor vessel ow water level or high drywell pr essure isolation signals.
- 3. The operating time of the valve is approximately 15 seconds, thus ensuring that the valve will close prior to experiencing any harsh environment consequences.
- 4. There is a redundant valve (FOR-V-3) in series with FOR-V-4. Valve FOR-V-3 and its associated solenoid valve FOR-SPV-3 are qualified.
and provide a redundant means of containment isolation.
E IPMENT XSTIF ICATION f13 1.0 COMPONENT IDENTIF I CAT ION EPN: HPCS-FT-5
Description:
HPCS Pump Discharge Flow Indication Transmitter Component Type: Di fferential Pressure Transmitter Manuf acturer/Model: GE/Ba i 1 ey 555 2.0 ACCIDENT G)NDIT IONS LOCA HE LB Accident Profile: Use Code: 1/2 Operability Time: 24/4320 Hours Radiation Zone: R471B Zone Dose: 5.0xl05 Rads 3.0 COMPONENT SAFETY FUNCT ION Regulatory Guide 1.97 requires that the HPCS system discharge flow be monitored during and following an accident. This allows the operator to assess the operating status of the HPCS. HPCS-FT-5, located at the discharge of HPCS-P-1, provides control room indication of the HPCS flow to the reactor vessel. The flow transmitter must function for 24 hours to mitigate an accident, and must not fail in a detrimental manner for 6 months following the accident. 4.0 ALIFI CATION STATUS 4.1 Summar of uglification Status Flow transmitter HPCS-FT-5 has been qualified for the following environmental conditions: 169'F, 15.2 psia, 100% relative humidity. Radiation test data is not currently available. HPCS-FT-5 is being replaced by an environmentally qualified component. However, a systems justification is provided in the event that it is not replaced prior to fuel load.
4.2 Parameters Re uirin Justification Radiation dose. 5.0 JUSTIFICATION FOR INTERIM OPERATION Flow transmitter HPCS-FT-5 provides indication of the HPCS system operation as required by Regulatory Guide 1.97. The flow transmitter will function for a length of time, thereby providing an after indication of HPCS operation. Continued operation of the HPCS system, initiation, can be successfully monitored using alternate system indication. The HPCS system, upon initiation, will pump non-radioactive water from the condensate storage tank ( CST). 135,000 gallons of water in the CST is dedicated to the HPCS and RCIC systems. The HPCS pump rate varies from 1550 to 6856 gpm depending on the pressure in the reactor vessel. If the HPCS pump moves water at its maximum rate, the allotted CST water (135,000 gal. ) will be used in approximately 20 minutes. At this point, HPCS suction will switch over to the radioactive fluid in the suppression pool and HPCS-FT-5 will begin to be exposed to radiation. The flow transmitter will operate for approximately 43 minutes after HPCS suction is switched over to the suppression pool before it receives a harsh (104 Rads) radiation dose. Therefore the flow transmitter can operate for at least 63 (20+43) minutes. From examination of the pressure graphs in section 6.3 of the WNP-2 FSN, for even a small LOCA break (.lft~), the reactor vessel will depressurize within 15 minutes to the point where low pressure makeup can be utilized. Therefore, the flow tr ansmitter can be expected to operate successfully iomediately after and well into the accident. Should the flow transmitter fail during HPCS operation, there are alternate qualified instruments which the operator can depend on to monitor the HPCS system status. This instrumentation includes:
- 1. HPCS power indication lamp (input signal from circuit breaker HPCS-CB-Pl) and HPCS pump ammeter (i nput signal from HPCS-AM-Pl).
The circuit breaker indication light will tell the. operator if power is available to the pump, and the pump ammeter will provide indication that the pump is running.
- 2. HPCS injection valve HPCS-V-4 position indication (input signal from motor operator limit switch HPCS-LES-V4). This indicates that a flow path exists to the reactor vessel.
- 3. Reactor vessel water level recorder MS-LR/PR-623A (input signal from level transmitter AS-LITS-26A) is qualified for the first 10 years of normal plant operation. Therefore, it is qualified to perform of 24 hours.
its safety function for the required operating period This reactor vessel level indication will provide information that the reactor vessel is receiving water. The above instrumentation will allow the operator to monitor the operating status of the HPCS and to take corrective action such as manual depressurization using ADS, if necessary. I 6.0 CONCLUS ION Interim operation is justified on the following basis.
- 1. Flow transmitter HPCS-FT-5 will provide reliable HPCS system information for at least one hour into the accident.
- 2. Alternate qualified indication is available to the operator to monitor the operation of the HPCS after system initiation. The operator also has instrumentation available to determine if ADS operation is required.
0 E(UIPMENT JUSTIFICATION tl4 1.0 COMPONENT IDENTIFICATION EPN: LPRM-DET-(Refer to Table A)
Description:
Local Power Range Monitor (LPRM) Detectors Component Type: Neutron Flux Detector Manufacturer/Model: GE/163C1154Gl Control 2.0 ACCIDENT CONDITIONS LOCA I HELB ~Rod Duo Accident Profile: N/A N/A None* Use Code: Operability Time: 0.17 Hours Radiation Zone: Zone Dose: 2.04xlOB Rads (Gama dose over 0.17 hours LOCA dose)
- No accident profile is available for environmental conditions within the in-core LPRM housings. Design and operability parameters are per GE specification.
3.0 COMPONENT SAFETY FUNCTION The LPRM detectors measure localized neutron flux in the reactor core over the full power range (i.e. 15 to 120K of full power) for input to the Average Power Range Monitor (APRM). The APRM, in turn, averages the signals from the detectors and provides a reactor scram when neutron flux exceeds predetermined limits following a Control Rod Drop Accident. 4.0 UALIF ICATION STATUS 4.1 Sugar of ual ification Status gualification data not available; however, design data is per GE specification. In the interim, a systems justification is provided. 4.2 Parameters Re uirin Justification Pressure, temperature, humidity and radiation dose.
l 5.0 JUSTIFICATION FOR INTERIM OPERATION These LPRM detectors are designed in accordance with GE Specification 22A2843AA to monitor reactor core neutron flux up to 120K of rated power (or 3.4x1014 NV). Similarly, the detectors are designed to withstand a gamma dose rate corresponding to the 1205 power level. Once 100K power ss exceeded, the operator is trained to decrease power levels back to normal (i.e., less than or equal to 1005 power). If corrective action is not taken by the operator to decrease power and the power level continues to increase, the LPRM input to the APRM system automatically initiates a reactor scram at 1205 power. Insertion of the control rods would then iomediately reduce neutron flux levels in the core. Subsequent use of the detectors is not required to perform any safety related trip functions. Since the detectors are housed in dry tubes within the RPV and are hermetically sealed from the drywell post-accident environment, operability can be assured for safe shutdown of the plant.
6.0 CONCLUSION
The LPRM detectors are designed to survive and operate under normally harsh conditions. Their intended trip function is not compromised since the time required to perform this function is less than 10 minutes. Therefore, interim operation is justified.
IN E UIPMENT JUSTIFICATION 015 1.0 COMPONENT IDENTIFICATION EPN: LPRM-CONN-(Refer to Table A)
Description:
LPRM Cable Connector Component Type: Electrical Connector Manufacturer/Model: Amphenol/CAT X901-200 2.0 ACCIDENT CONDITIONS Control LOCA HELB ~Rod Oro Accident Profile: N/A N/A None Use Code: Operability Time: 0.17 Hours Radiation Zone: Zone Dose: 7xl07 Rads (LOCA dose) 3.0 COMPONENT SAFETY FUNCTION The LPRM system measures localized neutron flux in the reactor core over the full power range (i.e., lX to 120K of full power) for input to the Average Power Range Monitor (APRM). The APRM, in tur n, averages the LPRM signals and provides a reactor scram when the neutron flux exceeds predetermined limits following a Control Rod Drop Accident. The LPRM connectors are electrical signal cable connections located dir ectly beneath the RPV. These connectors effectively protect the LPRM signal cable junction from exposure to the drywell environment.
- 4. 0 UALIF I CATION STATUS 4.1 Summar of uglification Status The LPRM cable connector was provided as an essential component in accordance with the specifications in the NSSS contract with General Electric. Environmental equipment qualification documentation for the LPRM cable connectors is incomplete at this time. Therefore, a justification is provided for interim operation while the qualification documentation is being completed.
4.2 Parameters Re uirin Justification Pressure, temperature, humidity and radiation dose. 5.0 JUSTIFICATION FOR INTERIM OPERATION The connectors are part of the LPRM system which are designed in accordance with GE specification 22A3008. This plant interface specification requires that LPRM equipment located in the drywell be capable of sustaining maximum drywell pressure, temperature and humidity conditions (i.e., 55 psig, 340'F and 100K RH). Furthermore, the LPRM connector is designed to withstand an integrated dose of 2.6x107 rads. Operability of the connectors after a control rod drop accident is only required for 0.17 hrs., and an integrated dose of 2.6xl07 rads will not be received by the connector even 0.17 hrs. after a LOCA. Since a LOCA creates a considerably more hostile environment than a Control Rod Drop Accident, operability of the LPRM System can be assured.
6.0 CONCLUSION
Interim operation is justified on the basis that the LPRM system is designed by GE specification to perform its safety-related functions in environmental conditions considerably more severe than a Control Rod Drop Accident.
EQJ IPMENT JUSTIFICATION 816 1.0 COMPONENT IDENT IFI CAT ION EPN: MS-R E-3A, 38
Description:
Main Steam Line Radiation Monitors Component Type: Radiation Element Manuf acturer/Model: General El ectri c/237X731 G001 2.0 ACCIDENT CUNDIT IONS Control LOCA HE LB ~Rod Oro Accident Profile: N/A N/A None Use Code: Operability Time: 0.17 Hour s Radiation Zone: R5010 Zone Dose: 4. 2xl06 Rads 3.0 COMPONENT SAFETY FUNCT ION MS<E-3A and 38 provide a main steam line high radiation signal which scrams the reactor and closes the MSIV's following a Control Rod Drop Accident (CRDA). 4.0 AL IF I CAT ION STATUS 4.1 Summar of uglification Status No qualification documentation is provided, but the radiation monitors were purchased per GE specification (See GE OEM Instruction Manual 237X731Gl. ). These REs are designed to withstand maximum steam tunnel temperature, pressure and humidity conditions for WNP-2. However, because the qualification data is incomplete at present, the following justification for interim operation is provided.
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- 4. 2 Parameters Re uirin Justification Radiation dose.
5.0 JUSTIFICATION FOR INTERIM OPERATION MS-RE-3A and 3B are used to detect gross fuel failure in the reactor core. When high radiation is detected near the steam lines, a scram is initiated to limit the release of fission products in the fuel following, a CRDA. These radiation elements are designed to meet the 40 year normal operation integrated dose requirements in GE specification 22A3008 of 1.8xl06 rads. The integrated dose after 18 months due to normal operation (i.e. before the first refueling outage) plus the 0.17 hour dose following an accident will be less than 2 x 105 rads. Therefore, failure due to radiation should not occur because the normal operating design dose is greater than the CRDA accident dose. In the unlikely event that the RE s fail during an accident, the following justification is provided to assure that no safety-related functions are compromised. Since the primary means of detecting a CRDA is reactor high power measured by the Neutron Monitoring System (NS), the Main Steam Line Radiation Monitoring System serves only as a backup. This radiation monitoring system's response is considerably slower than the NS because of its location. Once a CRDA has been identified by the NS, the MSIV's would close within 5.5 seconds. After the MSIV's close, the radiation elements will no longer be required. The operability time for these RE's is, therefore, limited to 5.5 seconds, during which scram and isolation functions will already have been initiated.
6.0 CONCLUSION
Interim operation is justified on the following basis:
- 1. MS-RE-3A and,3B are designed to function in a post-LOCA steam tunnel environment.
- 2. There are alternate means of detecting a CRDA. The NS would successfully accomplish safe shutdown of the plant.
- 3. No safety-related systems are degraded if these radiation elements fail.
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E IPMENT JUSTIF ICATION 817 1.0 I COMPONENT IDENT IF CAT ION EPN: ROA-POS-V/1
Description:
Reactor Building Outside (HVAC) Air Isolation Valve Position Switch Component Type: Position Switch Manufacturer/Model: NNCO Controls/70050100 2.0 ACCIDENT (DNDIT IONS Accident Profile: N/A Use Code: Operabi1 i ty Time: 4320 Hours Radiation Zone: R572F Zone Dose: lxl06 Rads (estimate) 3.0 COMPONENT SAFETY FUNCT ION ROA-POS-V/l. is the position switch for the air-operated valve ROA-V-l. This valve isolates the reactor building from the outside atmosphere. Regulatory Guide 1.97 requires indication of this emergency ventilation damper position. This position switch is required to function only for LOCA conditions. 4.0 ALIFICAT ION STATUS 4.1 Summar of uglification Status For LOCA conditions, the position switch is exposed. to maximum accident conditions of 128'F, 14.7 psia and 64K relative humidity. The position switch is designed for a maximum temperature of 194 F. and a maximum humidity of 95%, indicating the switch should function properly under these LOCA conditions. However, due to the lack of documentation, the switch is scheduled for replacement with a documented, qualified switch. In the interim, a justification for operation is provided.
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4.2 Parameters Re uirin Justification Temperature, humidity and radiation dose. 5.0 JUSTIFICATION FOR INTERIM OPERATION ROA-POS-V/1 provides the open or close position of valve ROA-V-l. ROA-V-1 is normally open and is automatically closed by the following isolation signals during LOCA conditions:
- a. Isolation Signal A - Reactor vessel low water level (Trip 2)
- b. Isolation Signal F - High drywell pressure
- c. Isolation Signal Z - High radiation level in the reactor building exhaust venti 1 ation system.
ROA-V-1 closes approximately 4 seconds after receipt of an isolation signal and thus will close prior to experiencing any harsh environment. ROA-POS-V/1 will indicate the cor rect valve position to verify reactor building isolation for approximately 7 minutes after the accident until the mild environment radiation dose (104 rads) has been exceeded. It would take approximately 54 hours into the accident to achieve the maximum temperature of 128'F and 350 hour s to achieve the maximum humidity level of 64K. Therefore, the correct valve position indication can be expected for at least the first 7 minutes. Within the first 7 minutes, ROA-V-1's safety function of reactor building isolation can be verified by the operator. The reactor building isolation function is assured in any event by the qualified valve, ROA-V-l, and its qualified redundant back-up valve, ROA-V-2. Also, both ROA-V-1 and ROA-V-2 fail in the closed position, which provides further assurance of accomplishing reactor building isolation. In the event that ROA-POS-V/1 fails after the first 7 minutes, there are three possible consequences. The position switch could indicate open, closed, or neither. In each case, the correct position indication was available for at least 7 minutes, during which time the operator could verify reactor building isolation. After this time, no further operator action is required because no credible event can cause both isolation valves to open.
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6.0 CON CLUS ION Intet im operation is justified on the following basis:
- 1. ROA-POS-V/1 will be operable for at least the first 7 minutes, during which time the operator can verify containment isolation.
- 2. Valve ROA-V-1 and its associated solenoid valve ROA-SPV-100 are qualified and will be automatically closed by any one of three isolation signals generated by qualified monitors.
- 3. The valve will isolate in approximately 4 seconds, prior to experiencing any harsh environment.
- 4. There is a redundant, qualified valve (ROA-V-2) in series with ROA-V-1 which provides another means of reactor building isolation.
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Egl IPMENT JUSTIFICATION 818 1.0 COMPONENT IDENT IF I CAT ION EPN: SRM-D ET-1 D
Description:
In-Cove Source Range Detector Component Type: Neutron Flux Detector (10"65 to 10% power) Manufacturer/Model: GE/368X432G001 2.0 ACCIDENT CONDITIONS LOCA Accident Profile: N/A Use Code: Operability Time: 4320 Hours Radiation Zone: Containment Zone Dose: 5xl06 Rads
- No accident profile is available for environmental conditions within the in-core SRM housings.
3.0 COMPONENT SAFETY FUNCTION SRM-DET-10 is a reactor core neutron flux detector which measures the power level over the range of 10-6X to 10K of full power. This detector is part of the equipment required by Reg. Guide 1.97 to provide long-term post-accident monitoring capability. 4.0 ALIFICATION STATUS 4.1 Sumar of ual if ication Status gualification data is not available. Design data is per GE specification. Therefore, a justification for interim operation is provided.
- 4. 2 Par ameters Re uirin Justification Pressure, temperature, humidity and radiation dose.
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5.0 JUSTIFICATION FOR INTERIM OPERATION This SRM it detector is housed w h i n dry tubes inside the RPV. It is effectively sealed from drywell pressure, temperature, and humidity. Ouring startup, the detector is fully inserted into the core region of the reactor and is used to measure power levels between 10-6 and 10-3X of full power. At full power it is retr acted to a position below the active core and measures the power range between 10-3 and 10% of full power. In the retracted position, the detector is designed to maintain operability up to 120% of full power according to GE specification 22A2843AA. At 120%, the APRM will scram the reactor, thus reducing neutron flux levels and protecting the SRM detector from exceeding its design limit for neutron flux. The subsequent post-accident radiation environment would be less severe than the normal operating environment due to a lower flux level. Since the normal operating environment is considerably harsher than post-accident conditions, long-term operability is assured for the SRM detector.
6.0 CONCLUSION
Interim operation is justified on the basis that the GE design of this detector provides for long-term service in a harsh environment due to normal operation. The harsh environment due to an accident is considered less severe, and, therefore, provides reasonable assurance that the detector will function satisfactorily for the 6 months following a LOCA.
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E UIPMENT JUSTIFICATION 019 1.0 COMPONENT IDENTIFICATION EPN: SW-M0-187A,188A
Description:
Standby Service Water System Valve Motor Operator Component Type: Motor Operator Manufacturer/Model: Limitorque Corp./SMB-00 2.0 ACCIOENT CONOIT IONS LOCA HELB
.Accident Prof ile: N/A Use Code:
Operability Time: 4320 Hours Radiation Zone: R548L Zone Oose: 1.5x105 Rads 3.0 COMPONENT SAFETY FUNCTION SW-MO-187A and 188A provide alternate cooling to the Fuel Pool Cooling (FPC) System heat exchangers under abnormal plant conditions when the Reactor Building Closed Cooling Water (RCC) System is unavailable. SW-MO-187A and 188A open valves SW-V-187A and 188A to provide Standby Service Water (SW) System cooling to FPC-HX-lA. 4.0 UALIFI CAT ION STATUS 4.1 Summar of ual ification Status The motor operators were purchased qualified; however, unqualified motors were inadvertantly used. The vendor is scheduled to provide qualified replacement motors.
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The following justification for interim operation is provided in the event that the motor operators are not replaced prior to fuel load: 4.2 Parameters Re uirin Justification Temperatur e, pressure, humidity and radiation dose. 5.0 JUSTIFICATION FOR INTERIM OPERATION Alternate cooling water is provided to FPC-HX-lA through valves SW-V-187A and SW-V-188A. Although fuel pool cooling is not required until the first refueling outage, the valves must remain closed in the interim so that operation of the Standby Service Water System would not be degraded. The motor operators will be qualified prior to the first ref uel ing outage. Currently the motor operators are fully qualified except for the motors. Review of the construction of the operator and limitorgue test data on operators with similar motors identifies no failure mode of the motor that would cause the operator to open the valve.
6.0 CONCLUSION
Interim operation is justified based on the following:
- 1. SW-NO-187A and 188A are not required to operate post-accident since fuel pool cooling is not needed prior to the first refueling outage.
- 2. There is no failure mode of the motor that would cause SW-MO-187A and 188A to open their associated valves to an undesirable position.
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E UIPNENT JUSTIFICATION 820 1.0 COMPONENT IDENTIFICATION EPN: SM-PS-1014
Description:
Service Water Supply to H2-02 Analyzer Pressure Switch Component Type: Pressure Switch Manufacturer/Model: ASCO/SCllAR 2.0 ACCIDENT CONDITIONS LOCA HELB Accident Prof i le: N/A Use Code: Operability Time: 4320 Hours Radiation Zone: R548E Zone Dose: lxl06 Rads (estimate) Tmax ~ 107'F, Pmax Atm. 3.0 COMPONENT SAFETY FUNCTION Pressure switch SW-PS<<1014 is associated with the supply/return of . cooling water to the containment hydrogen-oxygen monitoring system under LOCA conditions. Instrumentation to monitor containment hydrogen and oxygen is required in accordance with Regulatory Guide 1.97 to provide i nformation to indicate the potential for breach of the primary containment. Under LOCA conditions, the Standby Service Mater (SW) System is initiated. When SM pressure reaches 100 psig, pressure switch SW-PS-1014 closes valves SM-V-204 and SW-V-212 and opens valves SW-V-201 and SW-V-213. Thus, the H2-02 analyzer equipment receives cooling water from the SW system through these valves under LOCA conditions.
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4.0 UALIFI CATION STATUS 4.1 Summary of ualif ication Status Pressure switch SM-PS-1014 is located in a mild environment for temperature, pressure, and relative humidity. The pressure switch is required to be qualified for radiation conditions resulting from a LOCA inside primary containment. The radiation dose for its location is currently being reanalyzed. Previous radiation dose calculations show a dose of 9.0xl03 rads based on shine from primary containment and nearby piping. It is estimated that the radiation dose will increase to approximately lxl06 rads when dose contr ibutions from the H2-02 analyzer's process stream are taken into account. Therefore, based on the dose estimate of 106 rads, the equipment is not qualified and a justification is included. for interim operation while the pressure switch is being qualified. 4.2 Parameter Re uirin Justification Radiation dose. 5.0 JUSTIFICATION FOR INTERIM OPERATION Since plant operators will manually initiate the hydrogen-oxygen recombiners within 2.75 hours after declaration that a LOCA condition exists (See JIO 86 for CMS-AY-1,3), the H2-02 analyzers do not perform any vital safety function post-LOCA. Therefore, SM-PS-1014 is not required post-LOCA because its only safety function is to control SM-U-201, 204, 212 and 213 which support H2-02 analyzer operation. In addition, the failure of SW-PS-1014 will not degrade other systems or components that are required for safety. The failure of SM-PS-1014 would result in the loss of Standby Service Mater for cooling the H2-02 analyzer process stream. This could result in failure of the H2-02 analyzer. A failure of the H2-02 analyzer would not have an adverse safety impact since operators will manually initiate the hydrogen-oxygen recombiners without reference to H2-02 concentrations.
6.0 CONCLUSION
Interim operation is justified since operator action to initiate the hydrogen recombiner system precludes the need for the Hp-02 analyzer. The failure of SM-PS-1014 will not degrade the safety ot the plant and is not required to perform a safety function.
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E UIPMENT JUSTIFICATION 821 1.0 COMPONENT IDENTIFICATION EPN: SW-V-201, 204, 212, 213
Description:
H2-02 Analyzer Supply/Return Cooling Water Valves Component Type: 0.5" Solenoid Valve Manufacturer/Model: Marotta Valve Corp./MV229MS-L2 2.0 ACCIDENT CONDITIONS LOCA HELB Accident Prof ile: N/A Use Code: Operability Time: 4320 Hours Radiation Zone: R548E Zone Dose: lx106 Rads (estimate) 3.0 COMPONENT SAFETY FUNCTION The solenoid valves are associated with the supply/return of cooling water to the containment hydrogen-oxygen monitoring system (H2-02 analyzers CMS-AY-1 and CMS-AY-3). Instrumentation to monitor containment hydrogen and oxygen is required in accordance with Regulatory Guide 1.97 to provide information to indicate the potential for breach of the primary containment. Air samples are drawn from containment and are passed through cooling coils and moisture separators before being analyzed. After passing through the H2-02 analyzers, the air samples are exhausted to
~ containment. The coils are cooled by two water sources. Under normal operating conditions, the Plant Service Water (TSW) System provides the water. Under LOCA conditions, the Standby Service Water (SW) System is initiated. When SW pressure reaches 100 psig, pressure switches close the TSW valves SW-V-204 and SW-V-212 and open the SW valves SW-V-201 and SW-V-213.
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~ Suranar of ual ification Status The solenoid valves are being tested, and the test data is currently not available. In the interim, the following justification is provided:
4.2 Parameters Re uirin Justification Temperature, pressure, humidity and radiation dose. 5.0 JUSTIFICATION FOR INTERIM OPERATION Since plant operators will manually initiate the hydrogen oxygen recombiners within 2.75 hours after declaration that a LOCA exists (See JIO 86 for CMS-AY-1), the H2-02 analyzers do not perform any active safety function post-LOCA. Since valves SW-V-201, 204, 212, and 213 only support H2-02 analyzer operation, they are not required to be functional post-LOCA. In the unlikely event that valves SW-V-201, 204, 212, and 213 fail, their failure will not degrade the Standby Service Water System. The required service water flow rate will be maintained, by the system independent of the position of the valves. Safety-related electrical systems will also not be degraded by the failure of these valves.
- 6. 0 CONCLUSION Interim operation is justified on the following basis:
- 1. SW-V-201, 204, 212, 213 are not required in the event of a LOCA or HELB since the H2-02 analyzers are not necessary to perform the required safety function.
- 2. Failure of these valves will not degrade other systems or components required for safety.
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E UIPMENT JUSTIFICATION 822 1.0 COMPONENT IDENTIFICATION EPN: TIP-V-1, 2, 3, 4 and 5
Description:
Traversing In-Core Probe (TIP) Isolation Valve Component Type: Explosive Activated Shear Valve Manufacturer/Model: GE/PP13681302G002 2.0 ACCIDENT CONDITIONS LOCA HELB Accident Profile: Use Code: 1/2 Operability Time: 1 Hour/4320 Hours Radiation Zone: R501P Zone Dose: 1.0x 106 Rads 3.0 COMPONENT SAFETY FUNCTION These shear valves provide containment isolation. They are located iamediately outside primary containment and are mounted within a ball valve/shear valve isolation assembly. Closure of the shear valves is manually initiated by a switch in the control room and results in an explosive shearing of the drive cable and sealing of all TIP guide tubing. 4.0 UALIFICATION STATUS 4.1 Summar of ual ification Status These valves meet pressure, temperature and humidity qualification requirements following an accident. However, due to lack of documentation for radiation qualification, the effects of a high post-LOCA radiation environment cannot assure long-term operability. The following justification for interim operation is thus provided.
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Washington Public Power Supply System P.Q. Box 968 3000George Washington Way Richland, Washington 99352 (509) 372-5000 Docket No. 50-397 June 30, 1983 G02-83-590 Director of Nuclear Reactor Regulation Attention: Mr. A. Schwencer, Chief Licensing Branch No. 2 Division of Licensing U. S. Nuclear Regulatory Commission
'I/ashington, D.C. 20555
Dear Mr. Schwencer:
Subject:
NUCLEAR PROJECT NO. 2 JUSTIFICATION FOR INTERIM OPERATION This letter transmits the Supply System's revised Environmental Equip-ment gualification Report for Safety-Related Equipment. This revision. provides an update to the equipment list and responds to concerns raised in Supplement 3 to the Safety Evaluation Report (NUREG-0892). The list. included in this report contains electrical equipment important to safety as required by 10 CFR 50.49(d). A major element of the revision is full development of our Justification for Interim Operation (JIO) of safety-related electrical equi pment as required by 10 CFR 50.49. The NRC Supplement No. 3 to the Safety Evaluation Report (SSER) raised a number of concerns. Some of those are responded to within the document being transmitted by this letter. It is appropriate to address some of the concerns in this letter itself. The SSER requested discussion on how changes to equipment items resulting from IE Bulletins, Circulars, and Information Notices have been, or will be, evaluated for their impact on qualification. The Supply System has .a controlled, documented process for review and disposition of such NRC documents. Plant changes flowing from such responsive actions are con-trolled by the design change control process. The SSER called on us to reevaluate the adequacy of the plant walkdown performed before the staff's audit. The Supply System has caused system reviews to be conducted to add or delete items from the ClE list. These reviews improve the quality of the program and allow us to make the con-clusionary judgments necessary regarding environmental qualification of this facility. Additionally, in the efforts to complete the lists., incor-porate plant changes and complete data entries, mistakes in the list were observed and corrected.
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Oocket No. 50-397 Oirector of Nuclear Reactor Regulation Attention: Hr. A. Schwencer, Chief Page Two June 30, 1983 G02-83-590 The SSER requests our review of the Equipment. gualification Program to establish thai all data reported is applicable. Me have reviewed the program. As discussed above, system reviews were conducted to improve and verify the equipment lists. Computer programs were developed and used to assure that other programs had communicated properly in the gener ation of lists, and checking and reviewing were emphasized activities in conduct of activities leading to this docu-ment. Me can state that significant effort has been exerted to cor-rect the data and we judge that there are no significant omissions or errors. On June 27, 1983 our Architect-Engineer identified four main steam system pressure switches which were recently added to the plant design in response to NRC requirements and belong within the scope of our quali-fication program but were not discussed in these documents. This situa-tion is highlighted here because of our responsibility to be open with you and also to make a point. Me expect that minor changes will continue to be made in the list and qualification program. All such changes are addressed by our qualification efforts. At the time of fuel load, we expect to be able to judge the plant properly qualified. Last moment changes are expected. Me know they must be addressed and require quali-fication documentation. Me believe such changes are a reality and must be taken into account in our operation and your evaluation of this qualification process. The SSER directs us to submit the results of our review for all safety-related mechanical (SRN) equipment located in a harsh environment. Our SRM program addresses all active safety-related mechanical equipment in a harsh environment. The results of our review are included in this Environment gualification Report. As the NRC is aware, the Supply System has an independent Verification Program underway. The Equipment qualification Program was reviewed and errors were found in the calculated profiles associated with high energy line breaks (HELB). Me have calculations underway correcting the errors. Mhere this document shows equipment status as qualified, it is qualified to LOCA and "old" HELB environments. Preliminary calculations show that most of the HELB profiles will not change significantly. There-fore, it is our judgment that most equipment wi 11 remain qualified to the
Docket No. 50-397 Director of Nuclear Reactor Regulation Attention: Mr. A. Schwencer, Chief Page Three June 30, 1983 G02-83-590 revised HELB profiles when they are available and have been evaluated. We will provide verification of this judgment with the supporting document revisions as soon as we can, and certainly by July 30, 1983, to minimize the pressure on your review. We are available to discuss these documents with the staff. Very truly yours, G. D. Bouchey, 340 Manager,- Nuclear Safety and Regulatory Programs GDB:KRW:st Attachment cc: Mr. R. Auluck, NRC Mr. W. S. Chin, BPA Mr. A. Toth, NRC, Site
ENGINEERING REPORT
'NP-2 ENVIRONMENTAL QUALIFICATION REPORT FOR SAFETY RELATED EQUIPMENT July 1983
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TABLE OF CONTENTS VOLUME 1 PAGE
1.0 INTRODUCTION
1 2.0 CLASS 1E EQUIPMENT LIST 3 3.0 ENVIRONMENTAL SERVICE CONDITIONS 8 3.1 NORMAL AND ABNORMAL CONDITION 8 3.2 ACCIDENT CONDITIONS - HARSH ENVIRONMENTS 8 3.2.1 Temperature/Pressure Inside Containment 8 3.2.2 Spray 9 3.2.3 Radiation Inside and Outside Containment 9 3.2.3.1 Beta Radiation 10'3 3.2.4 Flooding 3.2.5 Temperature/Pressure Outside Containment 15 3.3 MILD ENVIRONMENT AREAS IN SECONDARY CONTAINMENT 16
- 4. 0 I QUAL IF CATION METHODS 18 4.1 EQUIPMENT EVALUATIONS 18 4.2 TECHNI CAI APPROACH. 19 4.2.1 Equipment Inside Primary Containment 20 4.2.2 Equipment Inside the Reactor Building (Seconary Containment) 21 4.2.3 Equipment Located in Mild Environments. 22 4.2.4 Margin 22 4.2.5 Aging 23 4.2.6 Instrument Accuracy 24 4.2.7 Interface Qualification 24 4.3 SAFETY RELATED MECHANICAL EQUIPMENT 26 4.4 DOCUMENTATION 32 4.5 MAINTENANCE AND SURVEILLANCE 32 5.0 QUALIFICATION RESULTS 35 6.0 JUSTIFICATION FOR INTERIM OPERATION 37 7.0
SUMMARY
40
8.0 REFERENCES
42 APPENDIX A - CLASS 1E EQUIPMENT LIST AND SAFETY RELATED MECHANICAI LIST APPENDIX B - ENVIRONMENTAL SERVICE CONDITIONS VOLUMES 2 AND 3 APPENDIX C - EQUIPMENT QUALIFICATION REPORTS APPENDIX D - JUSTIFICATION FOR INTERIM OPERATION
- 1. 0 INTRODUCTION The original equipment qualification requirements for Washington Public Power Supply System (Supply System) Nuclear Project Number 2 (WNP-2) were described in the PSAR. These requirements specified that NSSS equipment be designed to good nuclear industry practices and Balance of Plant equipment be qualified to IEEE 323-71. Initial equipment purchases were made to these requirements .
In November 1974, Regulatory Guide 1.89 was issued identifying IEEE 323-74 as the generally acceptable level for qualification of Class lE equipment. Based on construction permit requirements, the Supply System was not required to upgr ade the qualification status of the equipment. A review was made to determine the impact of the revised'uidance. Based on this review, it was determined that there was not a need for general upgrade of equipment. NUREG-0588 (Reference 1) was issued for covalent in December 1979 to promote a more orderly and systematic implementation of equipment qualification pro-grams by the industry. It also provided guidance to the NRC staff for its use in ongoing licensing review for new as well as for the near-term oper-ating license plants. The WNP-2 Construction Permit SER was issued prior to July I, 1974; therefore, the basis for the WNP-2 review was the Category II requirements.. In February 1980, the NRC requested (Reference 2) that the Supply System perform a review of the existing environmental qualification program to identify the degree to which the program complied with the criteria and positions in NUREG 0588. Deviations from the NUREG were to be justified. The Supply System provided comments to the NUREG in April 1980, expressing concern regarding certain criteria and positions. Revision 1 of the NUREG was issued with responses to the Supply System's concerns as well as other concerns raised by the industry.
The code of regulations was revised in January of 1983 by addition of 10CFR Part 50.49. Within this revision are provisions which require WNP-2 equip-ment qualification to meet the criteria of NUREG 0588 Category II. For equipment within the scope of 10CFR50.49 that can not be documented as meet-ing the applicable criteria, a justification that WNP-2 can be operated , safely pending completion of a gualification Program is required. In addi-tion, replacement equipment must be qualified to the provision of 10CFR Part, 50.49 unless sound reasons to the contrary exist. The Supply System has undertaken an aggressive equipment qualification program to assure that Class lE equipment is qualified to NUREG 0588, Category II. Class 1E equipment at WNP-2 have been identified. Normal, abnormal and accident service conditions have been defined for plant areas that could be exposed to a harsh environment. A detailed review of the available qualification data has been made for the equipment in harsh envi-ronments. Actions have been initiated to upgrade the qualification documen-tation where deficient and to requalify components, when necessary. This report describes the scope, methodology and results of the equipment quali-fication effort. This report also provides an analysis of the capability of operating WNP-2 during an interim period between September 1983 and November 30, 1985. This Justification for Interim Operation identifies critical electrical equipment required for safe operation. It also identifies equipment requiring further environmental qualification corrective actions and establishes the priority for our environmental qualification activities until commercial operation.
2.0 CLASS 1E EQUIPMENT LIST The first step of the Environmental Qualification Program is to define and identify the scope of systems and equipment involved. In order to ensure as complete a list as possible, the following assumptions and process were used. Assumptions: 1. All potential systems and equipment important to safety are powered from the Emergency Power Elec-trical System and the HPCS Emergency Electrical System.
- 2. All equipment important to safety that is added due to regulatory requirements (i .e., Regulatory Guide
- 1. 97 Cat. I and II, Rev. 2, TMI-2, Lessons Learned, etc.) wi 11 be added to the list as the designs are finalized and implemented.
- 3. Some equipment may not serve a safety related func-tion but could be exposed to accident environments and should be evaluated to determine if its failure could affect safety system performance.
Based on these assumptions, a detailed review was performed on the Emergency Power System. This review identified all equipment that could potentially be important to safety. In addition, a procedurally controlled process was implemented by our A/E that required any new designs to be reviewed and equipment evaluated to determine if it should be added to the Class IE list. This assures that as design changes are finalized, the equipment is added to the Qualification Program. The next step was to perform a design review to determine what equipment identified by the list are required to perform Class 1E functions. Addi-tional detail is provided in Section 6 and Appendix D.
Class 1E was defined according to IEEE 323-74 (Reference 3). The following definition was used: The safety classification of the electric equipment and systems that are essential to emergency reactor shutdown, containment isolation, reactor core cooling, and containment and reactor heat removal, or otherwise are essential in preventing signif-icant release of radioactive material to the environment. Instrumentation for the operator to follow the course of an accident was also defined as Class 1E. This includes instrumentation identified as a result of TMI-2 Lessons Learned and Regulatory Guide 1.97, Rev. 2 Category I and II. Specific criteria were then developed to determine the equipment that is Class 1E. The criteria and instructions for application of the criteria ar'e contained in Reference 4. Plant systems were reviewed in accordance with these criteria. The sources of information for the review were the FSAR, Technical Specifications, System Flow Diagrams, Electrical Diag< ams and Technical Manuals. Additional operational data were determined during the documentation review. The following information was determined for each component: o Use. The equipment use during accident and/or normal plant shutdown conditions . This is based on the categorization of equipment suggested in Item 2, Appendix E of NUREG 0588.
~5f t F t1 . Th C1 lK f ti f ti Pi f equipment or system is required to perform or monitor. Safety functions include emergency reactor shutdown, containment isola-tion, reactor core cooling, containment heat removal, reactor heat removal and prevention of release of radioactive material to the environment.
o Required Operatin Time. The time a component is required to be functional or retain its pressure integrity following a Design Basis Accident. This process assured the following equipment was included in the Equipment gualification Program.
- 1. Safety related electrical equipment (Class 1E) that is relied on to remain functional during and following accident exposure by LOCA, HELB, and rod drop design basi's events. To ensure:
A. The integrity of the reactor coolant pressure boundary (use code 1X, safety function code Bl); B. The capability to shut down the reactor and maintain it in a safe shut down condition (use code 1X, Safety Functions A, C, E, L, and J), C. The capabili lty to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the 10 CFR Part 100 Guidelines (use code 1X, Safety Functions A, Bl, B2, C, D, E, F, L and J).
- 2. Other equipment whose failure under postulated environmental conditions could prevent satisfactory accomplishment of safety functions defined above (use code 2, Safety Functions G, G1 and G2)(Reference Appendix D for methods used).
- 3. Post-accident monitoring equipment identified by Revision 2 of Regulatory Guide 1.97 Category I and II and NUREG 0737 (use code 1X, Safety Function I).
- 4. Equipment that may be exposed to the postulated environmental conditions but whose failure under any mode would not prevent satisfactory accomplishment of the safety functions defined above. (Use code 3X, no safety function for accidents defined above.)
- 5. Equipment that is in an area of the plant where environmental conditions would at no time be significantly more severe than the environment that would occur during normal plant operation, including anticipated operational occurences. (Use code 4X, safety functions A, 81, 82, C, D, E, F, I, L and J.)
A documentation review and plant walkdown was performed to determine the manufacturer's data for electrical equipment important to safety. The walkdown included verifying manufacturer, model, serial number and loca-tion. Equipment without unique identifying nameplate or markings, or not yet installed, was identified through applicable purchase and receiving documents. During the walkdown and documentation review the location of the equipment in the plant was documented to assist with the definition of the required service conditions and the calculation of the radiation exposure. A list of the electrical equipment in 1, 2, and 3 above is provided in Appendix A and constitutes the scope of electrical equipment important to safety on WNP-2 that are in the WNP-2 Environmental gualification Program. Section 4.5 of this report adresses the scope of the safety related mechan-ical qualification program. Appendix C contains the Equipment gualifica-tion Reports (Summary) except for electrical equipment that must maintain a passive mechanical function only (Use Code 2, Safety Function G). These are available fo'r review if required. The definition, identification and classification of equipment important to safety continues throughout plant life. This is due to plant revision required by regulatory changes, plant improvement modifications, replace-ment equipment, etc. WNP-2 has developed a program which requires that equipment important to safety be added to the Safety Related Equipment
Lists, preprocurement review for assuring procurement of qualified equip-ment or justifying sound reasons to the contrary, establishment of quali-fication files, and installation in a manner consistent with the equipment qualification . This report presents the scope of equipment important to safety defined by WNP-2 Class 1E List, Rev. 8 and Safety Related Mechani-cal List, Rev. 9. As design changes occur throughout the life of the plant, the WNP-2 equipment qualification program will ensure procurement and installation of equipment consistent with the requirements of 10CFR50.49.
3.0 ENVIRONMENTAL SERVICE CONDITIONS The normal, abnormal and accident service conditions were defined for all areas of pr imary containment and the reactor building containing Class 1E equipment. The service conditions were defined as described below. 3.1 NORMAL AND ABNORMAL CONDITIONS The temperature, pressure and humidity ranges expected during normal opera-tion were defined based on Reference 5 and 6. Abnormal conditions due to temporary HVAC failure are also defined in Reference 5 and 6. Appendix B presents the normal and abnormal conditions for primary containment and the reactor building. The 40-year normal radiation dose is included in the radiation doses dis-cussed in Section 3.2.3 of this report. 3.2 ACCIDENT CONDITIONS - HARSH ENVIRONMENTS The primary containment and most areas of the reactor building will be exposed to a harsh environment following a postulated LOCA/HELB. A harsh environment is defined as: An area that would be exposed to a significant increase in the maximum temperature, pressure and humidity during design basis events AND/OR the total radiation dose (normal + accident) is above 104 rad. 3.2.1 Temperature/Pressure Inside Containment and Mainsteam Tunnel The accident environments inside primary containment and mainsteam tunnel are defined according to Reference 5 and the WNP-2 FSAR (Reference 6). The accident profiles (1, 2, and 3), presented in Appendix B, were obtained from a General Electric analysis of the response of a BWR Mark II containment to a full spectrum of possible LOCA/MSLB. A WNP-2 plant specific analysis of the primary containment (Profile 1) was performed (Reference 20). The
purpose of the plant specific analysis is to demonstrate that sufficient margin exists between the GE Generic Profile and WNP-2 plant specific pro-file and to define the post accident conditions as required by NUREG 0588 (Figure C2). 3.2.2 ~Sr~a Operator initiation of a demineralized water spray could be used in primary containment at WNP-2 to mitigate the effects an an accident. No credit for this operator action has been taken in defining the temperature/pressure conditions inside containment. However, since spray impingement could occur (by operator action) its effect on Class 1E equipment inside primary con-tainment has been evaluated. 3.2.3 Radiation Inside and Outside Containment The accident radiation environments in the primary containment has been defined according to Section II.B.2 of NUREG 0737 (Reference 7) and NUREG 0588, Rev. 1. The calculated accident environment is based on the most severe nonmechanistic design basis accident during or following which equipment must function. This includes consideration of the entire spectrum of FSAR Chapter 15 accidents which can lead to a degraded core condition. The source term assumptions for postulated accidents are consistent with those defined in NUREG 0588 and Regulatory Guides 1.3 and 1.7. The source terms are calculated using the ORIGEN code (Reference 8). For the review performed in this report the radiation environment in the primary containment was defined per Reference 19. This is a plant specific evaluation. Reference 19 contains the methodology and results of this eval-uation. The results of this evaluation are provided in Appendix B for gene-ral areas of the containment and Reactor Building and also in Appendix C for each specific piece of equipment. The radiation environment in the reactor building (secondary containment) is defined according to Section II.B.2 of NUREG 0737 (Reference 7) and NUREG 0588, Rev. 1 and includes the sum of direct accident gamma dose, airborne gamma dose and 40-year normal gamma dose. The airborne dose is conservatively based on the maximum primary containment to reactor building leakage rate. Airborne activity in both the containment and reactor building was calcu-lated using the plateout assumptions of NUREG CR-0009 (Reference 9). The reactor building (secondary containment) was divided into zones to define the equipment doses. The worst target (Class 1 component with the highest dose) in each zone was then choosen. The total integrated dose (TID) to this component was calculated using the gAD-P5A computer code (Reference 10). This TID was used as the required qualification level for most equipment in the zone. In some zones multiple target specific TID's were calculated where there was wide variation of radiation levels within the zone. The methodology and results of the zone dose radiation evaluations are docu-mented in calculation packages in Reference 11. Appendix B of this report contains a table of the radiation doses inside primary containment and the radiation zone maps for the reactor building. It should be noted that the containment and the reactor building radiation levels are six month inte-grated doses. Doses for equipment with shorter operating times were deter-mined from the calculated packages in Reference 11 and are detailed in Appendix C on Equipment gualification Report Summary Sheets. 3.2.3.1 Beta Radiation The beta radiation effects on equipment important to safety are evaluated . Beta dose calculation and methods are presented in Reference 19 and 26. Host equipment important to safety on WNP-2 are installed with watertight electrical conduit and are constructed such that beta produces surface effects only to the external housing of the equipment. Internal sensitive components within the equipment are not exposed. There are, however, some equipment that are of a ventilated design. This equipment could experience an atmosphere internal to the equipment that is the same as the external atmosphere. The beta radiation dose to this ventilated equipment is dependent on the internal volume size of the equipment. The beta dose is determined through the use of energy dependent geometry factors, a ratio of the internal equip-ment volume to an infinite cloud, and the dose to a target at the center point face for a hemispherical cloud of gases. Thus, various volume sizes are evaluated and finite cloud beta doses determined. Beta radiation doses generally are less significant than gamma radiaton doses for Equipment gualification. This is due to the low penetrating power of beta particles in comparison to gambia r ays of equivalent energy. Of the general classes of ventilated equipment in WNP-2 (motor control centers, power panels, NEMA 1 junction boxes, NEMA 1 connection boxes on equipment, and cable), electrical cable is considered the most vulnerable to damage from beta radiation due to its relatively thin insulation thickness . The NRC has issued guidelines (Ref. DOR Guideline to IEB 79-01) relative to this area. Within these guidelines are criteria that allow further beta radiation effects reductions in order to address the degree of degration caused by beta versus more penetrating gamma. These guidelines allow a factor of 10 reduction in the ganma equivalent beta dose for the first 30 mils of electrical insulation material and another factor of 10 reduction for the next 40 mils of insulation (total 70 mils insulation equals reduc-tion factor of 100). Research programs within EPRI and the NRC are continuing in order to better understand and quantify reduction factors to be used. The Supply System is abreast of these programs and will factor these results into the equipment qualification program. Until better information is available, only a reduc-tion factor of 10 is used for equipment within the qualification program to equate to the testing and material radiation threshold levels based on gamma radiation. For cabling, reductions up to a factor of 100 are used provided adequate insulation and jacketing thickness are present. Reactor Buildin (secondary containment) Equipment important to safety located in the secondary containment Reactor Building receive radiation from several sources. These include direct, shine, and airborne. The gamma radiation from the direct, shine, and air-borne ganma sources greatly exceed the airborne beta in many areas of the secondary containment. Mhere the beta finite cloud dose (assuming a conser-vative equipment volume of 106CM3) is less than 10K of the radiation received to the equipment from direct, shine, and airborne gamma,.the beta contribution is considered insignificant and no further evaluation is required. In areas of the secondary containment where the infinite cloud beta contri-bution is greater, than 10K of the total integrated dose from the other sources, the equipment within the zone is evaluated to determine if the test or analysis level (i.e., qualification level) contained sufficient margin above the required level to account for beta. If this was not met, then the equipment details were reviewed to determine if the design and installation of the equipment would allow ventilation of its internal spaces to occur. This equipment is identified and the galena equivalent and volume adjusted beta radiation level (i.e., reduction factor of 10 plus a reduction factor dependent on internal volume) is added to the radiation levels from direct, shine, and airbodne gamma to form the total integrated dose for qualifica-tion. As beta effects are equipment specific, the beta values are not included in the radiation zone maps. The infinite cloud dose and various finite cloud doses based on volume for ventilated equipment in Reactor Building secondary containment are provided in summary form in Appendix B. Primary Containment Equipment important to safety in the primary containment will receive a beta dose several orders of magnitude higher than equipment located in the secon-dary containment. However, equipment important to safety in primary containment is designed and installed to function in conditions which include high pressure, mois-ture, and water spray. Watertight electrical conduit and non-ventilated NEMA 3 and NEMA 4 equipment enclosures are specified. Where required, cable entry ports are sealed using gualified Equipment Sealing methods. Where these conditions exist, beta is shielded from internal components and an evaluation of beta radiation is not required. Some equipment, even though not of a ventilated design, has the potential during long-term exposure to varying high pressure and beta conditions of receiving some beta exposure. Equipment important to safety within the pri-mary containment is reviewed and three potentially susceptable equipment types (terminal boxes with pressure relieving ports, continuous duty motors, and a few locations where exposed cabling was necessary) are identified. For these equipment types, detailed evaluation which included beta effects are performed. Beta Evaluation Status Evaluations have been completed for most equipment important to safety identified in the Justification for Interim Operation (Table A, gualified Status) contained in Appendix D. Completion of this scope of equipment will occur before fuel load. The remaining equipment important to safety, Table B and Justified components from Table A, will be evaluated as described above and any corrective actions completed prior to November 1985. Results of evaluations to date have not determined any cases where the equipment qualifications status has been affected by the Beta evaluation. 3.2.4 ~Fl oodin The top of the main vents from the drywell to the suppression pool are approximately 12 inches above the drywell floor. This is the maximum flood level since any excess water would drain to the suppression pool. No Class lE equipment or connections are located between the diaphram floor and the top of the downcomer vent pipes inside the wetwell except for the wetwell level system and wetwell temperature monitors which are totally enclosed in water tight conduit systems. As required by NUREG 0803 (Reference 15), the effects due to line breaks in the Control Rod Drive system have been evaluated. No safety related equip-ment would be submerged due to a break in this system (Reference 16). The possibility of flooding in the reactor building has been evaluated. The reactor building flooding analysis was completed (Reference 23 and Reference 25). The procedure used in this evaluation consisted of the following sequential steps:
- 1. Calculate the maximum source flow (gpm) for each room. Includes high energy and moderate energy lines.
- 2. Calculate the water depth using the following assumptions:
o Floor area used to determine water depth excluded floor open-ing areas, in order to allow for curbs and lips. o Twenty minute operator reaction time was allowed per NRC guestion 211.059. o No exit flow was considered.
- 3. > Perform a field walkdown to identify room exit paths (for drain-age) and equipment located below conservative flood level for impacted areas.
- 4. Re-calculate flood level considering water exit paths.
- 5. List equipment still flooded following recalculation.
- 6. Perform a safe shutdown analysis.
- 7. List flooded equipment required for safe shutdown.
- 8. Protect, relocate, or qualify equipment required for safe shutdown that is impacted by flood analysis.
Modification of the drain system for 4 MCC Rooms was required by the Mode-rate Energy Pipe Break Analysis. The results of this shutdown analysis for flooding (Reference 24) indicate that WNP-2 can be safely shutdown with alternate safety related equipment not affected by flooding. Based on this> flooding is not a required qualification parameter for WNP-2 safety related electrical equipment in secondary containment. The operator is alerted to flooding by leak detection sensors and can take action to initiate alternate safety systems if required. 3.2.5 Temperature/Pressure Outside Containment Class 1E equipment in the reactor building could be exposed to two postu-lated accident types: a LOCA/MSLB in primary containment or an HELB in the reactor building. These conditions were determined from References 5, 6, and 12. As explained in Section 4.0 of this report, equipment is evaluated to the worst accident environment in which it is required to function. A LOCA/MSLB in primary containment would cause an increase in the reactor building s temperature and humidity. The maximum conditions are presented in Profile 4 of Appendix B (Reference 21). The pressure/temperature effects of all postulated high 'energy line breaks in the reactor building were determined. Breaks in the following high energy lines were considered: 26" main steam line (envelops feedwater line break) 4" RCIC steam line 6" RWCU steam line 4" RWCV steam line 4" Auxiliary steam 3" Auxiliary steam Temperature/pressure profiles were developed for all areas that could be affected by these breaks. These profiles are presented in Appendix B. In addition, it was verified that pressure/temperature affects from moderate energy line breaks in these areas are less than the high energy line breaks. The accident profile due to a main steam line break in the steam tunnel'as determined from Reference 5. The remaininq temperature/pressure profiles in the reactor building were developed using the RELAP4 and COMPARE MODIA com-puter models (References 13 and 14). Detailed modeling of compartments and fluid flow paths were made. Heat sinks were modeled using appropriate heat transf er correl ati ons. for areas of the reactor building secondary containment where no high energy line breaks affects are postulated but moderate energy line breaks could be present, it has been determined that the pressure/temperature profile number 4 (effects in secondary containment of a LOCA/MSCB inside contain-ment) exceed the moderate energy pressure/temperature conditions. There-fore, equipment in the secondary containment is qualified to at least the pressure/temperature profile No. 4. 3.3 MILD ENVIRONMENT AREAS IN SECONDARY CONTAINMENT A mild environment is an area in which the maximum temperatures, pressures and humidity are not expected to chanqe significantly during or following design basis events. In addition, the cumulative radiation dose to equip-ment in these areas is below 104 rad (Reference 17). Some of the motor control center rooms in the reactor building are classi-fied as mild environments. These rooms are isolated and serviced by Class 1 HVAC systems so the temperature, pressure and humidity conditions will not change significantly. Also, the total radiation dose (normal + accident) in these rooms is less than 104 rad. Section 4.2.3 provides the Supply System's position on environmental quali-fication of safety related equipment in a mild environment. 4.0 QUALIFICATION METHODS The purpose of the equipment qualification evaluations is to ensure that all equipment important to safety wi 11 perform its safety function during its installed period when exposed to a harsh environment through which the equipment must operate or not fail. To accomplish this, the Class 1E equip-ment at WNP-2 was evaluated in accordance with the guidelines in NUREG 0588, Category II. The Equipment Qualification Reports in Appendix C summarize the evaluations that have been performed. Backup documentation and calcula-tions are contained on file at the Supply System's offices. 4.1 EQUIPMENT EVALUATIONS The following steps are involved in evaluating the qualification of the Class 1E equipment:
- a. Data Coll ection Available test data and analyses were sought for the Class 1E equipment. Data sources included the equipment vendors, the NSSS supplier (General Electric), the architect/engineer (Burns and Roe) and other utilities with the same equipment. Additionally, the Supply System is par ticipating in the generic qualification programs of the Electrical Power Research Institute and is a member of the EPRI Equipment Qualification Data Bank.
- b. Acce tance Criteria Definition The acceptance criteria to which Class 1E equipment qualification plans, tests and analyses are evaluated have been developed.
These criteria are based on NUREG 0588, Category II. The Supply System Engineering Procedure, titled "Acceptance Criteria for WNP-2 Safety-Related Equipment Qualification" (Reference 18), doc-uments the criteria that have been developed. Section 4.2 of this report highlights the major points of the acceptance criteria.
- c. Documentati on Revi ew The qualification data are evaluated to determine whether the equipment is qualified in accordance with the acceptance criteria.
Supplementary analyses are performed to complete the documenta-tion, when necessary. The Equipment gualification Reports in Appendix C sumaarize the evaluations that have been performed.
- d. Resolution of gualification Deficiencies In cases where insufficient documentation is available, requalifi-cation is initiated. The requalification method is chosen based on a number of factors, including the available test data, the severity of the accident environment and the complexity of the component. Evaluations, such as analysis of the materials of con-struction and failure modes and effects analysis, are performed when required. Replacement, testing, shielding and relocation are also used to resolve qualification deficiencies.
4.2 TECHNICAL APPROACH The technical approach, contained in Reference 18, was used to determine the qualification level of each component. This meets the intent of the guide-lines in NUREG 0588, Category II, and in many cases are more conservative. The selection of qualification methods is based on the severity of the acci-dent conditions and the function of the component. Two controlling types of harsh environments'ave been determined at WNP-2.
- 1. Severe Harsh environments This environment is created by a LOCA/MSLB inside containment and is characterized by high tempera-tures, high pressures, high radiation levels, steam conditions, 10OX relative humidity and possible demineralized water spray.
This condition is found only in the primary containment. These conditions in primary containment can produce harsher envi-ronments than would be present during normal operation in the sec-ondary containment. With the exception of the SGTS, Hydrogen Recombiner, and ECCS spaces which have high radiation levels during postulated LOCA conditions, the Reactor Building is charac-terized by moderate radiation, temperature and humidity levels. Significant changes in pressure and steam conditions would not occurs
- 2. Moderate Harsh environments This environment is created by a high energy line break outside containment and is characterized by high to moderate temperatures, steam conditions and increased humidity.
Neither prolonged high pressure nor high radiation are present in this environment. Flooding is described in Section 3.2.4 of this report. 1 In conformance with Appendix E of NUREG 0588, safety related equipment that must function during a harsh environment has been classified. The specific environment that this equipment will experience, and the accidents for which it is required to operate has been provided in the Equipment gualification Summary Sheets in Appendix C and Table A and B of Appendix D. 4.2.1 E ui ment Inside Primary Containment In the containment, where equipment will experience the direct effects of a LOCA, a rigorous set of criteria was established. This approach was taken due to the severe harsh environmental conditions that occur. A sequential or simultaneous test (aging, radiation, temperature/pressure under steam conditions) was a required element of the qualification. If this was not available, justification that the aging mechanism was not significant was required. The evaluation included verifying the estimated life, radiation exposure, steam temperature/pressure levels and duration were adequate to envelop the containment environmental service conditions. However, when test durations were less than the required period of operability, evalua-tions were performed to establish the test duration deficiency was ade-quately covered by a greater than required post LOCA test condition. In order to ensure adequate margin in the pressure/temperature testing, the GE generic profile shown on Profile 1, Appendix B (first 24 hours) was used. Plant specific profile beyond 24 hours was used as per NUREG 0588. The test results were reviewed to verify that the component met its required perfor-mance characteristics before, during, and after testing. 4.2.2 Equipment Inside the Reactor Building (Secondary Containment) For equipment in the Reactor Building secondary containment harsh environ-mental areas, analysis of material thermal and radiation capability was allowed. In all cases, elevated temperature testing with steam conditions or high relative humidity testing data was available to demonstrate the components'apability to the temperature, steam and humidity conditions. However, documentation addressing the components capability to withstand the radiation levels was generally lacking. Therefore, an evaluation to verify that the material functional threshold levels were greater than the service conditions was performed to supplement the documentation. For equipment that contains sensitive transister and integr ated circuit solid state com-ponents radiation testing was required. The functional radiation threshold for a component was based on the materi al and functions of each non-metallic part. The applicable material property (i.e., compression set, elongation, etc.). was considered. In some cases the material functional threshold was found to be greater than the radiation level that first causes a noticeable change in the material (threshold level). These cases were generally static applications such as gaskets and O-rings. Mater ial handbooks were consulted to 'determine the humidity susceptability of selected materials of construction such as gaskets and 0-rings. Test data was required for nonsealed electrically energized parts such as motor windings and solenoid coils. Sequential radiation testing in conjunction with the steam line break was not a required element. Material radiation effects evaluation as described previously was allowed. This approach is acceptable because it is not required to postulate that both the HELBs in secondary containment and LOCA/MSLB in primary containment occur simultaneously. Therefore, the steam conditions and the radiation conditions would not occur simultaneously as they are produced by separate accidents. 4.2.3 Equipment Located in. Mild Environments A mild environment is defined to be an environment that would be no more severe than would occur during normal power plant operation or during antic-ipated operational occurrences. Class 1E and safety related mechanical (SRM) equipment located in mild or benign environments satisfy general quality and surveillance requirements applicable to safety related equipment, including 10CFR50 App'endix B. The Class lE and SRM equipment pur chased and documented under the above quality requirements satisfy the environmental qualification requirements for safety related equipment located in mild environments. Electrical and electronic equipment in the Reactor Building MCC rooms that contain solid state electronic equipment are evaluated for radiation induced damage even though the radiation environment is less than 104 rad TID. Results of this review conclude that these components are qualified to the required TID. 4.2.4 ~Mar tn Margin, or conservatism, is added to the aspects of the equipment qualifica-tion procedure. This is done to account for normal variations in commercial production of equipment and reasonable errors in defining acceptable perfor-mance. 22
The qualification requirements were established using conservative assump-tions and analytical procedures. The reactor building thermal hydraulic profiles have been developed using conservative computer codes. The required radiation doses were developed using conservative source terms, as discussed in Section 3.2.3 of this report. A minimum test time of one hour was used for equipment that is required to perform its safety function within a short time into the event and, once its function is complete, subsequent failures are not detr imental to plant safety. In some cases, an instrument senses a parameter below the peak condition of that parameter during the accident. A system and component function evaluation was performed to determine subsequent failure would not affect plant safety. Testing for greater than one hour was performed but not to the peak conditions. These cases are identified in Appendix C. 4.2.5 A~~in The purpose of evaluating equipment aging is to assure that equipment will perform its safety function in an advanced life state during or following the hostile environment of a LOCA/HELB. The program developed by the Supply System addresses this issue within the context of current aging technologies. The WNP-2 Equipment gualification Program addresses aging as required by NUREG 0588, Category II. Valve operators committed to IEEE-382, 1972 and continuous duty motors committed IEEE-334-1971 located in a potentially harsh environment address aging through methods consistent with the Cate-gory I aging criteria of NUREG 0588. For other equipment important to safety located in potentially harsh envi-ronments, the WNP-2 qualification program addresses aging by identifying age sensitive components within the equipment. These components are evaluated and an estimated life is determined. The estimated life is based on current analytical techniques (Arrhenius model, 10 C rule, temperature index, etc.) and accelerated aging test data when available. The results of the WNP-2 Equipment gualification aging evaluations are inputed to the Plant Maintenance and Surveillance Program (see Section 4.5). 4.2.6 Instrument Accuracy Instrumentation performance may vary due to exposure to accident condi-tions. Instrumentation may demonstrate a change in output not related to input variation due to environmental exposure beyond its design rated envi-ronmental conditions. During qua1ification testing, performance character-istics such as accuracy are measured to determine the degree of accuracy variation under accident conditions. Dependent on the application of the instrumentation in the plant, the required accur acy for an instrument type may vary significantly. The WNP-2 gualification Program includes comparing the required accuracy (dependent on system application) to the accuracy demonstrated during quali-fication testing. Supplemental analysis may be used when the test environ-mental condit'ion greatly exceed the plant specific environmental requirement in assessing adequacy of'he instrument performance during test. Instrument Accur acy Status Required accuracy has been determined for those instruments identified in Table A of the JIO. Where a significant change in instrument accuracy under accident conditions could affect plant safety, the required accuracy per-formance under accident conditions and the demonstrated accuracy are pro-vided in Appendix C for the instruments statused as qualified (g). 4.2.7 Interface Oualification (Reactor Buildin Secondary Containment And Primar Containment Reactor Buil'ding Secondary Containment Electrical interface connections to instruments important to safety in the secondary containment were evaluated. Liquidtight installation as defined by ANSI/NFPA 70 was used. This type of electrical interface protects against liquid and vapor intrusion into the electrical instrument housing. Postulated high energy line breaks in the reactor building secondary con-tainment result in short term local elevated pressure conditions. Modeling of the postulated break conditions show that the short duration pressure pulse is generated due to air displaced by the break prior to steam exposure to the instrument. The pressure pulse is quickly relieved to neighboring open areas of the reactor building which is maintained at a negative atmos-pheric pressure level. During the postulated HELBs, there is no prolonged elevated pressure to drive the steam or moisture into the equipment hous-ing. Therefore, liquidtight connections are adequate to prevent steam and moisture intrusion into instrument housings. gualification testing has been performed to assure the liquidtight flexible conduit can perform under the steam conditions for the duration of the HELBs. Documentation of this has been reviewed and is being obtained. Primar Containment The method of electrical interface for equipment important to safety located in the primary containment of WNP-2 has been reviewed. Based on this review it has been determined that the interface method employed is not adequate. Corrective action has been identified and a program for implementation is in place. The final installation will employ the use of equipment seals at the equipment/conduit interface except for installations where the conduit run enters from an elevation lower than the equipment's elevation. Flexible conduit, qualified for pr imary containment conditions, will be installed when this field condition exists. For installations where the conduit run enters the equipment from above a qualified seal assembly will be installed to establish a moisture block at the conduit/equipment interface a final field verification will be performed for equipment important to safety located in primary containment to assure implementation of the corrective action prior to fuel load. This corr ective action will be closed as part of a 10CFR50.55e notification. 4.3 SAFETY RELATED MECHANICAL EQUIPMENT (HARSH'ENVIRONMENTS) General Design Criterion Four requires that plant equipment be designed to accommodate the effects of and be compatible with the environmental condi-tions associated with normal operation, maintenance, testing, and where applicable, postulated harsh environments including loss of coolant accidents. The procurement of appropriately specified and qualified mechanical equip-ment is part of the overall procurement program for the Washington Nuclear Power Station Unit 2. Mechanical equipment specifications were based on the environmental conditions of the GE Generic Specification 22A3008 (Rev. 5). However, post-TMI NUREG-0737 proposed the use of more severe harsh environ-mental conditions and these have been utilized in the WNP-2 Electrical Equipment Environmental Qualification program. These environmental condi- 'tions are used for the mechanical equipment review. Mechanical equipment has not generally demonstrated the sensitivity to radi-ation exposure that electrical components have. The metallic portions of the equipment are particularly resistant to radiation. Nonmetallic parts of the mechanical equipment, while more sensitive to radiation and temperature, are subject to planned maintenance, which assures adequate performance under a harsh environment for the required function of the component. This section summarizes the environmental design conformance review of active safety related mechanical equipment located in the harsh environments as identified in the "Safety Related Mechanical Equipment List" contained in Appendix A-2, for the Washington Nuclear Power Station Unit No. 2. Equip-ment reviewed includes: pumps, fans, motor-operated valves (valve only), air-operated valves, safety and relief valves, check valves, and dampers. Methodolo y The review consisted of a six step process as follows: Identification of Active Safety Related Mechanical (SRM) Equipment Identification of Environmentally Sensitive Subcomponents Identification of Environmentally Sensitive Conditions Identification of Environmentally Sensitive Component Material Capabilities Aging Analysis Evaluation of Environmental Effects Identification of Active Safety Related Mechanical Equipment In order to identify active safety related mechanical (SRM) equipment that would be evaluated under this program, a detailed review of the SRM list was made. All active safety related pumps, fans, motor operated valves (valve only), air operated valves, safety and relief valves, check valves, dampers, located in the Containment, Reactor Building and Steam Tunnel were included in this program. A compilation of this equipment is presented in Appendix A. Identification of Environmentally Sensitive Subcomponents The applicable revision of the manufacturer's bills of material for the equipment to be evaluated was reviewed and a nonmetallic subcomponents list was prepared for each group or piece of equipment. Where necessary, addi-tional information was obtained from the equipment vendor to explain specific trade name materials, or incomplete/unsatisfactory material identification. 27
Identification of Environmental Service Conditions The environmental conditions used for the review of mechanical equipment are presented in Appendix B of this report. The parameters considered are: radiation temperature pressure chemical spray humidity In addition, the fluid conditions (temperature, pressure, radiation fluid chemistry) processed by SRM equipment is considered in the environmental qualification . The qualification of this equipment also considers the indi-vidual equipment operating time required for performance of the safety func-tion. Chapter 3 of the FSAR contains the design requirements for SRM equip-ment handling process fluids. Where process fluid conditions are equal or more severe than the accident environmental service conditions, the normal design process condition sufficiently documents the environmental qualifica-tion of this SRM equipment and further evaluation is not required. SRM equipment that must function during an accident and are located in areas susceptible to LOCA and HELB effects are qualified through a materials analysis in addition to the surveillance and maintenance program activi-ties. The equipment and materials analysis identifies any susceptibility to radiation, temperature, pressure, humidity and chemical spray. Any effects from the above are factored into the surveillance/maintenance program. Where the analysis indicates a potential subcomponent failure, a failure modes and effects analysis was performed to determine the subcomponent fail-ure impact on the equipment's ability to perform its intended safety function. Identification of Environmentally Sensitive Component Materials and Their Capabilities Environmentally sensitive materials were identified based on the material 's threshold radiation level and maximum design service temperature. The threshold radiation is the lowest radiation exposure at which a property change in the material is documented. The maximum design service tempera-ture is the maximum steady s'tate temperature a material can be subjected to without loss of function. Materials handbooks, textbooks, industry and government reports were researched in order to obtain material data. In some cases, vendor data was utilized to supplement the above sour ces. Thermal A in Analysis Thermal aging calculations based on Arrhenius methods were performed for elastomeric'nd plastic materials utilized in SRM equipment. The test data required for the thermal aging calculations was obtained from industry pub-lications and test reports. These calculations were used to determine the maximum recommended service life of a subcomponent utilizing these materials. The maximum recommended service life of a material was calculated by deter-mining the base life of the material at normal plant operating temperatures and subtracting the equivalent effect of the most limiting normalized ther-mal transients. The transients considered incorporate an BoC margin over the postulated peak temperatures for additional conservatism. Each material is analyzed for service in the Primary Containment, Reactor Building and Steam Tunnel environments. The results of these calculations in conjunction with manufacturer recommen-dations and accepted industry standards are used to recommend change out intervals for the subcomponents under consideration. In some cases however, a material displays excellent thermal characteristics but is known to degrade significantly via 'another environmental parameter (typically radiation). Discussions related to such cases are considered in the Engineering Analysis Sheets (EAS) or Subcomponent Data Sheets (SDS) for the equipment under consideration. Evaluation of Environmental Ef'fects The equipment and materials analysis was performed as follows:
- 1) Nonmetallic materials (seals, gaskets, lubricants, hydraulic fluids, etc.) were analyzed for susceptibility to the normal and harsh environmental parameters, mentioned previously. The methods used in this analysis consist of an analytical comparison of the postulated environmental conditions with the material's maximum service temperature, pressure rating (if applicable), threshold radiation resistance, and susceptibility to chemical spray and humidity.
Those items which are shown to be unacceptable based on the initial comparison are evaluated in further detail considering: degree of material degradation
- b. material properties affected co equipment/component function
- d. degree of functional degradation
- 2) Nonmetallics (valve packing, O-rings, etc.) exposed to process fluids (primary water, etc.) were evaluated for their ability to operate properly under the harsh environments. The evaluation consisted of a review to assure that the design specifications and manufacturer's documentation demonstrated that the component was compatible with the postulated environment. Where this did not demonstrate acceptability, further analysis was undertaken.
f 'I
- 3) Active metallic components (valve stems, pump shaft) were evalu-ated with respect to susceptibility to corrosive attack due to exposure to moisture and spray.
- 4) Flooding - The top of the main vents from the drywell to the sup-pression pool are approximately 12 inches above the drywell
floor. This is the maximum flood level since any excess water would drain to the suppression pool. No safety related mechanical equipment is located between the diaphragm floor'and the top of the downcomer vent pipes inside the wet well. The possibility of flooding in the Reactor Building has been eval-uated. The results of the shutdown analysis for flooding indicate that WNP-2 can be safely shutdown with alternate safety equipment not affected by flooding in the Reactor Building. Based on this submergence is not a required qualification parameter. Acceptance Criteria In order to be considered qualified, environmentally sensitive subcomponents of SRH equipment must either: a) be shown to be acceptable for the plant environment by exhibiting threshold radiation values and maximum service temperatures above the maximum postulated environmental conditions, or b) be shown to be acceptable for the plant environment by analysis that demonstrates that the safety function of the component is not compromised. Documentation and Results The Equipment Qualification Report Summary Sheets in Appendix C-2 summarize the qualification evaluations that have been performed. The back up docu-mentation used to demonstrate that each Equipment Piece Number is qualified for its application and meets its specific performance requirements is con-tained in Qualification Information and Documentation (QID) Packages that are on file at the Supply System. The results of this review demonstrate that the environmental effects due to normal operation and postulated harsh environment will not result in a degree of degradation of nonmetallic subcomponents that, will prevent the equipment from performing its active safety function. 4.4 DOCUMENTATION The Class 1E Equipment gualification Reports in Appendix C summarize the qualification evaluations that have been performed. Tests, analyses and other documentation used to demonstrate that each component is qualified for its application and meets its specific performance requirements are on file at the Supply System. 4.5 MAINTENANCE AND SURVEILLANCE A comprehensive Maintenance and Surveillance Program has been developed for WNP-2 plant equipment. This Program is based on the recommendations of Regulatory Guide 1.33 Rev . 2, "guality Assurance Programs Requirements (Operational)". Procedures have been developed which implement surveillance testing and scheduled maintenance activities. The WNP-2 Scheduled Maintenance System (SMS) Program consists of periodic inspections, tests, and work items designed to improve component reliabil-ity. The Power Plant Information and Control System (PPICS) computer is utilized as the Data Base Management System for SMS. The Master Equipment List from which the Class 1E and Safety Related lists are derived forms the data base. Also a part of this system is the Equipment History Program for historical data storage. These elements of the SMS Program provide the necessary tools to assure the following is achieved: o Timely replacement of equipment with an estimated life less than the power plant life. o Performance of maintenance required to sustain qualification of installed equipment. o Performance of surveillance to identify equipment in a deterio-rated condition. o Proper performance of maintenance activities by maintenance personnel. Input to the SMS Program will be made which defines specific maintenance required as a result of the Equipment gualification Program. The Mainte-nance department will evaluate this input along with their experience with similar equipment, manufacturer's maintenance and surveillance recommenda-tions, and industry/NRC notices such as NOMIS, IEBs, IENs, IECs, NPRD, etc. From these sources an equipment maintenance and replacement schedule will be determined. Review and documentation of any replacement or maintenance schedule nonconservative with the recomnendations from the Equipment guali-fication program results is required. Modification of the qualification file with proper justification and documentation will be made. The frequency and scope of the surveillance and maintenance programs will be reviewed throughout the life of the plant to identify any abnormal degrada-tion. Adjustments to maintenance, surveillance, and replacement schedules will be made based on experience gained through service and evaluation of malfunctioning equipment where necessary. The Supply System intends to per-form analysis to identify trends associated with deterioration of equipment. gualification of Replacement Equipment Replacement equipment procured for WNP-2 will be purchased to comply with the intent of 10CFR50.49(l) which states, "Replacement equipment must be qualified in accordance with the provisions of this section unless there are sound reasons to the contrary." A reasonable effort will be made to procure in kind equipment which are qualified to the requirements of 10CFR50.49. When this is not feasible, replacements in kind will be used which, at a minimum, preserve the present qualification status of equipment at WNP-2 (NUREG 0588, Category II). If it is not feasible to procure in kind replacements, then a reasonable effort will be made to procure engineering approved substitutes to the requirements of 10CFR50.49. If it is not feasible to obtain such replace-ments then engineering approved substitutes will be procured to the present qualification status of equipment at WNP-2 (NUREG 0588, Category II). Subcomponents In kind spares will be used when available. Subcomponents used as identical spares for existing qualified equipment will be qualified to Category II as a minimum, by virtue of qualification of the related component. If an identical spare is not employed, an engineering approved substitute, which does not degrade the qualification of the original, may be used . 34
5.0 QUALIFICATION RESULTS The environmental qualification of the components identified on the Class 1E Equipment List and active Safety Related Mechanical List (Appendix A) has been ev'aluated. The status of the evaluations is presented on the Equipment gualification Reports in Appendix C. A justification for interim operation pending completion of the full Environmental gualification Program is com-plete (see Section 6 and Appendix D). The Class 1E List (Appendix A) identifies the safety related electrical equipment, along with its respective qualification status. The available qualification documentation has been obtained and reviewed for this equip-ment. The reviews, supplemented by engineering analyses, have determined that most of the components meet the intent of NUREG 0588, Category II. In some cases, it has been determined that there is insufficient documentation to support complete qualification . The method for completing the qualifica-tion is included on the individual Equipment gualification Reports in Appen-dix C. This equipment will be qualified by November 30, 1985. The following closure activities are being completed for equipment important to safety identified in Table "A", Appendix D (i.e., that equipment required to be qualified or justified prior to fuel load). Field verification that equipment whose corrective action is replacement with qualified equipment has.been accomplished prior to fuel load. o Completion of corrective action of equipment interface has been accomplished prior to fuel load for primary containment equipment. o Receipt of qualification documentation and completion of /ID files (g'tatus). o Completion of beta effects evaluation of secondary containment equipment prior to fuel load. o Final verification of pressure/temperature profiles for reactor building secondary containment areas. The following activities are being completed for equipment important to safety identified in Table 8 and that equipment statused as "J" in Table A of Appendix D (i.e., that equipment that must be qualified by no later than November 30, 1985). o Development of a detailed plan that assures final qualification of all equipment important to safety by November 30, 1985. On-going activities that will continue throughout the life of the plant include: o Review of Design Changes to assure qualified equipment is procured and installed in accordance with 10CFR50.49 as implemented by the WNP-2 equipment qualification program. o Continuing input to the scheduled maintenance system for equipment replacement or required maintenance based on the Equipment guali-fication Program results. o Continuing review and modification of the SMS program based on results from Maintenance/Surveillance activities. o Review of spares and replacement equipment to assure qualification level is maintained and upgraded where necessary. o Continued participation in and review of industry efforts in the equipment qualification area and review of NRC notices, circulars and bulletins as they are received with respect to the equipment qual ifi cati on. 6.0 JUSTIFICATION FOR INTERIM OPERATION To obtain an operating license for WNP-2, the Supply System has been noti-fied (Reference 2) that safety-related electrical equipment shall be review-ed using NUREG 0588, Category II, "Interim Staff Position on Environmental gualifications of Safety-Related Electrical Equipment", as the basis for determining the adequacy of the safety-related equipment's documentation. Furthermore, the NRC staff has informed (Reference 22) the Supply System that where there are deficiencies, the Supply System should commit to cor-rective action consistent with the requirements to establish qualification. If fuel loading occurs before complete qualification can be obtained, justi-fication for operation until corrective actions are completed must be provided. In addition, the NRC Staff's proposed final rule 10CFR50.49, regarding envi-ronmental qualification of safety-related electric equipment, states that "the applicant for an operating license shall perform an analysis to ensure that the plant can be safely operated pending completion of environmental qualification ." For WNP-2, the Equipment gualification Program is in process and, as demonstrated by this report, many components have been shown quali-fied by existing documentation . However, it is unlikely that all safety-related electrical equipment will be fully documented before desired full power operation of WNP-2. Therefore, a Justification for Interim Operation (JIO) has been performed. It is concluded that, upon documentation of the qualification of a minimum set of safety-related electrical equipment, WNP-2 can be safely operated pending completion of the Environmental gualification Program for all safety-related electrical equipment. The basis for this conclusion is in establishing the ability to accomplish the following six safety functions with a minimum set of safety-related electrical equipment.
- 1. Emergency Reactor Shutdown
- 2. Containment Isolation
- 3. Reactor Core Cooling 4~ Containment Integrity
- 5. Core Residual Heat Removal
- 6. Prevention of Significant Release of Radioactive Material to the Environment The methodology and results of this JIO analysis are provided herein (Refer-ence Appendix 0, Justification for Interim Operation Report). The following are the main elements of the analysis.
o Accident Definition The accidents that potentially result in a harsh environment inside the primary containment or reactor building were identi-fied. Eight types of accidents were identified: three Loss-of-Coolant Accidents (LOCA), four High Energy Line Breaks (HELBs), and the Control Rod Drop Accident (CRDA). The environmental con-ditions (pressure, temperature, humidity, and radiation) associ-ated with the LOCA and HELB accidents, and the areas of the plant affected, were determined. LOCA radiation profiles were used to envelop the CRDA conditions. o Safety Sequence Analysis For each postulated accident, Safety Sequence Diagrams (SSDs) were prepared. The SSDs identified the. systems required to mitigate each accident, shut down the reactor, and maintain it in a safe n condition by accomplishing the necessary safety functions. o Saf ety-Rel ated Equi pment The safety-related electrical equipment for the systems identified in the Safety Sequence Analysis was taken from the safety-related equipment list. 0 o Failure Nodes and Effects Analysis A Failure Nodes and Effects Analysis (FNEA) was performed for all equipment in the systems identified by the SSA. As a result of the FMEA, a list of equipment that need not function to achieve the six safety functions was defined. All equipment whose failure was determined to have no adverse effect on plant safety or acci-dent mitigation need not be qualified for any accident environ-ment, but it will be qualified for its normal service environment. The analysis also determined the technical basis for classifying the equipment that need not be environmentally qualified for a harsh environment. o Selection of Minimum Required Equipment Safety-related electrical equipment, in the systems identified by the SSA, that is required to operate to achieve the six safety functions, or must not fail in a manner detrimental to the six safety functions, was evaluated. For all evaluated accidents, a minimum set of this equipment .in a single shutdown path to achieve the required safety functions was identified . This minimum set of equipment wi 11 have documentation of environmental qualification or adequate justification prior to fuel load of WNP-2. The docu-mentation of this minimum set of safety-related electrical equip-ment ensures that the required safety functions will be achieved for all evaluated LOCAs, HELBs and CRDAs that potentially result in a harsh environment. The safety-related electrical equipment not requiring documenta-tion prior to fuel load, which include redundant or diverse systems, fuel pool cooling, and Control Room chillers, will be documented to establish its qualification prior to Hovembei 30, 1985. 7.0
SUMMARY
This document summarizes the evaluation of environmental qualification of Class 1E equipment in WNP-2, performed in accordance with NUREG 0588, Cate-gory II. It provides a summary of the Environmental gualification Program that has been undertaken by the Supply System. The program ensures that electrical equipment important to safety and safety-related mechanical equipment will perform its safety-related function during normal, abnormal and postulated accident conditions. The program includes the following: o Normal, abnormal and accident service conditions in primary con-tainment and the r eactor building (harsh environments) have been defined. o Electrical Equipment important to safety and safety-related mechanical equipment has been identified by the tag number along with its required safety function, use, and required operating time. The location and manufacturer's data for this equipment have been determined. o The qualification reviews and status of this equipment has been determined. o Corrective actions are being taken to resolve qualification docu-mentation deficiencies. o A Justification for Interim Operation of WNP-2 pending completion of the Environmental gualification Program has been completed. o Continued development and implementation of the maintenance/ survei llance program to address the results of the gualification Program. o A replacement part program is in place to assure the highest level of qualified equipment available.
8.0 REFERENCES
I. NRC Office of Nuclear Reactor Regulation, "Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment", NUREG 0588, Rev. l.
- 2. NRC Division of Project Management, "Qualification of Safety-Related Electrical Equipment", letter from D. F. Ross (NRC) to Operating Licenses Applicants, February 1980.
- 3. The Institute of Electrical and Electronics Engineers, Inc. (IEEE),
"IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations", IEEE Standard 323-1974.
- 4. EDS Nuclear Inc., "Review of 1E/1M Equipment Lists for Safety-Related Systems", Project Instruction No. 7, Job No. 1140-001.
- 5. General Electric Environmental Design Specification. No. 22A3008, Revision 5, April 1977.
- 6. Washington Public Power Supply System, "WNP-2 - Final Safety Analysis Report".
- 7. NRC Office of Nuclear Reactor Regulation, "Clarification of TMI Action Plan Requirements", NUREG 0737, Rev. 0, October 31, 1980.
- 8. ORIGEN, "Isotope Generation and Depletion Code", RSIC Computer Code Collection, Oak Ridge National Laboratory, Updated September 27, 1979.
- 9. NRC Office of Nuclear Reactor Regulation, "Technological Bases for Models of Spray Washout of Airborne Contaminants in Containment Vessels", NUREG CR-0009, 1978.
- 10. QAD-P5A, "Point Kernel General Purpose Shielding Code", RSIC Computer Code Collection, Oak Ridge National Laboratory.
- 11. EDS Nuclear Inc., Zone Dose Calculations, Series 0740-004-xxx.
- 12. Washington Public Power Supply System Calculations:
NE-02-81-06-0 NE-02-81-13-0 NE-02-81-17-0 NE-02-81-07-0 NE-02-81-14-0 NE-02-81-18-0 NE-02-81-08-0 NE-02-81-15-0 NE-02-81-19-0 NE-02-81-09-0 NE-02-81-16-0 NE-02-81-20-0
- 13. Idaho National Engineering Laboratory, "RELAP4/MOD5, A Computer Program for Transient Thermal Hydraulic Analysis of Nuclear Reactors and Related Systems", Volumes I and II, ANCR-NUREG 1335, September 1976.
- 14. COMPARE MODlA - NUREG/CR-1185.
- 15. NRC Office of Nuclear Reactor Regulation, "Generic Safety Evaluation Report Regarding Integrity of BWR Scram System Piping", NUREG 0803.
- 16. Washington Public Power Supply System, "Supply System Response to NRC SER Issue - Pipe Breaks in BWR Scram Discharge Volume", Memo R. 0. Vosburgh to C. D. Taylor, December 30, 1981.
- 17. EPRI, "Radiation Effects on Organic Materials in Nuclear Plants",
Report NP-2129, Project 1707-3, November 1981.
- 18. Washington Public Power Supply System, "Acceptance Criteria for WNP-2 Saf ety Related Equi pment qua 1 ifi cat i on", TDP 3. 32.
- 19. WNP-2 Final Shielding Evaluation Report, September 1982.
- 20. Washington Public Power Supply System Calculation NE-02-82-39-0.
- 21. Washington Public Power Supply System Calculation NE-02-81-14-0.
- 22. NRC Office of Nuclear Reactor Regulation, "Safety Evaluation Report",
NUREG-0892, March 1982.
- 23. Flooding Analysis, BER Calculation 5.51.055.
- 24. Shutdown Analysis for Flooding, BIIR Calculation 5.51.058.
- 25. Moderate Energy Systems Pipe Break Analysis, BIIR Calculation 5.51.054.
- 26. Washington Public Power Supply System Calculation NE-02-85-08.
APPENDIX A CLASS 1E EQUIPMENT LIST AND SAFETY RELATED MECHANICAL LIST
Appendix A contains the following information: o Class lE/SRM List Users Manual: A description of the use fields and abbreviations on the Class 1E List o System Code List: A list of system abbreviations used on the Class 1E Equipment List o Component Table: A list of the component abbreviations used on the Class 1E Equipment List o Class lE Equipment List o Safety Related Mechanical List
WNP-2 Class 1E and SRM Equipment List Users Manual: Descr iption of codes used on the Class lE List and SRM List Column Designation Descri tion
- 1. CONTRACT The contract under which the equipment was purchased.
The contracts beginning with 02 and Contract 59 were with the NSSS supplier. The two-digit contracts are for equipment purchased through our A/E and the three-digit contracts indicate equipment purchased through contrac-tors at the construction site.
- 2. COMPOSITE NO. The composite, such as instrument rack or valve, on which a component is located.
- 3. E(UIPMENT NO. The equipment piece number (EPN) is listed. It is com-posed of the system designation (a complete list is enclosed), a component code ( list enclosed) and a unique identif i er.
- 4. MFG Manufacturer: Contains the code prepared for the indus-try by Southwest Research Corporation indicating the com-pany who manufactured the equipment. In a few cases where the manufacturer has not been determined, the sup-plier's code was put in this column until the manufac-turer has been determined.
- 5. MFG MODEL NO. The manufacturer's model number. In the cases where this has not been determined, General Electric purchased part drawing number or other applicable information is supplied.
A.2
- 6. Q.I.D. The Qualification Identification is a six-digit number indicating a file which contains all the qualification documentation for that EPN along with summary forms and plant walk-through records.
- 7. LV Level assigned to equipment. An identifier which will permit the sorting of the 1E/SRM list into moor pieces of equipment, instrumentation and subcomponent parts.
Level 1: Class 1E/SRM composite equipment which requires qualification of the overall assem-bly. Each composite piece of equipment will be identified with a unique Equipment Piece Number (EPN) and will'have the symbol "+" added to the end of the EPN. Motor operated valves would be listed as composite equipment with a level designation of l. Other examples would include the diesel gener-ator skids, pump skids, air handling units, filter/dryer assemblies, air compressors, etc. Level 2: A Class 1E/SRM component or instrument func-tion which requires individual qualification. The instrument function is described by an instrument loop which could include a sensor, a switch, an alarm, an indicator and/or a con-troller. Whenever an instrument loop is iden-tified as Safety-Related, the sensor will receive a Level 2 designation and all other instrument loop components will be designated Level 3. A.3
Example 1: For a motor-operated valve, the valve body, valve motor, and external limit switches (if they have a Safety-Related function) are all Level 2 components. Example 2: An instrument consisting of a flow element, flow transmitter, flow switch and flow indi-cator would have the flow element as Level 2 with the other components as Level 3. Level 3: Any 1E/SRM instrumentation component not included in Level 2. Example: A flow transmitter associated with a 1E/SRM flow element would be designated as Level 3. Level 4: A subcomponent of a class 1E/SRM component. Example: Internal limit switch to motor operators for valves, dropping resistors, pressure trans-mitter circuit boards, wiring, indicating lights, etc.
- 8. Safety The Class 1 action that a piece of equipment or a system Functi on is required to perform or monitor that makes it Safety Related.
A component may provide one or more of the safety func-tions listed below.
~S>bol Function A. Emergency Reactor Shutdown including SCRAM Signals and Reactivity Insertion .
A.4
~Sbol Function B. Containment Isolation Bl Primary Containment B2 Reactor Building C. Emergency Core Heat Removal D. Containment Atmosphere Control E. Core Residual Heat Removal, including Long-Term Cooling F. Prevention of the Release of Radioactive Material to the Environment G. No Active Safety Function but a Passive Mechanical Integrity Function G1. No Active Safety Function but a Passive Electrical Integrity Function G2. No Active Safety Function but Both a Passive Electrical and Mechanical Integrity Function H. Emergency Electrical Power Systems, AC and DC. Instrumentation to Follow the Course of an Accident Compartment Heat Removal for Equipment Oper-ability or Personnel Habitability L. Necessary for Mitigation of Rod Drop Accident A.5
- 9. PLANT LOCATION The location of the component within the plant by build-ing, elevation and coordinates.
- 10. Q.S. Qualification Status (second column) indicates the envi-ronmental qualification of the equipment. The following list shows the meaning of the codes used.
A - Acceptable, thermal aging completed B - Acceptable, thermal aging being covered by surveillance C - Acceptable, not installed D - No documentation in files G - Being requalified by modification of the hard-ware or the environment M - Being requalified by analysis N - Not Acceptable P - Purchasing qualified replacement R - Documentation not complete T - Being requalifi ed by test The first column shows the seismic qualification status.
- 11. F/0 HOURS The time, in hours, a component is required to function following an accident.
- 12. EQUIPMENT A description of the equipment function.
DESCRIPTION
- 13. DRAWING The plant PAID on which the component appears.
- 14. USE Contains codes which describe equipment use during acci-dent and/or normal plant shutdown conditions. The USE field is based on Item 2 Appendix E of NUREG 0588.
The "USE" input field is a two-digit field. The first digit shows the equipment operability requirement for accident mitigation and the second shows the equipment operability requirements for Hot or Cold shutdown conditions. X X The equipment is not required before, during or after an accident. Example: Equipment in this category provides no active function, but may provide a passive function by containing radioactive material outside the Reactor Building. It need not be qualified to demonstrate operability, even under non-accident service environments. Equipment that will experience the environ-mental conditions of design basis accidents for which it must function to mitigate said accidents, and that will be qualified to demonstrate operability in the accident envi-ronment for the time required for accident mitigation with safety margin to failure. A.7
Equipment in this category is required for accident mitigation of accidents analyzed in the FSAR. This includes: pumps, valves, electrical equipment, instrumentation to follow the course of an accident, etc. Equipment will experience environmental condi-tions of design basis accidents through which it need not provide an active function for mitigation of said accidents, but through which it must not fail in a manner detrimental to plant safety or accident mitigation, and that will be qualified to demonstrate the capability to withstand any accident environ-ment for the time during which it must not fail with safety margin to failure. Equipment in this category must not actively fail in a manner detrimental to plant safety, e.g., a motor operated valve that is normally shut would be categorized as a "2" if its inadvertent opening would be detrimental to plant safety. Equipment that provides only a passive integrity function on a potentially contaminated system will be categorized as a "2" and will have a "G" placed in the "EC" column. Category 2 will include all manual boundary, integrity, test and root valves which may be exposed to post-LOCA and radioactive drain systems components (FDR and EDR).
'A. 8
Equipment that will experience environmental conditions of design basis accidents through which it need not function for mitigation of said accidents, and whose failure (in any mode) is deemed not detrimental to plant safety or accident mitigation, and need not be qualified for any accident environment. Example: Equipment in this category is limited to the 1E/SRM equipment in the "harsh environments" which is Safety-Related only to prevent the release of radioactive material and will not be exposed to post-LOCA radioactive fluids. This category will include the components of the Reactor'Water Clean-up System downstream of the second containment isolation valve. Equipment that will not experience environmen-tal conditions of design basis accidents and that will be qualified to demonstrate opera-bility under the expected extremes of its accident service environment. This equipment would be located in isolated NCC rooms in the Reactor Building and areas outside the Reactor Building. Second Digit X X The equipment is not required to operate to shut down the plant during normal conditions. A.9
The equipment is required to operate for Hot Shutdown only during normal plant conditions. The equipment is required to operate for Cold Shutdown only during normal plant conditions. The equipment is required to operate for both Hot Shutdown and Cold Shutdown during normal conditions. EC Equipment Classification. Code A - Active for Class 1E equipment, active components must provide or receive an electrical signal in order to perform its safety-related function. A.10
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POSITION SMITctl 02C12 1 0
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H52tti *=<< J>>I R". 528 LS/a.R C4 I i jf% I g i ~ CRO HCU"0227+ e,;-.> "i'll 'agf2.;, CRO".POS."4'2<OPi=-.".. POSIi foN %kifcH,.'.- ':;- V:"- -;=
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A POSITION SMITCH R 528 LD/Jt ~ 4 CRO-HCU-0631+ 02012 3 4 I 0 A H520 r4 cRD pos+1260630 =" H302, DZE6 2R)72 248003 612 POST it Ptt, SMI TCtt .R 528 Ke/0 ~ 4 A D N ~ ~ CRD" 617 tttl/~0635'RO-POS 1260639 H302 HZE6-2RtJ72 A 8 24ano3 612 N ~ 17 02C12 H528 POSI'Tlott SMIT'CH 8 52 tt K 2/P ~ 'I CRD-HCU-0643> 02C)7 A I 0 R H5 2 Jt r4
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~t t EPN HFG S E 010. TH HL TEST ANL FO C,
'ODCi.
FREP AGING DBE C HOURS DF SCRIP 7 ION BLOB ELElf DETAIL ZONE ROOH ACCURACY COHPOSITE EPN
- 26RakZ I
~
f'OSI TION SMITCH OPC12 I 0 . A, I HS2841',. ti R 528 K2/8 ~ 4 t
=.
Ch CRD HCU"0647+
~
CRO-f'OS $ 2610 11 titI02 - I:;.BZEft 2)972,,' ~,f'I'"-3C.=., ,t~ -" P'- P48tt051 <<'.~'ttl .."6i2 =;"-"': "+I'-ttf '.I~t~.f3" ':-- "'='-'- "' IT IO(t )MlTCH P
'QgR+5/8 $ ff'I"'1 re', "t tt ~t 'ft'~ g~'~gt t ff -4 ft~f t gRII~ycg~idiit,
. f .ei7 POS 4 -I,",g
+> <-,
CRD POS 1261015 HS02 BZE6 2RN72 A 8 24 8003, N 612, ~ 17 POSITIOtt SMITCH 3 - A 1 0 '. A '.- ",,",' f H$ 28 f t R 1
',"' 528 LS/Bah Ch
' " ." "- ',t ..... >>
CROftHCtt-1019'2C12
, ~
0()":I, ':.= CRO~PQS~1'2)$ POSITlOHSlllTCH , =gP)P~.;-,I".Qg
.'-.."" -.".,-";.~ .w (0PPT(.
;-,',";s(
W '< *;~:!;,,$ 'z'"~, 52)
*,'.> A",B.,
lb]8 4;,"- '"248+
'~, .1~'P i'.-,6yh'+,'I-"+,".P~"<<..'.""@,',,>-".w;
'y '>< ".".,:;-'-.".'~ '",'.I~ttRDMHgg:Id234.
'-'.<'<./TIJOU -',.~...":
CRO-PQS p 02C12-1P610PT . H302 ZE6~2RN72 " " ' A 8 2480(l3 N 612 ~ ' i7
'I
~
- POSITIOtt SMITCH R 528 $ 5/8 ~ 4 CRO-HCU-1031+
4," OPC12 A 1 0 A H528 Ch CRD POS 1261035 POSITION SQITCII '. - Hgtt2:QZ$ 6~2RNT2,;-'t;
'.;", .;,~-., -,',-," -
'l.:.,
',k";-'28 )2(8 A 8 24800$ -'.'
=, Q f,gi2 -.'" '.
". -: '.W;~ -".-PRO~HCV-10354
'- ='
017/ n CR O-POS-1261039 H302 BZE6-2RN72 248003 111!'IL- A',' A 8 N 612 ~ 17 02C12 . - 3 A' l,0,. /if f .',; ~~ -H528. t POSITION SMITCH 02C1P R 528 KP/8 ' CRD HCU 1043+ A 1 0 A HS28 rh CRD-POS-1261047, -tI3ttg, BZE6i.g Rtl72 A 8 248003 tt 612 rd ITiOM SMITCH 528 K2/8.4 CRO HCP 1047+ CRD"PPS-1261051 H3ti2 BZE6-2Rtt72 P nh 0003 N 612 417 0 02C12 tl
' A 1 0 II528
-in-POSITION SIIITCII CRD HCU-1407' OPCI P 3 J' 0 Hf 5 2ll
~,
m UQL+C POy pl QS4g~ff.'E00$ PHEIIT L)SY-;- -; ~.,'.<<-. -;: .:..-=';. -, .'-~ - .. 6 E-t)0-00
,-DATE t)6/je(83.--:;=.-.
EPN HFG NOBEL S E OlD TH HL TEST ANL FO C FREQ AGING DBE C HOURS DESCRIPTION BL DG ELEV DETAIL ZONE ROON ACCURACY CO!IPOSITE EPN
~
P<SI TI Ott SMITCH ' 'RD-ttCU-Ihll+ 02C) 2 3 1 0 H528 I.R
~, '". LS/8)4:
528 C4 i " . t:hO;rOS-i2(3$ - - Bg i%2)f 72'Pg +3iQQ i+4 5 SyiICH 5iS'OSItled
$$ (g 4
;j'jjcL'5' -g iP>,
gjP,P . '<> 4~L>;4) 4-PP2
'P,<<is'(,"'" t~ Si:,R~'"; j4"-,<<i CRD'~flCU<< F4154 CRD-POS-1261419 H302 ZE6-2Rtt72 A 8 248003 k "
1 61)Z ~ 17, 02C12,
-3- t i.'f- .. A 5 4 ~"~" 'At . "kl'52$ ""'!'" '> -'h ~."<' <<
02C12 SMITCH 3 A 1 0
<<-<< 'L"
H52$ Rt 528 L5/8 Ch
' 'OS1110tt CRO HCU 1423k CRO POS"i2614$ 7 POSI Ti Og SMI T'Ctt
~-t)302
~-",'i I<<,'~ .'.>Ntt<< ~<<g<<'/
7,, ~
'.,<<4ggyg2+)2'c,'-~..l<<i~<>II .jI '
~
-y~ 4qh> ii <<<<, '-~<<g%.
'"!. .; P 248003'
)2B'MLS(R $4- '.> ~ .<<<<
"..-'k!:-,6)2," " .."! '
'i i i >)4<< '-<<s+ )"," ~'=<<7
~
.".'*~"..
P<<; g:~"
'"<<y"
'CRg+HCU~i4274 ilf ';".
CRD-POS-1261431 H302 BZE6 2RN72 A 8 248003 612 N - p F 17 i 02CIP, ', i3 POS1110tt SMITCH
<< ~ 3
' ,R '.528 K2(each ., CRD-HCU-14354 02C12 3 A 1 0 A ' tl528 Ch CRO-PnS-126i43P ',-H30g -:.'QZE6>> ttk72-' "-':L'>'"'4 ti.'248004.;->>. N<>=,
I-'28 '52(8 ~ 4 .',,;. t .,'." .'k ..~~"'~<<~'g'.'<<-""- .~ <<gi",-'." 'CRD,"HCU~1439> CRP-P~)S-1261445 02C12; ' a I;0-H302
'. A BZE6 PRH72 "c. -;.--";,>'<< . '~...~'528"-I- ', ctt a 8 248003 k 612 ~ 17 POSl f1 ON SMITCI) R 528 K2/each CRO-HCU-1447+
02C12 . 3 a 1 0 A H528 C4 CRO-t 45-)26145)" POSIT)ON Stt ITCH kt30$ QZE6-2RN72
'528 K2/0 ~ 4 A l} 248003 N 612,; ~
LRD w HCU~145it ei7 CRO-PI!S-1261455 H302 PZE6-2RN72 A 8 248003 612 p N ~ 17 i'2C12 3 a ) 0 a POSIT) Okt 02C12 SMITCII 3 <<1 0 a HBPR R i28 Lr/!I CRD~HCU )803+
~"
'4 I
.-,,-.-;;, '/<V'! =':.<=:"-r~'-'~i~~S
,a
$ .-. "~<",. 7:'-,.;.T~ 'N.~Z'4meh-'.1g. OU<eggNT Cgct->>, ."4; ~:-",~"-;0 "4,'-:i.-:.:-'~~-"~'" ~" 'ttYE'5$ tt58 A3>'-'- -'" "
EPN DESCRJPT10it, HFG' ~, HODEL .
.BLDG.ELEg g
DETAIL S E GID ZONE TIt HL TEST ANL FO ROOH ACCURACY C, FREG AGING DBE C COMPOSITE El'N HOURS POSI TION SMI TCII 52 Sic M2/8 10 .'A R 'CRO"4CU 18594 02C12 3 A CltP-IOS;$ 2tt2303 PDSI IN S(fNfl
-..;:~; 4 '~
IIQOq ',"<<~%ZE5PNAVR>"'~~-'- KjP@'-~~$ ~T".,'4 -4<< =A"Q*'.t"- fP ttO)XP,',~-"ttg~;5)j~- ~".<'-'>"-,"',,:~.",-;"'.",':-'41..'.",=;,:
~>+~~=:ftt')W>$5lB'i$:g.V<+ .'< F%',W~w~P~p""./;I;g'g" ~'+.';~t"--,$
., -.'-, t- -"',-'oi't,-
": -, 5 t='-"';~(.,Q j'"W~~~'-;"'.",,;.'~~2, -",!,'-
m.",IItD+8g)~2)03<'RO" POS-1262207 H302 ~ . BZE6 2RN72 ",
'- A 8 248003 N 612 k
~ 17
..0,2C12; POSI TlON Sttl TCH ' R. 520-L5/8-4 b2C12 '
~
CRD-HCU 221l< A 1 0 A H528 C4 CRO-. POS.'gu6galh.
;;I'OSITf05; I'tttlfCNP, gag",~ ig44gRfiy2,:- "-':-"'- -"4 )- '--".
".I;,.",.P.="~;*,,~~i;=.I '.+@" ~':.P4".P~. 820
.' ~'S" 2>agog --'="-i'--fjg'-"' -':- "-'-""';"-'='
.-~ '-.-'"." ".*.',"= ..='~~".'-.-.w.s-.';-,i;:.lPRO,
~ '-'~- '- - -: --,-tie '""-=-"
Y."- t)5/8 ~ 4-"'~'.-- " HCV. pter.0+-4'4 t )2154.'RO-POS-1262219 M302 BZE6-,PRN72 A 8 24000$ ' 612- ~ 17 tt2012;-i~=.,- t rh."', " <<w"~ ~X ~- '-f ~. ~g', jqQgP '~ r,"P.-.< 4tH', i y" dy~'".-.>>i",,"..y~'-"qt,,g" ">>. 02C12 SVITCH CRDTP08-106222Z.-','. A
-';- =..:
1 0 '- A gj)f;--'>~~' I!')Z'jgj2 jjg2~p'0';, ~i~'. H528 , R
'28
-;$ <<IT>> 't~" "4
'4LB/8e4
. '";. A
'S,. 2)800$
'g'qf'OSlllOtt
"-'~.- .0"642.'.--.-"
*' '-= "'""
. =
~.-..~t t~"-~'i-'". ="" -"-" 4 -~
CRD IICIl 22234 POSITtOIt=8tiIRDH..; I".~ f i ~. ~,j,. .<<,<'~';.'$'-= j.~4 > g r 528, L5 jg>4'~'= ~;- ~ -wi.- '.~14'> >i s~~ ~,';CRDgttt UG2227+ 0'RO-POS-1262231 M302 BZE6-2RN72, A 8 248003 tt 612 F 17 02CI R.', 0 POSIT IO'I SMITCH P. 528 K2/8 ~ 4 CRO-HCU 2235> 02C I? 4 1 0 A M 520 C4 CRP-POS~I)6223) M402 PZE6")RN72: I POST TIOtl SMITCH R '(8 $ 2/8.4 A 8 248003 CRO"HCU 2239> CRD-POS-1262243 M302 BZE6-2RN72 A D 248003 tt 612 ~ 17 02812 . 3"t A 1 0 SltlTCH 'OSITION R 528 K2/t' 4 CRD-HCU-224 7t 02C12 3 A I 0 b M524 { 4
<< ~ / .T
<>T i>l'IL /'Ii'i ';ISSPQ', L'!ISSL>>S!PSUIPIIESt LSSf/'Tjg/6.
. --'; ';~! '"". "L':,',<<"'.'", .'T>><<! ", Sll":06/SSTIS>><<, .<<,,'-, I o ii
;6 SD EPN '" ~, " ' 'FG '. '- ': 'HODEL S E OID TH HL TEST ANL Fo C FRED AGING DBE C HOURS DESCRIPTION
. LDG ELEf DETAIL ZONE ROON ACCURACY COHPOSITE EPN P
02C12 I POSITION SVITCH 3: A 0, A ~ .-, H528>>,f %
... R: 52B.K2/8 ~
I,j C4' 4, L. w. =;, ',, ~ ~ .; ', ~ - CRO HCU-2251n
$ 3f16>> == I'BZE)>2IINTQ;l>>>>'g l)6: -+ ~@ -"$..8.". 24)ODQ<,P'. <f'g"'612~:9"""."'- -',,'f+:.,'>>",:~s'"T "-'..'":" '.'17'.
CRO"POS 1262255', - svITcv:...: ": l,',;:".ll;."1-4: .',"":,P= .:;.'c'kg'(~~a:=,.528.i(giw;g'--j,,'- .-.,;- -A"'>I"'--:.I-",4-=;s:,:0;, 4;Pg~jgp-.(CU-gj54
-'osITIOII
/
CP.O-POS-1262259 H302 BZE6"2RN72 -:, ' A "8 248003. ' 612 p~ T 6 POSI 02Cie IION SVII'CH CAD-POS-IPbg60Y;" POSITIO8'II)TCII.'. ~ 6 . ~ Tl,: ~R528-LS/BP4
-<-'Igpg:;. j>,BZL'4+gtN72'6'5gf$.'. +Y j'.ig~gj":g"-> " 8 6'48QO) l*=~.."-.",II $ 2 '-'~A'PP.=.~ f"'4M;"-~;','".""
,.=:..":>>-'f-
"".', i.:
='4.)n >5'5'"'T~">~8'-="~C p,'5~2II'-45/BP ~ $4,@~I-'---'"- p4'K<<s-'-Zi'~ "-'I'"'>>> 4"'6 ~i" ".CRp'HGU-;-,:
/ T
.',TL. I'-'-.~i . ',I CRD HCU
.~l.
2603> 2607n, A'i7 CR n-POS-1262611 ~ H302
/
6" PZE6-2RN72 '"..',.',,, ";,
' 8'48003 ~ . '.
N '. 612 .", t
. ( . 17
'//
~
~
g., J. <<& 6+6 6
'4 l'f "6'.
' c<<l/<<,'8a~ae*;
02C 1 2 A 6 I+ 0 ~ '1. 6 66k>>6 f l<<4 g
~ -"*
fP ~ '-$>I " -'sk. ~'<<) Is>> SS T T 'I
---'T 0 POSITION SVITCH R, 528, L5/8 ~ 4 02C12 3 T
A I 0 '. A '.-,,- '528. CRO HCU-2615s CRO "POS 12(26)9", '>>l02 ","/; 8 ( .$ 8$ j2--.6,':,",.'> g$ ". f: '."-'.;- . 8'29800$ ,~ '.~"'I',,N'."';612 j'"e~~~"; '-~V: .-, ~,- I;.~i=<<"-- Pos I ri OII'IIITcii A ",-..', . 17
, ~
I << CP l>>-Pos-1262623 H)0$ BZE'6 2RN72 A 8 248003 II 612 ~ 17 PO 02C12... - 3 6 A 1-.0~'$ AP~~( ~ '"".-='., ';,;g~ Hg(8': ~,P: ~6 - sy Cn -;,:-.- T POSITION SVITCH R 528 L5/Ron CRO-HCU-2627+ 02C12 3 a 1 0 H528, C4 CRO-Pos~1262631 IOSIiIb>>I'SSfiIidII 1
."H30$
A PZE6 2RN72
' -"-=.. '"' .
6
/R.'528 L5/8.4 6
4
- 8. 248003 "
N 612 CRO .HCU"'2631's
-CiY CRO-Pos 1262635 H302 BZE6-2RNYO A 8 240003 N 612 POSITION S II ~ 17 l<<
I 02CI2 A I 0 A, H528 Cn 6 T a POSITION SVIITCH R 52>>i KP/fl CRD HCU 2639n 02C16 3 A I 0 A H52P rn o /." ' TQ
V 1
', ~-+t-..:-::;-",-g ":$ g,;;-~ ."'.-..>yNf~g f)PICji~ lgauI jSEIP LIST; -; ".,=;-.':;-0-"-,'=.:., ':;;"'~~'<~.'-..--'.-'-.",~", --..-.. UPTE-.05i28/BS ': =;"-
EPN HPG NOBEL OID S E TH HL TEST ANL FO C fREO AGING OBE C HOURS OF SCRIP T ION BLDG ELEV DETAIL ZONE BOON ACCURACY COHPOSITE EPN l C POSITION S'MITCH ~ ,III" 528,K2/8 ~ 4 >> CRD "HCU-2643> 02C12 3 A 1 0 ~ A
- CRD-POS-.j~gki)7:,.:;;,,A%a'..:
POSI Ti <ii .GM ffP9," I "">:-"..1:g'l "". $8 rbijSPC"" g'l,'.'<..
< ~-=:
~:: ~='".:,: l'.iT4P'i)OkP',.-..-,~.";-'$ .5$ P'.,:;-;.-; -='--.-,:,i-.;.,~~ .-'.-.,','=-,".~:. -..'--=
-"~~."'.5-"..4~8 ~PA+i5$ @XO/'8.'"' .l/ ~ -',""~~j'~~8'4:<<'~A'." I V..:.'-- /4 'CR$ i)Cy-26)7i CRQ-POS-1262651 H302,BZE6 QRN72, g <<~,",,- I A 8 4
248003 N, 612 4
~ 17 '0 C12'",-,;3... a:- >.~'-O,I,-,: g <,;-'p >,',,;>>'='.'..',~.g(2':.
POSITION SNITCH 02C12 CRO-POS t/262gq', 3 A I =, ', ~ 1 0" ~ A -,, H528 ', R 528 )<2/8 ~ $ C4
@<'g)Ogq:,"."',pZ)6A2Wtl't5 y'.'.,""4 ~) 7',-"":."<<',...:.:,'-B ')48003 -~>'~N =4lPw'g~"-~'-g"'+-'<'~'.:l - ':='- '-
CRD HCU-2655+ J 0
.:;'~,';." ~,~y'j,",1 <~,,';;; .j.-.~ <i+$ <-g)j>j2)B~ $ -" . ';"-. '~~i'"".-q>',-"'$%~ <<~> -,';>, <,'<~cw~~~'~>
< j'4;:C)5+)CU"2659>
CRD POS 1263003 HS02 "$ 246"2RN72 A 8 248003 N 612 "
~ 17 POSITION SMITCH 02C12 3 A I 0
=,,":.=-
- A..
=
H528," R ~ 5?8. j.5(8e4 C4" ~ 1 CRD".HCU 3007< tHO-POS. 1/630)i,,'-~.~ OSiyIbik 5MIT'tP'--
',- .'jH$02 iPp:$ 2$ 6$48)ff".".g~.~~:;;-&Fr- i '," Q,II ',2l1800+';;'
r-'.~ SiP~'r/A 7',,'.O'Vg)8':4yj8'lj.'.-.i :-'>- .<'- 'Q';",4)4
~:. @~', r; g t '.<;-r~-","z'<-,.~< ~
.'-'-.-'~,<z,,"" ". P,"=".>>.; ')D~H)V~30114 0 CRO "POS-1263015 H302 DZE6-2RN72, A 8, 248003 612 4
~ 17 02CI2 '. -) - A ..i 0;:.: P," =.
-:-=-'-.'; H028' " " C4
~CU 3044 POSITION SMITCH R 528 L5/RE 4 CRD-HCU-3019+
02CI? A I 0 A P 528 CR D-POS-1a63023 8%02 . BZE6-hRN72 A, 8 248003 N. 612 POSITION SMITCH R '28 L5/R ~ 4 CRD HCli"3023> CRD-POS 02C12 . 1263027 POSITION SMITCH A 1 83C2 0 A OZE6-2RN72 k 524 R
'4 528 K ~ P/3 A
~ 7 8 240003 N 612 CRO HCU 3031+
~ 17 02CI 0 A 1 0 8520 r4
>>A~>>>>
6'
' RAG t-',>-~'. >',.;-.','pe;,.k.'CLASStgE EGU'IPHENT Liet.. -,,, DATE 96/28/8>>3<<,
~
f f 4 EPH HFG HOOEL S E GID TH Hl, TEST ANL FO C FRED AGING DOE C HOURS OFSCRIPTION BLDG ELEP DETAIL ZPHE'ODH ACCURACY COHPOSITE EPN [~M POSITION $ MITCll CRD-POS $263b39 SM)TgH'*-.- A 1
, ~.;,'-,:
~Q 0, ', . H$ 28"'-.
f R~ 528 .K42/347 a f C4 ".,:,': ~
'Z':.~ 18ZE6 .'2R472'.-",,',-'.'.'-'"> <<>'," .-"".: l>> 4, 8> 24,80O> f; .',<<;( 5'))aX,=.'..=:-",=, "',. ';;,.-'.,'.-s; CRD-HCU-3035'2C12
~, -.ig', ='-- .
'>> f 4
00$ IT$ 0$ ~>> '~z ~: =" '><<>> .>,'>>t "--q6'C9giC'Wg(@i2!~>>$ '~~j<< '>>>>..,i>>~i'~'g,;1'ff~g~,",'1
>g '~y,-5 ..f'<<i;-~'gyes "kRP~ljCU~)0594 0
' ~ f CRD-POS" 1263043 H302 . BZE6-2RN72 248003 612'-
P A 8 N ai7 r 02C12 4 POSITION SMI TCll 3 A 1 O'-' -"s . -= = q<<t28
.R, 528, K~
2/3 C4 7, .'
'. ~
~
~
sf'<<';.', (RD~ HCU 30474 CRD~POS~Q'630)1 )>> i ~ . <<-H)og" <<.:;5@6>)tlN72~,~ "" y~ 'y'tj'V "-'"."(>>'0 '>'B, '48403.'; i~-$ <<P$ gQ'>>."*'"4 '-'=-'<<g'P'-'"fi>>>>s>>g'"" -<<' i'.'-*; ':';.',@~.'<>>h- (RPf>>H)V~30514
~
POS1TIOhf <PT4g.', "
.P. I;.~<'- ".,'.- >Q-.~<~~..;,rt>>'>>~Q'~'~~/(+';:cleft'>>Il>>glt3 7,'..'~ ., ~"i..'.-:.>>i>'~.
CRO POS 1263055 ~ H302>> PZE6"2RN72 ." I
-" ' ls A 8 248003 N -'612 ~ 17 POS T 02CI2>>4 -.:.
C
'p<.>>. A
- s)~0';, =,.<<'4 <<g>>s<<~j'-"z>>ff:;-H$ 28.9.',.;.'. ft; C4 ',>4 ~
f.
'RO P 0 POSITION, /MITCH 02C1P CRO-P0$ " f263403.--.",
PGS IT/ON SMITCtl : CRC-Pns-12634O7 3 A I 0 ,- A
',H)(f,"'.";jIZg642Rp72
=-
H528 P
~
'f 528 K>>2/3
','.;".='.-',;". 7:'-.. tf
! '"; ;""..";>,',>>g ".'"-".'.;-',."-;1 '.'.P 528-f H302 DZE'6 2RN72
~
'513a7 C4 A D '248004'=
<< '<< '"',"", "i:
."... 0;y612",'
"n,"
'<.;..'.~.", ";i>> '. =,
CRO "g'..'-'-'ij -'. "".,";.CRD"HCU HCU 3059+ 34934 248003 P 02C12, I tl.VT
=,
A 1.0:,.t; A ':;-':.'...,", =M.H)g8 <<>> ~~ <>7 CRD"HCU"3415+
>> 0 4 CRD POS-1263419 H302 OZE6-2RN72 A 0 248003 N 612 ~ 17 POSI l IOtt SMI CH /5 02C12 3- A 1 0 A
-PO - 34 POSITION SMITCtt R 928 L ~ r>> /3 ~ 7 CRD HCU 34234 02012 3 4 I 0 A HrPO ('4
4 HFG HODEL S E 0lo TH HI. TESl'NL FO C FRED AGING OBE 0 ., HOURS DEscal p TIDN BLDG ELEP DETAIL ZONE ROON ACCURACT COHPOSITE EPN POSITION SMITCH 02C12 --3 A 1 0 A H52h ".
- , R;,.528L~
~.
~ 5/3q7 C4 CRO-HCU 3427+
EI t ltPE>>i OS-.$ 96383i..,; -.-
<- g362)."-".'$(gg<2tIg72."g'>.",$ g,'";; '.~I~~",-,'"",,"( P>>.:,2,8ett3+@! Qg+Mih';"~< '.,
CRD-pns-i263435 H302 , BZE6"2RN72 A 8 208003 N 612 i17 0 4 POSITION SMITCH 02Cl 2 A 1 0; A:- ' H528 .*',
~ R '28 k+2/3@7,,
C4
'RP HCU 3 439<
,4
'- i7 '.
" = ~
',-" f.;bN4~'jRN72~: ~; "/: I'l';-~"--:;-':"'-':l:-. ~.,~'.-24003.-.,-<<',.'5>6<y'r.-".,
'p 0'l)tlat i<<g4>>, '>>~ &
CAD-t is-.ggI34~5;"-- :. ',-".'-'..--'-~;-,'=':-;"-- ";. - .'-. '-, =- pogf TI pg<< $ QI Tgtt >> jI '44$ . I 'I
>>I gfjf, g 4" g. EP,>>>>JL>>*
~Q >>$ $ 1I $ gg)$ f " %% 7, 4 C4 P>>' IE fop/AU 34430 4- 'I >>
v CRD-POS-1263'l47 H302 ~ BZE64>>2RN72 A 8 248003
' 612 ~ 17 q2c 12, "j
", y g Rt p- R 0 f 0 j'lf>$
~
44-, I gg>>+ g>>>>g E<<l >> tt82ttP/> -<<F>> $a .
<>4>> IR 4 RI j ~a>Qw," AgVp" 4 4 '4 W ECWR s-, 4=,.$ P " 0<<QN a p 4,.'OSITION SMITCH R 528 K>>2/3 ~ 7 CRO HCU-3451+
02C12 A I 0,. h, H 528, I C4 Ctfp-POS,-1263455,,-;-'-':.'.=:Hg(j,"- -;. Z$ 6~5%875"'I';-"; ".-. ~:,"..A ': ,"~~" -':, A.- 4 'tt80)8, "-
,"', lfi 'Bi2-:.>>;>>-"- .'-;,~ .-.",~-.;.;;=;.. ".
i- " . ",'?17--: '.=-
'ab POSI ltoii,.'SOITPH, . - ".-'..'-= '.-.;-".;-~::~4'~.~.~:l~ </".!;,.:=-;,t;jP" 528:,$ 0 j3 7 .',.';, .' --'~,.;.~> ~ ."-'."-"
4.""'l~',~j.P';:.lERO4ggII~3hssi can-pos-l 263459 H302 BZE6-2RN72 A 8 248003 N 612 ~ 17 a 4 4 *,, POSIT100 SVITCtl R 528 L o5/3 ~ 7 CRD HCU"3803'2C12 3 A 1 0 A C4 Ceo-t OS-I)63e0 j ', POSI Tl Otl SQItCHE i.-H)02 bZE6 2RN72 4520 L.5/3.7 A 0 248003 1j 612 - -:=, HCU"3807<
>17 I C CRII pns-I263RI I H302 BiE6-2RN72 A 8 248003 N 612 ~ 17 02CIP 3,' 1 0 A pnstTIott sullcti a 52tI LED/3 ' CRO-HCU-3815' 02C12 A I 0 H528
~ <<lA*h i
Il 1 r A I -P, 44 -:, ~>l .">4',~.',I/-'~,-.4MttP'-.j ICL'kqSVJE.'EO0IPHEtty,.'Limni,~"-: "-'-'-:-.~.'-~-" '-.'.1,:='y<.".'-'<<" ',"', . DATE n6/28/83;=."; Il A EPN ,HFG =', ".,', NOBEL
. 'A S E ~, 010 TH HL JEST ANL FO C 'FRED AGING OBE C HOURS PCSCRIPTION r BLDG ELF' OCTA)L ZONE '00$ ACCURACY COtlPOSITE EPN
~<~ .L,f IA4>.,< 'PII . g).'I ..i> .'>; Il- q,P- q I'A'PjS POSl TIOtl SMITCH 02C12 A 1 0 .', < .:; -.'
H52B.=S
",. R'28'L
.*'-'*l.l
~ 5/3+7
- 'C4
<<'RD~HCU-3819>
CRO-Pqy-126382'3.1 POSI TIO) -,SMI)Cff':
"'~
HM'2'..;.'q',BZEBARRNi2't'.4f,'.qi",~i 4,:.-.~<<~ g;=,:,=~". -.4"y,,'<<Z(80btt k-'<<~<<.84$ igti-"~'<< -~'..w;-P<<'-.3;". ~.,7 8 g.-ii 4 /+~1=-"."%<<0b- -;g~g4;kgtt. L-5 31 j:~ -.'-".a/~+~'"t <P~= - 2 f@i<<= ~.':~t +.,I /P',O'RP'>>jCU-3a23i-,
=
~ I I =.Si f ",.;-." =-;;
".",": ',ii CRO POS 1263827 h
; H302
~ I BZE6"2RN72 . "."
I
'.;, = ', ' '8 248003 '~N " 612~ '
"I F 17.
POSI TION SMITCH 02C12 ', 3
~ .IRf 528.) ~I /3+7',' ., =", .' << -.', .': "-CRD;,HCU 3831+
<< ll CRO-PgS-: $ 26383II".',
POSITIOtk SVfTQQ-,,
.'; 'j'."5"-Hgg','g<<IIZEP2g)P '.<<%g+','g,, F4/
- "4" -$ ~4 ')>'~~1, ~P$-")4R 4)RPP/3<7"-";'.I~,~~P.'.,I P
l';:"-<<-.,',-"'~z " f. tl'-:.< 248003: Pjg-'":6P'"<~ ""-.~~<'~r'g>"./gX"--":-'-' '=-':,
$ .+g4'<< "4"'.- ~<<4jk'<<.-Az- I~.QD~QCU~3$ 35$ '7." v', -":,-,:.
CRO-POS 1263839 '302 - BZE6 2RN72
. ' , . A 8 248003' '
"6i2 ' sl7 POSI 'llOtl SMITC)t 528 '/3 ' ';I 02C12: 3
'1 0 A, I '-; H528 R
.,~ K C4 h
CRD;HCU-38431 CRD~POS,".12tt384$ POSI1fOtl" SllITg)t; l.i A .;,;I'.."..=:g:f
~Pt tt302*>; ',;BZE6~gRN)2IA-~-.:=.+~:,">;QAP .;j~
-'~q@." a..'P-.~:.-I~ a 5)if"
';". A.;S.::,24800$ ,/,FP.,',;g',~61$ .;,,1;","IP'.-.~~ 4. t~P"y<<-.'I'" "..1<<;.-, ", - .* =,
0/tli7::1. '- ='"""i "<<14 '". '@ "~-'<<'.'I *
~;,;-;".j', . 8 J( - ~
CRD '@U<3847i' " 02C 2 CRO PAS-1263851 tI302 BZE6 2RN72 A 8 248003 fl 612 ~ 17 P S T OIt 8 POSITION SWITCII 02 Cl 2 3 II I R '28 K '/3 ' CRD HCU 3855+ 0 A H 528 C4 CRO POS 12t13859 " -:;;H302 -" BZE64)RN72 .'* ""-,'.,;-. ". - A jh- 248003 g1 7 TOStlION SMITCq CRD HCU"3859> l ~ I 2 Jh l 'I " CRD-POS 1264203 H302 BZC6>>2RN72 A 0 248003 tl 612 ~ J POS 7 Otl S I 0 lt P: I 02012, ." ~ 4 I 0 H528 A h I 0 ~ (4 P P0 9 I I I OH S'MI TC H CRO HCU 4207+ 02CIP 3 I I 0 A tl.120
~ ~
r r ~ . 7 '
,',.;.4;,'--.'-j; Std'~tZ'.ppqf~l( t'tIVgy HEN.,t,lip t';l,~",q.4<< 'I =,~",'.," *'.=-~;.,'=.'- '> ".<, ",=',"., -"'. - DATE ttSj( gq3;-...:.'--:
'a
) EPN HFG 'OOEL S E OlD 'H HL 7EST AML FO C, FRED AGING DBE C HOURS I rESCRIP TION SLOG ELEY DETAIL ZONE ROOH ACCURACY - COHPOSITE EPN POSITIOtt SVITCH 0, R. 528>Li5/3 ~ 7
'4 ',,'",, ~~, .' ~,CRD HCU-42ll+
02C12 ', l'IID-P68'tthttfg-; l'DSITtON. SMf f. tl 3
'.: A 1 A
"-:: ~; (-" -.')$%~:,'tt~'g'4r"-."4"'"-'
HEt28<<; ', ."$ <<
',;g ty3<<02,.-".-=je 4( jjIIN'723<.,-,";.4", ".- 'k~~<'-.%,'-'-';X',g;.<<g480j@+")I't jy$j~"- 'V'~ <<t'-Mi>"-. "';"-:g-":=-".~
'.).."~) W'kI<<lfpak)3<<I)~L/.',"."'- -"( .""'-~ '<<:",". r~w F42~'.'>+<"~:" .'-'Zap>@p 42)54'6
'l. g r CRO-POS"12642i'4 H302 'ZE6"2RN72 A B 248003 N;612; r ~ 17 r
<< 0 POSITION SVITCH 528 l.e5/3
- CRO"HCU 42231 0;.A 'H528" R, ~7 02C12 A I .. ~
C4 CRO-q08.-'ic4qe2'-..,. -: '; ~:,4.W4!,N~;==';.:.."j;.rt,"-'.".'-",;"..'.,'.".-<<, l:I r 'g';.=:,;4fgj~'jATe>>L't.;-~I:'-95<<
.:.,-.-.~ '.- -".
"w."M p'>+~-. ~ ~ 4$ $ $ 2$ +'5/3 A=.q,".,2480cty~ '-<<j17'-,'",=--
yjpff $ gfp Q '., r + <<g,",'-,."~Q.
-.-'..=
7<<.-", r,-: =='"<<;"<<<<<<r~'"r" '.-.'-',I<~~t g,","<<; ~P~'~',~jy;,CRD~(glJ;)2274-. CRO-POS-1264231 H302 BZ)6 2RN72 A B 248003 tt 612 ~ 17 ir f rw ra P OS I T I 0tt 02CI 2 CRD S POS'igg(23tt:- tt IT C H 3 A I* 0 A t528, R I 528
-~'<<H)02.".,g.'5Pjg'.gRH72,';;)i",';4 ">>'" "- -'," l-B.- <<248!03 K ~ 2/3+7 C4
~
~,.
- ..: " CRO HCU 42354 P-c't30$
PASITIOtt SVIyctt . /r + !..l<,"'l "<<' N Q~r,~;~ '.~~.'-?,"R'SZBr~ ~ 21<<3i7 CRO-Pt<<S-1264243 BZE6-2RN72 8 248003 612 P A F 17 02C.I2'::=r '-. 3-.. A 130~. -A.,". ~: C4 POSITION SVITctl 02cl? 3 4 I 0 A H 52<<t R 528 K '/304 ' CRD-HCU "4 24 7+ CRD-Pos-'1)64251,i t IH302 gZE6 4RN72 A 9 24 8003 612 posl IIDN, svlictt 528 K ~ 2/3 ~ 7 CRD HCU-425lt CRD-POS 1264255 tt302 87EI<< 2RN72 A 8 248003 612 ~ 17 02C12 3" A I 0 A M528 ', C4 POSITIOlt SVITCH 02ClP 3 h I 0 A R 528 K '/3 ' C4 CRO HCU 4259+
0 1 1 Eptt CE.SCRIPTIOtl tlPG ', '-~$ Pzj:,'.,"~'",-~.,~'.:.jOtt)jg. HOOEL BLDG ELEV L'k,
=,
gyiE Ed V $ DETAIL E tfht]f, Lilac..>'"-".'i,'.";,i',".".y".-"-~>, ~';-., 0IO ZONE TH HL TEST AttL FO RdOH.. ACCURACY C FRED P -;::..<<,';;;,j '.: AGIttG DATE DBE Og/g8/$ 3-',;.;"-"i C . HOURS COHPOSI TE EPt) '. *
~
POSI TIDE SMITCH 02C12 4 1 0 A R '28'Lj5/3+7 '
~, ' -
I
'. CRO'"HCU 50114 (RO-Pttg-Pt'Qtt)'%"-
PD>'ITgdg gy~yCH '-,- '
+ 2" "'TBZE6%+Q72,'"""~f".-'
'<'g:,""-. "..'.-$ -'"
~ -: '-'j-.', -"';
2> '"-~'~" '.9
-',"9@Page"
'P" 80ttp'".45 ~~t ifd'"- R'-/""-'~y~
CRO-POS-1265019 H302 BZE6-2Rtt72 ' 8 248003 =612 ~ 17 v + 2 .= l
" - 3 4 ~
Q vp} " '$= p' ~(<~ > 'p 'gy'~tt5$ i) "i g i"14 (s) ~
$ ( p.>yg ~- '11 Q~g4p w a'g;.;g4 p- p gP/g ~:~s "'~~ ~;~ ...-' "; i " s a POSI TIOtt SMITCH 02C12 3 A 1 0, A
~, H528 .=
R-;. 528 L+5/3 ~ 7 C4 ORO HCU 5023i CRD POS POSITEOH.
$ 26'$62'2:
SuriCH
't;",H)Pg
<,, +PZElf~jggtl)2'".,~q8)o:.g Q4i~'i ~'+ ..f(" g+":. A 82480') ' ~ "'ij ., fg>
'-.'",.;g:;:.';;j'~:=--,.-,;;-';:.,~:,, t -<:,;-jk fa8.,%..b 3i)i'-":;:,,~ '----, -.',"."~".,-q j "-'-f> ~" ~
y+ 'r '
~/ '
)7
.-"',-='~'i"-";j-:";-;,Ciii-ACtj~5tt274'RD-POS 1265031 H302 BZE6-gRtt72 .'. -
= A P 248003, 6ig ~ 17.
POSI TIOU SltITCtt R 528 K ~ 2/3 ~ 7 CRO"HCU"5035+ ~ 02CIP 3 0 A 1 HS28 A =
. C4 It CRP" POQ? I265P3):-,
POSI TIbtt SQITCtj
'Pt'@'.;If"I"~gli47)-'~ '".- I"- '-'~'- -'-, '-"
g.-<<,',.",,."- -,.'.,4 ',.'5/8'Il;2/3 f""'
- 4 0:-.24)004 " -'>'-612".'l'-~,'<.* ~-
-,;=- "'.."-;";-', ';;- "-~:,.'.;" '.-,'- ":" ='.,CRO~!IClfA5039 CRD-PnS 1265043 H302 BZE6-2RH72 248003 A 8 612 ~ 17 02CIO I~0:", . A -: "~. -~.. ~ tl528 ",'. -. '=- .. C4 L
Ak POSIT IO"I SltITCH 5?8 '/3 ~7 a'RD-HCU-5047+
'3D2 R K 02C12 3 4 I 0 A H 528 C4 CRO-PnS-156505$ RZE6 QRN72 4 0 248003 612 17 POST Tldtt SMITCH R 528 g ~ 2/3 7 CRD~HCU "5051' CRO-POS-1265415 tl342 D/f 6 2RN72 4 8 248003 612 ~ 17 02C12 3 A 1 O' H528 POSI TI OH Slt ITCH R 528 L '. ~ /3 ' CRO-HCU-5419+
02C12 3 4 ) 0 4 H528 C4
}
l ">>>t,' ~ << i = Alt
~ ~ t I'
-LEAPT-." <<
',t=.,g'T;"'>sy, 4;-'"-8"~.'r <"':4""fffttgwg'-CLAS@lf 'EOU<<IPHEtIT t>v"P) j:-"".',"~p .", <<"'xg-,'.<<'~= '. ~'-"', i", ', DATE<<06/28/43-E Ptl OESC RIP ZION HFG .' ".'. 'IOOEL " '"';;
BLDG,ELE'f DETAIL E AID , ZONE'OOH TH HL TEST ANL FO ACCURAC)
~
C, PREO AGItIG DBE C COHPOSITE EPN flOURS ~ O
~
t' 4 pnslTIDN svITcH " R. 528 Le5I3e7 'RD HCU~54234 3
'2C12
~.
t 4
~
cRO-Pos>>12654 7 + 342'7tztf} 2 NP '->"1 "' ~-'- + If'>409"- ~ "<<'>> ~ 2'-
':+
Posl Tfolj qvI)cp ; ='-'.
";P; z}",-",;j~~',:-* ..0(ltd'%28,'..5/>>l='~.=';-"-j@4".-'.-fj~}.~'~':~"-~s~
1 P~>>. -.-.=.'" j "."-'-'"-.-'~RH ",-"54%;.'=-"-':-.<";- t t - t ', CRD-FOS-1265431 H302 BZE6 2RN72;i .'. ': "s', *! A 8 248008. <<612",' '" "" '- . '. - ~ l7 t 02clg ',".-.'- ="3 s-I r <<<< i , ~ pnsITIGN svITcH 02CI2 - t3 A 1}41 .
';j r<<'-' i4'528 A:$ ~PI",
s 4 ~ ~
' == <<'.',R"'28 K+2/3+7 ft't,; '. C4
~ *,,'. - .' ..~.. ',. -..;-;
0 Q
. "'. -;CRP HCU
% - t W <<
POS IT[ON" $ VfTClf1<< '
.'. s,~'H30I}
'!i'" ~S
",~~+!Z$ 6<21~7$ <Ir>. v"I~C)<<T
~j>'4 >f,l 't'<<r'.'"g'<<~- >"~
gj',;b A '8."i 2480 $ "-(+~~}+<<.6jg'Z >"> - "='>'"-Ir'+~'<<- s " 5435'RD-PQS-1265439"i.s 5 ) %III'$28: <<2)3p7 . '"- I<< "..<<".~'.-.t <<>t' w -r<g< >c'-'"'Ii<<t -<<i~< j "'. pb+ftCU~5439> -. I I CRO "POS-1265443 t}302 5ZE6"2RN72 4 ' " i .. 248003- .' -'6i2 ' I A 8 i17 POS 02C12 I T I 0 tt CRO-PO(-1265819, POLI T>ot] SVITCit S VIT C H A 1
'S:-
0 Ngg)
'<i~<<<;
A
<<<<<<pE6+)tf)72'-y'. I;.z If)28',
~
R; 528
-"y<<i 3
~', ~i"."':';r,'...,."-~".j 528'tj <<5/47
'4 K
.'.I'."q'
'/3+7
. 4.8'./48003tt
'-'<<<<I'<<<<,;6$ Q'r".,:r.<<i<<< "<<'<<".s<<<<tS>"-
f
; '; ,.-, CRD i'-.
HCU-54471
';"'. '=, - <<i7" Ti@- '4I -
r ',.-'.~",-j}:-~s> i !I)-;<<}ti;-." ~".=&i."s,'..i,"":p<<;CI1D+)tCU 5II19< '-, .;;- s-;.. -'-<<~7 .- ~ . E J CRO-P OS -1265823 H302 BZE6 2RN72 4 8 248043 612 T 02C12 ' '
'I .A 'gg "t.'r A b"' <<. '<<6<<y<<< 4" f<<" 828j ti;
$ s','. 4; Ch -,.I +" r =
". b" <<'- 'Q6 "'
b "~I" v'4',S<< " '" j' ': '"';
" ' <<' +*
POSITION SVITCH * - R 528 LE 5/3 ' 3
<<,, I CRD-HCU"5827'2C12 A 0 A N528 C4
's .... = ';..'.; "-:. ';-
CRD-POS-It26503X PnsITfott sv7) r
.H,;, ~ >>",ttpttP ~,.
"'.. BZ)$ .OR/72
< .~ ~ .R 4528 A 8 248003;..
tt;2/3.7;<<:-,.'"
$ 12
~
'..-;t >...'. :CRD-Hcu-:583ii
!. ~ - .'t'),7',.'-,
cPD-Ins-1265835 nc r o2C12 PS ~
"" 3 1
8302
- 0. "A ",
BzE6-2RN72 N528 A I} 248003 612 ~ 17 POSITION SIII TCH R 52t<<K ~ P/3 ~ 7 CRO HCU 58391 02ClP A 1 0 A H528 ra
~ I t ~
t EagtettEg f L)S '.'"; .'.- ",-...<'~'-'-..->"'-3.'."<<;:~ 'l"-. '-".~-'-.;;"-), F - 0~~0, DlfE'06/28lel3<< ?i
<<I -,-...','f
- 2'I Eptt
~ ~ *t HFG HODEI, s E 0 ID TH HL TEST AttL FO C FRED AGIHG DBE C HOURS DESCRIP TI OH BLDG EL/V DETAIL ZottE ROOH ~ kCCURACY CotlPOSITE EPH . ~
I
<'l,,
I POST TI Ott SMITCH '~',I R, 528cKg2/3o? * ~t ~ HCU-5843< 02t:12 A 1 0 A;
,, . H528-h C4, t
CRD CRO. F$ 5~12?ltgg,.'t";. 2R)$ 2,4~-'~'~,~'~'. tel,",'~.-~.'A 0:P)8<<ltd .<S~@ g,P)g";-+'P 'g":S',.'<-'v~'='<<>.t,',
';)'g<t,:$2 -.,4';,'tt@$
=;3,',. (.H i'oak ko(.NQI)bH.,;, j '.,'$ ".."j~q,',.v.-':w~',i>%'-'PIC)k+'Lg/a~4:~<t":,"-'.; '.~ps'~4@'>>".-'. t'/2g .'~"=,":>."'~:egg:-gcu.-(tf9<l ',.'";;.';--~;.;;;,,: 0 can"Pos-i2?0223 H302 ~ BZE6-.2Rtt72 A 8 248003 tt 622 ~ 17
.. j;.".,~3:> .,
tjt2G 2 2 I
- ~
POSI Tlott SMITctt ; 528'L5/8 ~ 4 0PC22 . 0" i" ~' R CRD HCU-0227> 3 A 1 A H528 '~ C4 CRD-:. POS'-g27P23i':, ':;;-.'- .-880 ./t-:a@8-AyNy0;.'-: ';.4- < 'I~ "-'"';-'t-" t 4'8"." 248005k/'"'g "gk'- "-" - -'y17,t ';'~ f "; '-: f-.'.
"'"-';"'"~"'"='-',-'-~"'~'-'~,'-"'~"'o<
1 7IOIt- skitglf I
=,'- ~;="-,',"-"..;;.; -.g y.'*=",x.'-;.,-'~;.'-.5:,~f >","@.548~t 5'/e.4,,- -. ." -;.;: -".'~~. ':-- ~~.'-,",:-'-.";-"=.-."-.cjfiAH)u'dyl+
t CRD"POS-1270235 k H302 BZE6 URN?2 =.',;' a A 8 <<248003',<<
' H 612 af?
POSITlott SMITCH 02C12 3
' I ~0 ' I, "' tt528 R g 528 K2/8 ~ 4 '. - ' ' -.; '. CRD HCU",0239+
I CRD-PQS~t270243;.'--- SQITQH,. ';;
.. ~ -
".ft)0$ '-'"fjZg6>kflgfgq'~- '~.7,4'/'< ",<<='..,'2 ',. A,B" 2480j))~= "~.'. g-'6j2+".k
~".q-( '=g~~jl <,r~t,;j -"<'g~.'-".'t -tg-'t. $ 2(-"gkt'eq4 <<> . '~.."-: ',;~.'. ~".;~";;.~'. ".~'~',~",~.<.( "'.,'OSITIDH.
-.'j'$~ CAD>HCU~0243+-
- .ky .. ~ -, .--
'a
~ ~
0-pos-1270615 H30( BZE6 2Rtt?2 A 8 48003 tt 612
~. ~ 17
.,02C22: I C5 <<l<< ~
t ~ tt .' POSITI ott SMITCH 528 L5/RE 4 CRD-HCU 0629+ 02C] 7 3 A 1 0 A H528 C4
.Cao-Pttq-2270625 H302 PzE6 2att?2 I* A 8 248oo3 628 SMITCH " H 427 I
'OSITIOtt R 528 t.5/tto4 CRD <<HCU 06234 CRO-PPS-2270627 H302 BZE6>>2Rtt?2 8 248003 tt A 612 ~ 17 ocl2 $ A 1 0 4 tt520 POSITIOtt SMlTCH 42t'B/PE 02C22 R 4 CRO-ttcu-0632t 1 O A HBPO f4 pi
1 J'J 1 O. j~( t d]JJ t )II J'gkPJ$ +P(JP))E)7LI>7 / <>ggl ~ t t t w@DKTE06I$ 8$ t) EPN HFG tlODEL, =
. S E QID TH HL TEST ANL FO C, FREQ AGING OBE C HOURS PESCRIPTION BLDG ELEv DETAIL zoNE 800$ AccURAcT . cotlPosITE EPN
")J,, *HP ~t t, " ~g JIB"
~ /, t~ ~ P, Q J, ... + '"' "
- POSITION SVITCH -'- -R =-528.K2/8.4 CRD"tlCU 1047+
02C12 b 1 0 A .'Ht)28> 7 ~; Ip c4'. '- CRO-POS-gevli07 H302 - -'BZE6 2RN72 A 8 eheooh, N 612. ~ 17 02012 '>>>>7-',.-, 3 '-'":-A;-;ot j )$7 y: <<I xP~~+tt'Jgt<r'Qf)2$ +~ '-.j;wgt. g )4>>,',gj i ""='""...'.<"i 4~$4+p~~t~e.'";." POSITION" SVITCH 02c12 -3 b I, 0 . A'; '- t '; )1528."
,. R>> 528'L5/8 ~ 4,-,
-, '. c4=
.-',, CRO HCU 14llt BOSKET.ION.SPT,T4q ..
CRD POS 1271419 tt 302 J ~ BLE6"2RN72 .-, '.." ';-; 7, A 8 248003 ", )f .'612 .-, ~ 17 POSITION SVITCH 02C12 I 0"., A-,:,.
-"'> ,'>>>t)P$
.'.ttsee : '."528 =
L5/8~4,, c4
~ ~: . S .,p CRD~HCU~i423>
cRD~Pos. $ $ $ $ 42 ...; ":<@$2$ 6yettegg~q '.tt:$ ;~>'--".,'~: ',>P ..'I 44$ j>>>2 pe& i'PJ" )I'. 6iett V".~,~)-; '-".'l"',"'.."~P.,"-'; =;.".-"t'<- ","
'.-,,,'elf ',.:w
~t
~-P"'
PositIDM sfllTcH - ';-""..-.j~".'-t-'/.'tb-.."/t.-".';-=-.<-'.~",:",'tt';7ggy;$ 5/eely'..i"- '->>, <;".= =-j."<<"~.; '"..-'-"'t. -'.~-~~'"-'." ',-,; cRD, ttcU. $ 427+ 0 CRO-POS-1271431 ' H302 87E6~2RN72 8 248003 N 612 ilT 02C,I2 g >'" Ttpe >'*')t':,~ < <<'+> ';< ~ t)5aei ~ > b..-.-"t "C4 t Jt
't POSI TIO>>t SVITC)l 528 K2/8 ~ 4 R CRO")ICU-1435+ J 02Cle 3 A 1 0 A H 528 . C4 .>>t CPU-P0$ 412714H .H30 . = 'zE6-'2Rtt 72 A 8 eieoo3 ,4 "o'17
'J
>"ri
. ) . )>>OS)7[ON* SMITgt) 528 tte/8 ~ 4 >>t J R . = t:RD?HhU~1439i J tt t
")J, CRD-POS-1271443 H302 87E6 2RN72 A 8 P4 8003 612 N 17 qecfe ' A 1 0' 528 r4
~
POS)TIOH SVITCH R 52f K2/t< ~ 4 CRO-HCU-1447+ 02C12 I 0 a tt 52>>>> r4
<<u< u
<<l
, u uu R
~ '" '-"- ll = " 4 -q<'j g" 'fu'w i'u;" ~f<j,fj$tf$'$$
f LiAgs .$ g QQUIPHENT ~ Lfgb 'I '"-3""'- ">-'>~ '- ~</". "- '""i"j . EJllLJlgUt
'ATE>> )6/28/83'
>> a,
,'Q'3i
~
p, ~u ~ C tf AC I EPN ~ ul'ESCRIPTION
'!lPG A
.,'...' ", ' HODEI.- .'.:
SLO6 ELEV
-, - 3 E DETAIL GID
'ONE-
~
Ttl HL TEST =A(i, Fo g'e.,; ROOH = . ACCURACY FREQ
'OHPOSITE AGING OBE C --;
EPN HOUR$ I uu 0 u POSITION SMIICH 02C12 CPQ-PQS-1271455 POSITION SMlTC}f* 0 C u3 A 1 0 A, ',
)fly ',-'82E642tftf72',-"',,'."I >>@P.'I
.",.",0',.~:. j.'u H52$
=,
R,52(
"~ . ';..
'. g'.-
-';;P',!gi,'+.,:Q8"'QKk/)y$ 'l,~~.,
K2/8 4
~ '.
C4
~
8 ",24800),",.'-',>i>' 6fg!'-;P~~~':.'"->,'.>P ...".
".... j4'p.4','-p~ g(";.;"='-';.-'>>' Q. ffj'jjDRO,ACU f4554 u
CRQ flCU "1451>> ~ o I CRD-POS-1271803 P Off H302 BZE6 2RN72 C e 8 248003; . N, 612. 4 ,>>17 ~ ' 0 POSITION SMITCH 02C12 - u 3 A 1 O' A
-; I u =,
>'" ~
'H528
.'R "l 528 LS/8 ~ 4 C4 CRD-HCU 1807+
CRD-PQS-1271811'<<,'," ' 02 ".;.'-elf)(6~<<$ )NVg~.'.i.';Kg~'us"'='W. ' I'u --.
'.-'~="i'N<<"4.---"--
'u
~e A,- Q'..~2itP0054~'~i< 'grig,"'~~WP '.>g~'->:.:r" =-",> -'.
~ ':-=.,",". q$ Z" ~,".::,";~>>::
$ MiTdtf'=. -.'-'='- =<"+'4<'P.-".=='<<'=:Rf=))k>f5/h.'4'~~- =~-- -:.=-'- ~ ("'-="'-:::"'~~';..~.';-"-, 'QslrtoN.
'I H302,
4 CRD-PQS-12T1815 '
BZE6 2RN72 " ' ".
~
8 248003 N 612 617 02C12 T
") ~
"' " u7 ~ '$r~s'> ~ '~~'g>"u' H5$ 8'~..<z t'. "," '-, -v "'::.'pe '. <<-W." ~--,<< .y ~Xi>><,4=" eu; ."
4 << POSI f1 QN. SMITCH 02C12 CRD-PQS-1271823
. 3 A 1 0, A , 'l528
;; H302 q.~".ffZE6~gffp)2"'";;-c--'~',~g::<<~4 'f.u -
528 LS/8 ' C4 A ff"=" 248003~..u/",tf" 612 f." ~~ CRO
'/ j'i";~~"=..-f*," '.:'.'..
HCU-1819+ e,": '.~
- I
"...- . ~ 1'jf'-
Poslrlatf;sliIT'H- ; F" ";:.:- <; -",'.~-.~-'-'.'-.~4->> $ 28>>$ 5/$ .4;:..>>I -.-;.'-'. ~', - -"- . ~ 'u"'";".~';,'--CRDLHCU i823i. ~ 0 C, 2 u CRD "POS-1271827 tl302 BZE6"2RN72 A 8 248003,= N 612 ~ 17 P W R- $ - u POLI fr ON SMITCH R 528 L5/0 'I ~ CRQ-HCU 1831+ 02C12 3 A 1 0 A H528 r4 i' CRD-PQS~I27]835,, POSIT/d'0 SMITCH 0 C "= ':-',~ ' A
- <<0 H/02 A.-
>--BZE6~(R(72.-
~.
.~ u-.
~
R
.I -;
q528.K2/8 ~ 4 A 8 248003,,'u
.. ~
tt": '612.'
.. ~' ",<<.>>'. "CAD"HCLJ"1835>, ~; u u" ." .";"'-'-.' P CRD POS l271839 H302 87E6-2RN72 A 8 >48003 t) 612 <<17 POSI I IO>>
02C12 -:=:
'M TCH A I 0,; A -'; =.',, H528 C4 u
u
'u
>>uu
-C 0 4 uu Posl f ION SMITCH R 52P K 2/I' 4 CRD "HCIJ-184 3+ 02CIP 3 A I 0 A HSPfl ('4
q 4 q "d
': '. ddd " F':-.:-" g ',: ip.'. "'-d'xllql'-'sfsS("1L t'qlJ~Cgg L!Sy t'"P,."';";'- '-'s>@",q.... . t, ~ ','"""...'DATE 0612)/Q3;. ~ ..'=
d'FG DESCRIPTION tlOOE L S E'ID TH HL TEST ANL FO C FRED . AOING DBE C HOURS d BLDG ELEV DETAIL ZDNE . Rootl'CCURACY COHPDSITE EPN POSI Tlotl
'. '4 SMITCH 02C12 3 A 1 0 A- ". '528l':528-K2/Ba4 R',
V d CRO "tlCU 1847+ dS 0
'k,4";;; '80'-."-> i '4.gp"..'l-CRD POS-.)2718M POSIT)Oft %)IT)At '
-;xfd .'x;f,' '~ ';P'~~~<<'5'p(xtttx j wp .<'4/IT"tftf!Ash'st'"*>>
ei,g3(2':="~.BgL'paktt'/ZAN:,'j ")g-.,q+ i'-. ~':g.'s-+" I '.=
2$
+~'8'i<"-,~x i--ass
",",'-,,~t.',.>";,-'- -;-..-'- '. ': '=;.',,-, " .";:.yi7x ",:".:;, '"
a d S Q <<'-'.-s;g',- '.8 ~jtx '. <<CRJl+ACg~i8514
>g',. dd'X CRD POS-1271055 tl302 BZE6~2tlN72 A 8 248003 "~ N =6i2'
~ 17 POSITION SVITCH x; R-.
'4 528 K2/8.4 ',; .;, ';.;., ';
3 ' " A 0 H528 -
,4 CRD-HCU-1859'2C12 1 ~ -. ~
d d CAD-POS:-)ppO>:;,". POSfTlhlf SQXTC3
~;. ";.",;,jjttd
~.-+/4"
=...sj s
fiji)".jtfi7a'~p.,y.:"~ -"i",:-:;q.;. $ .)~~a)'BOO ~
~""X'j:.-">>:-:""yp '=~8-'~':-;.~'g~s,'981'LStt Bi 4',"; s'->>ls.:.I.:-c'$ 4i~k'> tg;:~'4t~ ~'-."-
y.':tj'$'i$ ~i+;;;;-',~,',.",:PZ, -.--.,:,:: '=.:,,--..'.- -
=',-.,j7-. ~
<~'a'",>> "'.'<<'ithd "HCll~k4403>
2RN72,<<s; "-,
'd CRO-POS-1272207 H302 d82E6 -; 8. 248003 '612,,
02CI2 '= px. '. 3
'. A,
- p.$
s y'g'<<C t!7'~+ A hl ~ 17 pf," J s
'-<<g7'gals d$ .;Qtffpgg+); s
<g 'Sgs<<Ph@s'dbw.
$ LlpP.
POSITION SVITCH 528 g5/8 4 ~ ~," g CRD HCU-2211> O?C12 Cqo~pOS-.1272215., POSI Tlbg SVIT4H', 3
-,,'.,,1;-T;;.-
A I
~
O Huff', '.-';<<SZ56 x ftt7k'>>~~::p'4W" '-":.'4'-'-:"'9= 24tt0.4 '-"'<<N.>~ 2""' y+.-;..>>;->>.=,:ic-'.>~;r~+ =,<g -02k'g5tt8of,';.,",', >,~ >-I~'P:"i'< "."~-.j" ~ 'i"gY'<-."'~vltq,CRO~xKCU 2215
<<gi I I
<<<<x 0
~ . d CRD POS 0? C I?.
I?72214
/, 1 H302 0
IIZE6-kRN72
.--0 - -"sH52II '
'- c4 A 8 248003 N x-'---<<i"-'d~-
612 ttCI)=22L'L
~ 17 POSITION SMITCH O?Cl? 3 > I R 528 L5/8 ' CRO-HCU-2223+
0 A 8528 C4 CRD-POS 1(72227 $ 302 8?E6 $ RN72, A 8 248003 N <<-612-PDSITI Of( Slti'ICH 528 L5/Bi 4 R ~ CRO KCIJ 2227+ s. 'd',. CRO-POS l272?31 H302 BZE6 2RN7? A Il ?48003 N 612 ~ 17 0?C12 -. 3 1 I O A I) 528 d POSI Tl oft SMITCH R 5?R K2/0 ' CRO HCU-2235+ 02 C 1? 3 t I 0 A tl5?8 r4
K PR 0 AHC
<$ .'.""IfNf'Pg Pf,'AS)P1t.",EOU/PJtENT LI)l!>>","'- "':" -I>> .;"..-,:,-" ." '.'-"+>>>>.-.",>>.,:"-"".',;" ", ',DATE'06/$ 8/83--;i-"h;;...,,
HFG . ~
' HOOEY'
';.: E Olb TH 'L TEST ANL FO C . FREO 'AGING DBE C HOURS.
CONTRA T L V DESCRIPTION C S
=. ~ BLDG ELE) DETAIL ZONE fIOOH ACCtIRACY -,. COHPOSITE EPN
'RD i"
02C12 P POSITION SMITCH 722 3 A 1 0 A; '62$ @g..;
--<<""R,.) %28(K(/8 ~ 4
'~g C4*, "
~
~ r g .':<< ',~'...I ' CRb-HClJ-2239<
O
~;.,'s f 4'PN CRD PQS 272243 . ~ -- '-"'3'-:,",;-'BZ)6,<<h? N72 :";q.'II 94i> s 'gg,',,.4'- A. II'- $ 48005'+>+ jN g.fi}2T-'i+>>J."','>> -'+XV'"", j+ P' . /:.'>>
SlJf'[CN s.>>' ~ '- k~l,-;.0"." .&'.-'..~Q ,
=;(gpglf2/$ <<4,j~~.'i'. Ic~f)Q+pitI',:~jg-," $ ;~;-.-."sag:~~~tJ. $ R0.)CO."2?JI3>
t'l7.'OSITION
~ .
- I J
CRD-POS>>1272247 II30?:,BZE6-2RN72, A 8 248003 N . 6B "-=
-)*,,
" ' l7 ~ g i POS 0?c12, ~ -,,
M CH yis A>> h ga OF>>,'hA;" ;.>>-~. rs>>L~ a+. ) It52g~ gLh~ yjhhM hI f4 j>> '."'" .g i- hh<<>> . ~h. '8"7'.
~
V h - .'i ~ . <<h~ .V ~ "h > ~ 4 t POSITION SMITCH 02C12 3 A 1 0,'i A ., >
'- , II528 '=
328 K2/8 ~ 4
, "; C4 ~
- CRD-I)CU 2251s h
h CRQ-POS 1272255 '., ';>'tt802 ~>>';..SZE6C2>>RENT?-;- ~-". I'y %:.;.'='l%' = ',' 8 2)80)4'> +"-<,pi 661+4.'--4'---i" '~l>>"=;"-"":: - ".> ',;. i >17'i -* 1 POSITIONh 0" StttrCg, '"'."~",j.-<<";P.4-" ':.i 4'~~'".:5- ~pl.'528$ ?gl;4<'-'.--" '!@."-.4'-'s-'-J",:-"'-",-:=;+%~ ".,4j'$; ', ~ $ 80 HgU 22~~54h 0 CRG-PQS-1272259 S I
~ \
H302 '-; BZE6~?RN72, "- ', ' 'i ) A 8 248003 ' N 612 h
+
o'17 cR 0 POSITION SMITCH 02C12,
- 3 h 1 0 .",.A',,", -;, 4.; t1528>>,"
R l 528 "L5/8C4 ~ 4 CRD HCO 2603s CRO POS POS 1?$ 2607' i Tt QII: 8 V I TCM. ', -'."
~
'Hfdf. '.."P)F6qgR$ 72'; ."*';-.',~>>'.;.I-;" ~ >>f '
8 I '" -:;<<'.$ "'- j 7
~
~'>. ',.'-.-'.:~a.J 5jS;-L5/8'%'
8 24S00$ <<-'.;hh N.',:l6)2'J",'.'>>"-Y.'..i '..";ms'~,'.,"-.= !
. "....;;.':::;.,=.-;-.
-'."~~;-"-~ 4-~'" '".'.,gjD-:>>HCU!26O'14 , ~
v CRD" POS-1272611 H302 248003 BZE6 ?RN72 A 8 N 612 ~ 17 POS T t S H 02C12,'";;i O'," *A, I',0".-"-'-'A>>;.'-' "-j,":".'-'"'(">><<>> II528;:"".<<.'-g,w ..'4 C4 i.~>>>> >>;...",m-+.'>><<'- g ',,!2>>","=;. x -.'<<,>>>>:.I-,,"i.->>" .
->>>> -)r ~
CRO POS I 26 h ~ ! POSITION SMITCH R 528 L5/JI ~ 4 CRD-HCU-2615<< 02C12 H528 A 0 - A 1 C4 CRD-PQS4127261$ .~H)tJg,"...'BZC6 RN72~ ',"'" '. 248003 ' lf PnSITIQN-.SVtT 02C H s>> ...'..It ~,528 .t 5/8 h
=
~
A Q 4 612 CRp~iHCII+561)s s CR D-POS-1272623 PO ITIOJJ SM H302 BZE6-2PN72 sn4 A 8 248003 N 612 ~ 17 0 '2C12 " 4 A I 0-'- A c4 H.<<iR7 R pn 7 POSI 'IIO'I SMITCH R 528 L5/re 4 CRO-HCU-2627+ 02C12 3 t' H528 co O U A
w...0 w r 1 w ~ ~ I :4 EPN DESCRIPTION HFG ,,
-* MODEL DLOG,ELEff S
DETAIL E DID ZQNE TH = HL TEST ANL Po RODN 'CCURACf C ~ Pw RED
=
AGING DBE C , HOURS COHPOSITE EPN POSITION SVITCH R.'28L5/8 ~ 4 '-,i CRD-HCU "2631<
!3 "i OZCIZ A I 0 . A ." '528 ,
'i",, t."',--'.c: C4 w
~ CRO-Pbf;427)68'5; ~ .-'. I:,C.< Ii 'g~~gSZg6j4)NPg;; -".'4@;,'-.~4.',:.'~'h',.',L;O'~2 88j'j 'P'=='. '~~"6Pi~~-,< -:"'.".".:>:j;;,-y-:.-.1 ="<8*="';=: -.'
. '.sos I TIO Qw CRO-POS-,1272639 H302 *.8+6+2RN72; ~; p ~ ' 8 248003 .
N '"$ 12 ',: +17 0
-.. -02cig . +-.'-.:, >,~3,:.> ,{1g $ ggwg~P<E7$ 0)/gr'wgpp gVg~+y'wpfw '
)4 gg " '
+4 $ ww'ww www g ~~445 Qw >$ @C ki'wpww g y i)p%gg 1/p +gkjZ'G '1
+~~ . 1y -< 1! wQw PDSITIOII SMITCH 02C12 ,- '3 A,I'.0>> <<1'A."",
-:i";" .',.~
~ .:.=
N62$
- - '>R$
;=I; 28$ ',~
2/a.4,:,. C4 .
=
=
~
cRd Hcu-2643~
- ,cg DTf.as>>,a)zfg4'7".."'-,.-',.: ~~~8$ )P>>jyjzfjMiIPj7g';g>,'-.<~~;~~-",;~'",g:. ""- A',:g; w~
) CRO-POS<<1272651'H302 =.bZE6 zRN72 !w
'. ;( A 8 248003 '.;;612 '~w=,," -'.>> *'.
e 17 4 I 4 ~ = wl. w&M 'C( 1, -~ ~ -, .4 .m ~ - 1 1 ~ -w PaSI TION SNITCH ' 0.; A,";;-; ;R ~ 528 g2/8 4 ~, 1 .,' ',,CRD+tICU
-cho OZC12 P'gs pe's5 3,
A 1
":tt(og -=-'gZ 6"4 $ %" ""
. H528 j "-
g
.'-, =., ".:=,,-""- '.""+'1>.';,', 'y>> "-,'..",.-.,I>>",5'. ,ggd'. HClf 2659>.',
CRO POS 1273003 8302 8ZE6-2RN72 A II 248003 N 612 ", -'; " - -
~ l7 2cl2/ .-.w,'z~,3'-,=. A-,'-':,fi'0','r, A'-'-1~ '4 ~'('i<'<'."- ptI5)8I g' "=;~.',+. G4,'f- w =. - - ~'v "~'-,.);-, ~~,:'.-g;;; v.,'wl>'.
w P w.> POSITION SNITCH R 528 L5/Gs4 CRO HCU-3007+ DZC12 3, A 1 0 A H 528 C4
-CPD-Pos~1273oii . N30(' . 5Z(6 2RN72* ."',>>; ~ - A D 248003" 612 POSITtoN'SUItCH .0'," 528 L5/8+4 N
CRdiHCU~3DI ii CPD-POS 1273015 8302 BZE6 2RN72 8 248003 A N 612 ~ 17 DZCIZ lLHClLXILKa A 1 0 A H528 1 " 41 w PnSITION SMITCH 8 528 L5/8 ~ I CRO-HCU-3019+ 02C12 3 A I 0 A 8528 C4
~ w: ~ 4JltV
~ \ <<
;= '",;.": .W,P,-;."t>>$ '>>g: >~-->>t." -ItHP,-" PAS$ ;:,$ K uKOV$8N>>EWJ- LIST,-;:.,'~'~:.'." .'4>>,,;',-,-<~"..~,~=,>:;-. -;:-.",. OAT E 06/28g83:..", .-'; >>
EPN HFG HODEL ~ S E OID THHL TEST ANL Fo C. FRKO AGING OBE C HOURS >> C C NTRAC DESCRIPTION BLOG ELE) DETAIL ZotIE. ROON ACCURACT COHPOSITE EPN I ~ Pnsl TloN sMITcH 02C12 3
'.:"R",,""528,LS/8 F 4 ""= -.,;-.>>','..j -'-'--,, . ', :, CRO HCU 30234 CRD POS IPT3027>>;-
POSTTI0$ .SMI)CH- +i"C,P >. ",'>>0" '" ".'>:-,":.-'k ':l '
~
+@ '-ging'g'8/'8~4'-."$ 4j".>>"- ~J' g~'<>>A~'q'='.~ .>>'y~'~~>><<'.,".:CRO~tICU<5027k 02C12 i' >>3 '.
CRO-POS I osl T 1273031-nN s H'
'02>> P?E6,2RN72'; '-. l A 8 248003,: ""'- N '. $ 12*-' "'
02C12." -r.; .43 ~,-. <<A,>>"-"~
$"'0>> T.:-.K +M.LQ~~>>>> '.Rip-'3t5$ 8".R;;, M.;i <<a" C4 '-'j<<+-~ '-<<P;r>>~'; <<c"'>> ~'r-i/Ag-.:-<<~'. -'.>>."-.- <.; ."-'-
- "~ 0 612'4 V * , ->>,P POSITION SMITCH 02C12 , 3 A . I 0.," A i .. ' -'..'..N528,.
- -'R I Q'. 528;K ~ 2/347
'C4
- CRD HCU 3035>
, ~
CRO<<POS-.12PO)P, I Iiog,;"qlilTcII;
~-'os I:-'~ lgPR'>>""';y(
fi2Rg72"'>><<>>-.~'r'>>'i
'14; 'g <w"'. >><<'~5'.'-4~y %.-"Q>';f28'~%jag/3'iT'-
~,.':>-~ -'~Up:;. ~, l-.II 2{8(034;. =-"=.;P, '61 - <<4<<" >>"=~==': 'g".'4
-: "'>> '~I>>~~'>>'>>".-+"j~l~+..~>>., ",:>
= =
'.=-'A~X-'yN-~'RAB HCU-30394
-.:- .-,", '.-)ll t 4 02CI C'
CRD "POS-1273043 '302 BZC6"2AN72 "' ' " A 8 -2{8003 ' PAS 0 M TCH 02C12 0 P 02C12 a POSITION SMITCH ' 3 "A '.0 " A.<
~ -, '
'; ~
g,8528
'l .
' '4
$ 28 K ~ 2/3r7 =
h 8'- 2)8003'j -"'<< '<<.,61('":+',u <'<<~'P'-'4>>"4~ .<< -"Tr'-'= -'" ' ' CRD~HCU-3047+ CRD-POS 127305L -- $ H3g. '-<<QNS~4RII72 '<<sj'>j~; '= >>', '"'-"'1 f"
' =
'= -'<< "
POSITION" SMITCH, := '- $ *.~ ".VI';">>'""w'I","",ge"~gr- <<<<.."v'0 558",K>2/3t7>><<.'". " .'-;-~<5-'-:X~4'":w,', ~"~-.." g..'~ g+'0,:ICgb~ACU 50514.. C 8'4800$
>>..'RD-POS-1273055 H302 BZE6 2RN72 "
A '. g -612 ~ 17 POSIT ON 0 CH 02C12 >> ~ CRD PO 30>> <<H 0 4 Pos! TI ON SMITCH R 528 K~ 2/3 ~ 7 CRO HCU-30591 02C12 3 8528 A 1 0 A U C4 P CRD-Pos-1273403 ; :H302:QZE6-2RH7$ : . '. ':,,". " '- -'" A 8 248003 r>>". POST TIOM SMITjll' -, ".R, 828 LES/347 '-- CNIo-Ht',U-){03t<< 1 CRO-POS-)273407 H302 BlE6-2RN72 A R 248003 N 612 POS T N <M
~ 17 02C12 A 1'0- A H528 C4 R -POS-P081110ll SMITCH 02CI 2 3
~ R 528 La 5 I I a7 CRD HCU 3411+
0 A 8520 re
k O g;-$ !( "I~<~~&='->>',",
~~2004 ;-
. .P'".: t".!'"-..TPNP~2,CL4@f f K
$ 0U jPttKttT f.IBJ.>>':,-' ~L "~'/,.".'~F,',>> DbiE 06i iala ,
,F
'.IIODKL K. 0 ID . TH EPN CXEI DESCPIPTIdN E
HFG
~ ...,:
'= "
, ~ BLDG KLEtt DETAIL ~
ZONE '60H HL- TEST ANL FO ACCURACY C
' FREQ AGING DBE C COHPOSITK EPN HOURS POSITION SNITCH 02C12 3 A 2 0 A ~,, ., ' ' 428 L ~ 5'/3o7 HS28$ '.. $ ':i";,'=-: ~
Ch. '-;
',.~ CRO~HCU"3425+
49.., ".';-;S302,-."-.--;@yQj'/>Sa~f~~= P;;.;,">g'<<~ ',r-.:.;4,".g~kgattyk~;,;-'S > af21.'~"-.<.,--,'~,.g, '-'.-<'.::,- P..'-;- '- -"..;-";- -".="-.';~>> -
'Po-PQS-12t~h TIO8-'BVITa9,;: -<',":~;.4 -.-".~-.-;..;.~--.,-~..;,'.:.P-t."'~~~ca '-; 2S"X<S'P3'+~.-;-.>>F,~: J~~,=4:4"-.'.~'"-.";-,"'.P"., ",;~;-;-~.";;:-'.-"-tjty'-tjcui3'hied;
'.-,'.-'OST r
CRD-POS-1273423 H302 w
~
'. 'B2K602RN72 s'>>F; '-i A B '48003 . N 612 .
, .!, ~
POSITION SNITCH 02C12 3 A I'>> A ~ -' "'. "'F'S28 ' Rl 528';Lo5/3a7
~ ~ I Ch
"., ', ~ .;;;.".. ':: ' L -'CRD"HC0-3427+
F CRD POq-.)'2734yg.>>, ...",. q yy0 C>>~)g(yy tj'lq'*/="~@~-'~ q~.'t à 4S"' $ =>><khagh5~'
+ 6jg~~~'."-"~+7- ~-""- ='- - =
.. -. :.,->>,;brig:,:.'".=,
.PQ I l1Dg gttf Cg . F z" >>
$ >> 'Ig~'>> +>>>> Q =Li >>~gh) 't"(i'Cjg<'pe II28~g P2 f3 $ . '.,'l F I, F+x~7'v'Fh~g @<g~+i~g~'T~4~~
~
-K i ' '"'
CRD "POS-1273435 <<8502 'ZK6>2RH72 > = A 8 '-24800$ ' g 42..!, of7 gg)g-,i @,-,."X-.C).i",..1 ->>;-.~.~F.- ~>>~'+:+~ ....'- >>~ ='-e'.g-',@>>-".'F,-*;:,I -~ POSITION SNITCH 02C12 " 3 A 2,0 ', A;,~ .. ' -'H528" . 525
~
K~ 2/3 Ch t7', .. c ~
' -. =':"- ~
CRD~HCU "3439~ 0 CRD-P94:III73443.:. POSITI O4'-.BVIT6g -" '-
-:,,-~q =-:";-': ",~P-=g "."-=--;".-'~j~828.kihlb'l4 .,';.,',I ;=,', i'-;P. $ ,;;=,V': '..-',:..": '-~P~-.: .:,'YRD"H6U".3<<>'; ', *
.'-. = -.-
CRD-POS 1273447 H302 82K6-2RN72 A 8 248003 = N 612 ~ 27 02C22 f F ' \ POSITION SMITCH R 528 K~ 2/3 ~ 7 CRD HCU 345)+ 02C12 3 H528 0 Ch A 1 A CRD POS '273455 tl$ 02 BgK6~2RHy2 8 248003 612 .- "- F. .427 POSITION. SMITCH It;528 K '/3 ' A
~ .,
", -;.. ".' -'i,CRD~HCU "3455>
1273459 248003
'RO-POS H302 BZK6 2RN72 A 8 N 612 ~ 17 02C12 A 1 O'A H528 r.h 0
POSITION SNITCH R 528 L ~ 5/3 ~ 7 CRD HCU 3803' 02C12 1 0 A H528 r4
V <<g~~>."
.g ~g IIPg ~> L'p '1K-'KQIII8$ENT<b'R .~~3~:~ pic~>~~~i. 5'/pYr'~a>~<<~i~'h<<Py~~<<,g,;~ISSSO/YES/06/2883pgbi+~~)
-,':" '>>>> EPtL',";-'. ',,OKSCR/PTIOtt
~ '.,';HFG;-"I,;:~;;.'.'"F~
..<<',!', f'-,'=.,
.'."'-'; HODEI, -'"'":.'-"-~' 'S E 410 ";"'.Tg'"HL: TEST, AIIL FDIIC: '>>" FRE43.','AGING DBK "C '", IIOURS '; <3,-', I
~,,'.<<~';t. BLDQ,'ELKS I DETAIL ZONKI.,<3 Rggtt',a, 3
'ACCURACY, ':,j,", --3;<<is<<','..~r COHPOSITE EPN'-." ~
"=
CONTRACT ' LK E ..'EC -- US 1 h Y I ~ ~ S<<t g j,CRO POS.12'$381'I>;:.-'4 .'<<3-' H3,,'QI 8+gi QRN72",i Y" 'j e<$ ">>'.<<'.> TSS; 'AS II j -2480Ps~" g V $ II~~~W+. ">V '&/$9X ~~JPgj'j "AI".PP'"- t,-~ 'tsim3~
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