ML20134C940
| ML20134C940 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 01/28/1997 |
| From: | Donnelly P CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-96-06, GL-96-6, NUDOCS 9702040173 | |
| Download: ML20134C940 (16) | |
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Consumers Patrick M Donnetty Plant Manaw MERING nlKNDGAN5 PROGRE55 Big Rock Pomt Nuclear Plant.10269 US 31 North. Charlevoia, MI 49720 January 28, 1997 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-155 - LICENSE DPR BIG ROCK POINT PLANT - 120 DAY RESPONSE TO GENERIC LETTER 96-06: ASSURANCE OF EQUIPNENT OPERABILITY AND CONTAINNENT INTEGRITY DURING DESIGN-BASIS ACCIDENT CONDITIONS On September 30, 1996, Consumers Power Company received U.S. Nuclear Regulatory Commission (USNRC) Generic Letter 96-06: Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions dated September 30, 1996. The NRC issued this generic letter to address the identification of several safety-significant issues that have generic implications.
Addressees have been requested to determine:
(1) if containment air cooler cooling water systems are susceptible to either waterhammer or two-phase flow conditions during postulated accident conditions; (2) if piping systems that penetrate the containment are susceptible to thermal expansion of fluid so that overpressurization of piping could occur.
In addition to the individual addressee's postulated accident conditions, these items should be reviewed with respect to the scenarios referenced in the
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generic letter.
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If systems are found to be susceptible to the conditions discussed in the generic letter, addressees are expected to assess the operability of affected h
systems and take corrective action as appropriate in accordance the f'
requirements stated in 10 CFR Part 50 Appendix B and as required by the plant Technical Spectficatfons.
9702040173 970128 PDR ADOCK 05000155 P
NUCLEAR REGULATORY CONNISSION 2
BIG ROCK POINT PLANT REPLY TO GENERIC LETTER 96-06 January 28, 1997 Requested Information Within 120 days of the date of this generic letter, addressees are requested to submit a written sumary report stating actions taken in response to the requested actions noted above, conclusions that were reached relative to susceptibility for waterhamer and two-phase flow in the containment air cooler cooling water system and overpressurization of piping that penetrates containment, the basis for continued operability of affected systems and components as applicable, and corrective actions that were implemented or are i
planned to be implemented. If systems were found to be susceptible to the conditions that are discussed in this generic letter, identify the systems affected and describe the specific circumstances involved.
The Big Rock Point Staff has determined that:
(1) The containment air cooler cooling water systems are not susceptible to either waterhammer or two-phase flow conditions during postulated accident conditions; and (2) Two piping systems that penetrate the containment are susceptible to thermal expansion of fluid so that overpressurization of piping could occur.
These determinations were made with the benefit of the questions and answers from NEI/NRC meeting of October 29, 1996 that were forwarded by NRC Letter dated November 22, 1996, and a Memorandum to the Division of Reactor Safety (DRS) Division Directors dated January 24, 1997.
The Big Rock Point Staff will perform an operability assessment prior to plant startup from the current forced outage.
The assessment is expected to conclude that the affected piping systems are operable, and the conditions identified above will not be in conflict with the Technical Specifications or the Updated Final Hazards Safety Analysis.
Pending further evaluation, the issue will be resolved during the 1997 refueling outage.
A preliminary investigation has concluded that the plant may have been operated in a condition outside the design basis.
This event is reportable pursuant to 10 CFR 50.72(b)(1)(ii).
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The requested summary report is provided as an attachment to this letter.
/Ly)5,n V
Patrick M Donnelly Plant Manager CC: Administrator, Region III, USNRC NRC Resident Inspector - Big Rock Point Attachments
CONSUMERS POWER COMPANY Big Rock Point Plant Docket 50-155 License DPR-06 Response to Generic Letter No 96-06 dated January 28, 1997 At the request of the Commission and pursuant to the Atomic Energy Act of 1954 and the Energy Reorganization Act of 1974, as amended, and the Commission's Rules and Regulations thereunder, Consumers Power Company submits our response to NRC letter dated September 30, 1996, entitled, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions". Consumers Power Company's response is dated January 28, 1997.
CONSUMERS POWER COMPANY To the best of my knowledge, information and belief, the contents of this submittal are truthful and complete.
act e /s' y
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u By m m-t Patrick M Donnelly V
Plant Manager Sworn and subscribed to before me this 28th day of January 1997.
0 n o u /t, YhNW nni Lynn Helms, Notary Public Ch evoix County, Michigan My commission expires August 29, 1999.
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ATTACHMENT 1 CONSUMERS POWER COMPANY i
BIG ROCK POINT PLANT i
DOCKET 50-155
SUMMARY
REPORT FOR 120 DAY RESPONSE TO GENERIC LETTER 96-06 i
ASSURANCE OF EQUIPMENT OPERABILITY AND CONTAINMENT INTEGRITY DURING DESIGN-BASIS ACCIDENT CONDITIONS Submitted January 28, 1997
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i 5 Pages
SUMMARY
REPORT - 120 DAY RESPONSE TO GENERIC LETTER 96-06 2
j l'icensee Respons_q ACTIONS TAKEN TO RESPOND TO THE REQUESTED ACTION The Updated Final Hazards Summary Report, Technical Specifications, Operating Procedures, equipment specifications and drawings were evaluated to determine if Big Rock Point's containment air cooler cooling water system is vulnerable to either waterhammer or two-phase flow conditions during any postulated design bases accident (DBA) conditions. This evaluation was based on the most limiting DBA for Big Rock Point, the Loss of Cooling Accident (LOCA) analysis, which includes break sizes up to and including the complete severance of the largest pipe in the system. Loss of offsite power, and the worst single failure of Emergency Core Cooling System (ECCS) Equipment (the emergency diesel generator (EDG)), was also assumed in the evaluation. The evaluation also considered a LOCA without a LOOP, and a LOCA with an automatic transfer of the 138kv line to the 46kv line.
Analyses / evaluations were also performed to identify those safety and nonsafety-related piping systems (Attachment 2) that penetrate containment which may be subject to overpressurization during the most limiting DBA, which in this case is a Main Steam Line Break (MSLB). To determine susceptibility, piping systems were considered to have sufficient protection against overpressurization if they contain air, gas or steam as the fluid, where pressure relief is provided either by relief valve or globe valve, where an expansion path (such as to a tank) is available, or where appropriate pressure locking and thermal binding modifications were made.
In addition, a structural stress analysis was performed to demonstrate that overpressurization and failure of selec+ed piping inside containment would not be credible. All nonsusceptible p a systems were verified to remain within limits allowed by Code.
CONCLUSIONS REACHED RELATIVE TO THE SUSCEPTIBILITY FOR WATERHAMMER AND TWO-PHASE FLOW IN THE CONTAINMENT AIR COOLER COOLING SYSTEM Big Rock Point containment heat removal design differs from the licensees described in the GL. Containment heat removal following a postulated design bases accident is accomplished by the containment spray system. The Big Rock Point containment air coolers and the associated Service Water System (SWS),
i are classified as nonsafety-related. No credit is taken for the SWS for mitigating the consequences of a LOCA. Cooling and air conditioning components utilizing service water are not considered essential equipment in order to achieve safe shutdown following a LOCA, and are classified as functionally nonsafety-related with regard to containment cooling.
The SWS is not susceptible to either waterhammer or two-phase flow conditions during postulated design bases accident conditions because of the following:
- 1) The system is rendered inoperable during a loss of offsite power - no flow, no waterhammer, no two-phase flow; or l
- 2) The system continues to operate during a LOCA if offsite power is not lost. Flow continues, no waterhainner, no two-phase flow; or
- 3) The system cycles on/off (SWS pumps deenergize, then autostart) during a LOCA following a 6 second automatic transfer between the two offsite power sources if one is compromised - flow is not suspended long enough to incur waterhammer or two-phase flow *.
SUMMARY
REPORT - 120 DAY RESPONSE TO GENERIC LETTER 96-06 3
The SWS feeds various equipment in parallel through 2, 4, 6 and 8 inch o
lines and drains to the lake discharge canal via a 12 inch line (the pumps are rated at 2100 gpm with a discharge pressure of 40 psig). When the service water pumps are not operating, a significant portion of the system drains to the lake. In addition, the low system pressure and large volume results in slow fill rates, precluding waterhammer events.
Although the SWS is the normal heat sink for Spent Fuel Pool (SFP) Cooling, this system is not relied upon for cooling following a DBA. As discussed in section 9.I.3 of the Updated Final Hazards Summary Report and section B(2)(a) of the Operating License, adequate SFP cooling is assured following a DBA via a SFP makeup line which provides flow to the SFP when the Core Spray Recirculation System is placed in service.
CONCLUSIONS REACHED RELATIVE TO THE SUSCEPTIBILITY FOR OVERPRESSURIZATION OF PIPING THAT PENETRATES CONTAINMENT Big Rock Point's functionally nonsafety-related demineralized water and treated waste to spent fuel pool lines which penetrate containment were determined to be susceptible to overpressurization during a DBA. These systems are not safety-related and provide no function associated with LOCA mitigation. Actions to provide overpressurization protection are discussed under " Corrective Actions That Were Implemented or are Planned to be Implemented".
All other safety and nonsafety-related piping systems that penetrate the containment and are susceptible to thermal heating following a LOCA were considered to have adequate protection against overpressurization. The basis for this conclusion is that the lines contain air, gas, or steam as the fluid, pressure relief is provided either by relief valve or globe valve, an expansion path (such as to a tank) is available, or pressure locking and thermal binding modifications to compensate for this condition were made, were.
i BASIS FOR CONTINUED OPERABILITY OF AFFECTED SYSTEMS AND COMPONENTS The Big Rock Point Staff will perform an operability assessment prior to plant startup from the current forced outage.
The assessment is expected to conclude that the affected piping systems are operable, and the conditions identified above will not be in conflict with the Technical Specifications or the Updated Final Hazards Safety Analysis.
Pending further evaluation, the issue will be resolved during the 1997 refueling outage.
A preliminary investigation has concluded that the plant may have been operated in a condition outside the design basis.
This event is reportable pursuant to 10 CFR 50.72(b)(1)(ii).
Discussion I.
Treated Waste to Spent Fuel Pool Line Function Treated waste is considered a normal spent fuel pool makeup supply and is only one of three independent sources of makeup. Other sources are demineralized water and safety-related fire water. The treated waste makeup function is,
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' SUMARY REPORT - 120 DAY RESPONSE TO GENERIC LETTER 96-06 4
therefore, nonsafety-related. However, the line does penetrate containment.
The safety-related containment boundary is protected by isolation valve CV-4049, Treated Waste to Fuel Pit inboard containment isolation valve, which is controlled by the Containment Isolation System (CIS), and will automatically close the isolation valve in the event of an isolation scram or loss of power.
The isolation valve supports the safety-related containment isolation function performed by the CIS and is, therefore, a safety-related interface.
Possible Failure Mechanisms The functionally nonsafety-related treated waste line was determined to be susceptible to overpressurization during a DBA. A preliminary overpressurization stress analysis indicated that treated water pipe hoop stresses would exceed their ultimate value assuming the consequences of a 0.63 ft2 main steam line break, solid pipe, and zero leakage of the valves.
Operability Determination During a DBA, the treated waste to spent fuel pool line will be isolated by the automatic closure of CV-4049. This line is safety-related for containment integrity only; it is not essential for safe shutdown. Procedures will be revised to provide surge protection for this line during a postulated DBA prior to plant operation from the current forced outage.
Therefore, with the procedure changes discussed above, there is assurance that overpressurization of the piping during postulated DBA would not be credible.
The treated waste to spent fuel pool line is operable.
II. Dominert.lized Water Line Function i
The Demineralized Water System (DWS) maintains an inventory in the i
demineralized water storage' tank to provide makeup capability for the Nuclear Steam Supply System (NSSS) and reactor auxiliary equipment. Maintenance of the inventory is not considered a safety-related function; rather, loss of inventory degru;ec plant reliability. During a LOCA, Emergency Condenser makeup from tre owS is at required. Safety-related makeup is provided by the Fire Protection System (FPS). Therefore, the DWS is not required. However, a segment of the demineralized water line penetrates containment in two places.
The safety-related containment boundary is protected by CV-4105, Demineralized Water Sphere Isolation Valve; which is normally open, and VMU-300, Demineralized Water Isolation Check Valve. CV-4105 will close as a result of a loss of power or loss of air supply. VMU-300 will close upon loss of inward propellant force (i.e., low demineralized water pressure). The other penetration is the demineralized water fill line in the shell of the emergency condenser. This short segment of piping contains CV-4028, Demineralized Water to Emergency Condenser valve, and its associated check valve, VEC-300, Demineralized Water Check valve.
Possible Failure Mechanisms The functionally nonsafety-related demineralized water line was determined to i
be susceptible to overpressurization during a postulated DBA. A preliminary overpressurization stress analysis indicated that demineralized water pipe hoop stresses would exceed their ultimate value assuming the consequences of a 0.63 ft2 main steam line break, solid pipe, and zero leakage of the valves.
SulWIARY REPORT - 120 DAY RESPONSE TO GENERIC LETTER 96-06 5
Water will be trapped in a pipe segment between CV-4105 and VCU-300, and 1
another segment of water will be trapped between CV-4028, and its associated i.
check valve, VEC-300, Demineralized Water Check valve.
This line is safety-related for containment integrity only; it is not essential for safe shutdown. To provide long term overpressurization 1
protection for this line, further analysis is required. Additional time is j
required for engineering, procurement, planning, scheduling and implementation of design modifications if required. For the short term, the Big Rock Point j
staff believes that overpressurization protection for the demineralized water line is provided by known and quantifiable leakage of CV-4105. Leak rate trending indicates that some leakage below the allowable rate can be predicted. Leak testing of CV-4028 will be conducted to validate adequate seat l
leakage, or the valve will be adjusted to ensure seat leakage occurs prior to l
exceeding code allowable stress for faulted conditions as delineated in Appendix F of ASME Section III. This limited amount of leakage is adequate to j
prevent overpressurization. Therefore, based on the discussion above, there is reasonable assurance that overpressurization of the piping under a postulated DBA would not be credible, Based on the discussion above, and the fact that the Demineralized Water i
System continues to satisfy the criteria as described in the Updated Final Hazards Summary Report (UFHSR 9.2.1.4.3; 9.2.3; and 10.4.5.3), the i
i Demineralized Water System is operable. The issue of providing surge protection for this line will be resolved during the 1997 refueling outage.
3 Plant startup from the current forced outage may commence.
l CORRECTIVE ACTIONS THAT WERE IMPLEMENTED OR ARE PLANNED TO BE IMPLEMENTED i
Subsequent to an extensive review by the Big Rock Point staff the following i
procedure / configuration control enhancements, procedure revisions and modification evaluation will be made to ensure that piping will have surge protection during operation:
Procedure / Configuration Control enhancements a.
Standard Operating Procedure (SOP) 3, Reactor Cleanup System, will be enhanced to ensure that the resin disposal line is drained following resin transfer operations to preclude thermal expansion:
b.
S0P 25, Heating and Ventilation System, will be enhanced to ensure that the system is not isolated during power operation.
c.
S0P 28, Station Power, will be enhanced to ensure that the SWS service system is returned to operation in a manner to reduce the probability of a waterhammer.
Procedure revisions a.
S0P 3, Reactor Cleanup System, S0P 9, Fuel Pool System, S0P 11, Radioactive Waste System - Liquid, and 0-TGS-1, Master Checklist, will be retf sed. The revision will ensure that a section of piping will be vented prior during plant operation to provide a relief path for treated waste return line globe valve CV-4049 venting under postulated post accident conditions.
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SUMMARY
R'EPORT - 120 DAY RESPONSE TO GENERIC LETTER 96-06 6
Plant modification evaluation a.
A surge chamber may be installed on the demineralized water line to provide overpressurization protection, which would ultimately ensure containment integrity. However, additional time is required for engineering, procurement, planning, scheduling and implementation of design modifications if required.
IDENTIFICATION AND SPECIFIC CIRCUMSTANCES INVOLVED FOR THOSE SYSTEMS FOUND TO BE SUSCEPTIBLE The functionally nonsafety-related demineralized water line and treated waste to the spent fuel pool lines were determined to be susceptible to overpressurization during a DBA. These lines are safety-related for containmant integrity only; they are not essential for safe shutdown. The Big Rock Point staff has determined that the treated waste to the spent fuel pool line would have been able to perform its required safety function, specifically containment integrity, before overpressurization could occur.
However, A preliminary overpressurization stress analysis indicated that demineralized water pipe hoop stresses would exceed their ultimate value assuming the consequences of a 0.63 ft2 main steam line break, solid pipe, and zero leakage of the valves. Therefore a one hour non-emergency report was made to the NRC Operations Center pursuant to 10CFR50.72(b)(1)(ii)(B), Condition Outside Design Basis.
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I ATTACHNENT 2 CONSUNERS POWER COMPANY BIG ROCK POINT. PLANT DOCKET 50-155 CONTAINMENT PIPING OVERPRESSURIZATION EVALUATION (120 DAY RESPONSE TO GENERIC LETTER 96-06)
Submitted January 28, 1997 6 Pages
l ATTACMENT 2 Page 1 l
CONTAHNIENT PIPING OVERPRESSURIZATION EVALUATION l
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I SERVICE /.
P0TENTIAL:TO _
PENETRATION #'
' DESCRIPTION TYPE FLUID' OVERPRES$URIZE7 COMENTS..
I 12' Equipment Lock Air No Hydraulic lines penetrate containment through Hydraulic the lock. The hydraulic fluid in these lines Electrical vent to the reservoir located inside the lock.
2 7'7" Personnel Lock Air No Same as #1 l'
Hydraulic Electrical 3
5'6" Escape Lock Air No Pressure should not exceed 23.5 psia Electrical (8.5 psig) l 4
Manhole None No No fluid present 5, 6 Eliminated None N/A N/A (No Penetration) 7, 8 24" Ventilation Air No Maximum pressure following a DBA is less than Supply & Exhaust 8.5 psig 9
14" Emergency Air No This penetration does not isolate.
t Condenser Vent Steam l
l 10 12" Main Steam Steam 90 Fluid initial temperature greater than DBA j
temperature.
11 10" Reactor Water No Liquid is not trapped - discharges through t
Feedwater check valves to steam drum.
12, 13 Service Water Water No Liquid in the containment boundary piping is Return and Supply not trapped. There are no isolation valves.
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14 4" Space Heating Steam No Steam temperature is greater than DBA Supply temperature.
15 2" Enclosure Dirty Water No Containment isolation globe valve vents in Sump Discharge direction of flow into an expandable volume of pipe, providing surge protection.
ATTACHNENT 2 Page 2 CONTAINNENT PIPING OVERPRESSURIZATION EVALUATION SERVICE /
POTENTIAL T0 PENETRATION #-
DESCRIPTION TYPE FLUID OVERPRESSURIZE7 COMENTS -
17 3" Treated Waste Water YES Water can be trapped between inside Return containment isolation globe valve CV-4049 and a normally closed manual valve VRW-52 outside containment. Standard Operating Procedures S0P 9 and 50P 11 will be revised to ensure that a section of piping will be drained or vented to provide a vent path for CV-4049.
This action will provide surge protection for this piping.
18 2" Demineralized Water YES Water can be trapped between CV-4105 (outside Water Line containment) and normally closed system (inside containment). A surge chamber may be added to the piping to provide surge protection depending;on evaluation.
19 3" Space Heating Water No S0P 25 will be enhanced to ensure that the Return system is not isolated during power operation 20 2" Instrument Air Air No Air in the portion of the piping which forms the containment boundary is not trapped.
It will flow to the header outside containment.
i This header has relief valve protection.
21 2" Enclosure Clean Water No Containment isolation globe valve vents in i
Sump Discharge direction of flow into an expandable volume of pipe, providing surge protection.
22 2" Reactor and Fuel Water No Containment isolation globe valve vents in Pit Drain the direction of flow into an expandable volume (Radwaste).
23 2" Resin Sluice Water No S0P 3 will be enhanced to ensure that the line is drained following resin transfer.
24 Spare N/A N/A N'A 25 2" Service Air Air No Pres. following a DBA will be < code allow.
26 Spare N/A N/A N/A 27 Backup Post Water N/A Piping protected by RV-5077, RV-5078 and RV-Incident 5082.
ATTACHNENT 2 Page 3 CONTAINNENT PIPING OVERPRESSURIZATION EVALUATION SERVICE /.
POTENTIAL TO PENETRATION i DESCRIPTION
-TYPE FLUID OVERPRESSURIZE7 ColglENTS 28, 29 Core Spray Pump Water No Piping is open to containment atmosphere and Suction has no potential to trap water.
30 Spare N/A N/A N/A 31 Inhibitor Water No This line is no longer in service.
Inside Recirculation Line containment, it is sealed with a spectacle blind; outside containment, closed valve VEH-8 and a pipe cap seal the line. During performance of SC 96-04, which cut and capped a 4" branch from the 6" inhibitor recirc line, water was drained from the system to permit welding. This provides an air pocket which will prevent overpressurization.
32, 33, 34 Spares N/A N/A N/A 35 CRD Supply Water No CRD pump suction relief valves provide over pressure protection.
36 6" Post Incident Water No Same as penetration #27.
and Fire Water Supply 37 Main Steam Line Water No This line has been removed from service, and Drain Steam is isolated on both sides of containment.
This line was drained and has been isolated for over 30 years 38, 39 Spare N/A N/A N/A
ATTACHNENT 2 Pag' 4 e
CONTAINNENT PIPING OVERPRESSURIZATION EVALUATION SERVICE /-
POTENTIAL TO PENETRATION #
DESCRIPTION
-TYPE FLUID-OVERPRESSURIZE7 CONNENTS-40, 65 Electrical Nitrogen No RDS electrical penetrations are normally.
81, 83 pressurized to 27 psig (28 psig maximum) and are sealed.
In a DBA environment, the internal pressure could increase to 51 psig.
(This assumes initial temperature is 0*F; if initial temperature is 60 F, pressure would increase to 43 psig).
Per EEQ file 4.90, these penetrations are qualified to 180 psig.
41 thru 64, Electrical No Fluid No These penetrations are sealed with potting 66 thru 76 compound.
77 ILRT (Blowdown)
Air No Cannot pressurize to greater than 8.5 psig.
78 thru 80, Electrical No Fluid No These penetrations are sealed with potting 82, 84 thru compound.
87 88 ILRT Air No Same as penetration 77.
89 PS-7064A Air No This penetration is open to containment atmosphere. There is no potential to trap fluid in the piping.
90 PS-7064B Air No Same as penetration #89.
91 ILRT Air No Same as penetration #77.
92 Nitrogen Supply To Nitrogen No Piping is protected by RV-5092 and RV-5093 if RDS Valves CV-9472 is open.
If CV-9472 fails closed, and assuming an initial temperature of 40*F, and initial pressure of 90 psig (maximum allowed by SOP-18), the maximum post-DBA pressure would be 136 psig.
Per FC-572A which installed the containment boundary
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piping, the piping is rated for greater than 600 psig pressure.
l 93 thru 95 Spare N/A N/A N/A 96 PT-173 Air No Same as penetration #89.
ATTACHNENT 2 Pcg'e 5 CONTAINNENT PIPING OVERPRESSURIZATION EVALUATION SERVICE /
P0TENTIAL TO PENETRATION #
' DESCRIPTION TYPE FLUID' OVERPRESSURIZE7 COMENTS -
97 Ventilation Probe Air No Isolation valves SV-9155 and SV-9156 are both outside containment.
Inside containment the line is open to containment atmosphere (via DPC-9071 which offers minimal resistance to pressure buildup).
98 PS-636, PS-665, Air No Same as penetration #89.
PS-667, PT-174, PT-187 99 PS-637 Air No Same as penetration #89.
PS-664 PS-666 100 Electrical No Fluid No N/A (Lighting) 101 Telephone No Fluid No N/A 102 Spare N/A N/A N/A 103 Spare N/A N/A N/A 104 Lighting Electrical No N/A No Fluid 105 Telephone No Fluid No No 106, 107 Spare N/A N/A N/A 108 Lighting No Fluid No N/A 109, 110, 111 Spare N/A N/A N/A 112 2" Core Spray Pump Air No Line is open to cont.:Nment.
Vent Water 113 4" Core Spray Pump Water No Line is open to the spent fuel pool, and has Discharge same RV protection as penetration #27.
114 Nanhole None No N/A
ATTACHNENT 2 Page 6 CONTAINNENT PIPING OVERPRESSURIZATION EVALUATION SERVICE /
P0TENTIAL TO PENETRATION #
DESCRIPTION TYPE FLUID OVERPRESSURIZE7 C0f0lENTS -
115A, 115B Electrical Air No Per EEQ file 4.86, these penetrations are (ASD Electrical qualified to 65 psia. Assuming these Penetration) penetration assemblies are pressurized to 27 psig and that the post-DBA temperature of the assembly is 235'F, the maximum internal pressure would be 62.1 psia.
(Reference FC-462J3 and QUADREX Report 1-81-855, page 4-2.
This report, which provided design input for designing containment penetrations H-115A/B, is the basis for selecting 235'F as the maximum temperature of the penetration assembly. )
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