JPN-97-019, Provides Summary of Detailed Evaluation of Postulated Scenarios Described in GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Conditions & CA for JAFNPP IAW Ref Commitment

Provides Summary of Detailed Evaluation of Postulated Scenarios Described in GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Conditions & CA for JAFNPP IAW Ref Commitment

ML20148E608
Person / Time
Site:	FitzPatrick
Issue date:	05/27/1997
From:	James Knubel POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-96-06, GL-96-6, JPN-97-019, JPN-97-19, NUDOCS 9706030188
Download: ML20148E608 (23)

Text

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123 Main Street l

WNte Pla.ns, New York 10601 914 681.6840 914 287.3309 (FAX)

1. > NewYorkPbwer s"L"vi"i'e's,oemt.oo i

& Authority c"*' ""cee' "-

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May 27,1997 JPN-97-019 l

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U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station P1-137 Washington, DC 20555

Subject:

James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 i

l Detailed Evaluation of NRC Generic Letter 96-06: Assurance of Equipment Operability and Containment intearity Durina Desian-Basis Accident Conditions i

References:

1. NRC Generic Letter 96-06: T. T. Martin, NRC to Operating Licensees, l

" Assurance of Equipment Operability and Containment Integrity During l

Design-Basis Accident Conditions," dated September 30,1996.

2.

NYPA letter, W. J. Cahill, Jr. to NRC (JPN-97-003), " Response to NRC i

Generic Letter 96-06: Assurance of Equipment Operability and Containment l

Integrity During Design-Basis Accident Conditions," dated January 27,1997.

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Dear Sir:

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This letter provides a summary of the detailed evaluation of the postulated scenarios i

described in Generic Letter 96-06 (Reference 1), along with corrective actions, for the James A. FitzPatrick Nuclear Power Plant, in accordance with the Authority's commitment in Reference 2. The generic letter requests licensees to perform an evaluation to determine:

i (1) if containment air cooler cooling water systems are susceptible to either waterhammer or two-phase flow conditions during postulated accident conditions; and (2) if piping systems that penetrate the containment are susceptible to thermal expansion of fluid so that overpressurization of piping could occur.

As stated in Reference 2, the systems evaluated for the GL 96-06 scenarios are either not susceptible to these conditions, or if susceptible, retain operability of their safety related function during design-basis conditions. The enclosed report (Attachment 1) provides a summary of the detailed evaluation, and identifies modifications to eliminate the susceptibility of thermal overpressurization. These modifications will be implemented prior to startup from

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the next refuel outage (RFO 13).

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l 9706030188 970527 PDR ADOCK 05000333 I

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PDR l

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l A review of the FitzPatrick design code of record (ANSI B31.1 Power Piping Code -

1967 Edition), and its licensing basis in the FSAR, identified that neither of these documents provide acceptance criteria for evaluating fluid expansion effects due to accident conditions.

Accordingly, acceptance criteria were developed for this evaluation and are defined on page 5 of the enclosed report. The criteria are based on industry standards recognized by the NRC, i

and is consistent with criteria approved by the NRC for resolution of the torus attached piping issue. This acceptance criteria will be incorporated into the FitzPatrick FSAR.

i The commitments made by the Authority in this letter are listed in Attachment 2.

If you have any questions, please contact Ms. C. D. Faison.

Very tru

ours, J.Knubel Senior Vice President and Chief Nuclear Officer i

STATE OF NEW YORK COUNTY OF WESTCHESTER Subscribed and sworn to before me this47* day of d 1997.

GERALDINE STRAND f

Notary Public. State of Neur M Qualifie a

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Notary Public conmus.d s

inggM2h ion I

Enclosure:

Summary of Detailed Evaluation for NRC Generic Letter 96-06 l

Attachment:

as stated i

cc:

Regional Administrator l

U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Ms. Karen Cotton, Acting Project Manager Project Directorate I-1 l

Division of Reactor Projects-l/II U. S. Nuclear Regulatory Commission Mail Stop 14 B2 Washington, DC 20555 c.

Office of the Resident inspector i

U. S. Nuclear Regulatory Commission P.O. Box 136 Lycoming, NY 13093

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r ATTACHMENT 1 TO JPN-97-019 l

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SUMMARY

OF DETAILED EVALUATION FOR NRC GENERIC LETTER 96-06 l

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l NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT i

Docket No. 50-333 4

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Attachm:nt 1

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JPN-97-019 Page 1 of 19 INTRODUCTION l

Generic Letter 96-06 (Ref.1) " Assurance of Equipment Operability and Containment Integrity l

During Design Basis Accident Conditions," dated September 30,1996 requires licensees to determine:

1 (1) if containment air cooler cooling water systems are susceptible to either water hammer or two-phase flow conditions during postulated accident conditions-j (2) if piping systems that penetrate the containment are susceptible to thermal l

expansion of fluid so that overpressurization of piping could occur.

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On October 27,1996 the James A. FitzPatrick plant entered a scheduled refueling outage. The f

l NRC staff informed the utilities at an NEl meeting on October 29,1996 that plants starting up prior to the 120 day (January 28,1997) submittal to GL 96-06 must perform an operability l

l determination pursuant to GL 91-18 for the systems susceptible to the GL concems. In order to support this requirement, report JAF-RPT-MULTI-02596 " Operability Assecsment for NRC l

Generic Letter 96-06" was developed and issued on 12/2/96 l'

The 120 day response submittal to the NRC, dated January 27,1997, presented the l

operability findings and committed to submittal of detailed analysis results to the NRC by May l-27,1997. The purpose of this report is to summarize the findings of the detailed analyses that l

l were performed to evaluate GL 96-06 and to present the Authority's long term corrective actions.-

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DISCUSSION ITEM 1: WATER HAMMER OR TWO PHASE FLOW i

This first issue is primarily concemed with safety-related containment air coolers which are credited for heat removal from containment during DBA conditions. The Generic Letter does point out, however, that water hammer in cooling water systems associated with non-safety related containment air coolers can also challenge containment integrity by creating a l

containment bypass flow path for interfacing safety related systems.

L At FitzPatrick the containment air coolers are the Drywell Coolers (68E-1 A to 1D and 68E-3A to 3D) which are non safety related and are not credited for heat removal from containment during DBA conditions, as documented in Design Basis Document DBD-068 "Drywell Ventilation and Cooling System". Therefore, the two phase flow concem of Generic Letter 96-06, which deals with degradation of the cooler heat removal capability during design basis I

accident scenarios, is not applicable to FitzPatrick.

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JPN-97-019 Page 2 of 19 6

I The cooling water system associated with the Drywell Coolers is the non-safety related Reactor Buildmg Closed Loop Cooling (RBCLC) system (FM-15A & B). An interface does exist from the RBCLC to the safety related Emergency Service Water (ESW) system (FM-46B).

However, at FitzPatrick, the interconnecting piping circuits from ESW to RBCLC are isolated by normally closed valves (46ESW-10A&B,14A&B,17A-D, and 15 MOW 102&103). Therefore, the water hammer concem of Generic Letter 96-06, which deals with a challenge to containment integrity by creation of a bypass flow path for interfacing safety related systems, is not applicable to FitzPatrick.

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Manual operator action to feed ESW to the RBCLC system is controlled by existing plant i

procedure AOP-11, Rev. 8 " Loss of Reactor Building Closed Loop Cooling". Within this procedure several cautions and instruct 6ons are given to the, operator which are directly related to the " bypass flow path" concem. The first CAUTION states: " Cross-tying ESW loops to RBCLC could divert ESW from cafety related loads". The operator is subsequently instructed

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to monitor crescent area temperatures to ensure adequate cooling water is being supplied to the safety related coolers. The next CAUTION specifically states: " Supplying ESW to the i

l Drywell Coolers could divert ESW from safety related loads". Here again the operator is instructed to monitor crescent area temperatures. Also, the operator is instructed that if the Drywell floor drain leak rate rises or cannot be determined he should isolate ESW from the RBCLC system. This is accomplished by remote / manual operation of RBCLC containment j

isolation valves (15AOV-130A&B,131 A&B,132A&B and 133A&B) from the control room.

There is the potential that a water hammer may occur if the operator supplies ESW to the RBCLC during postulated accident conditions. A procedure change was made to add a i

caution to AOP-11 concoming the generation of a water hammer in the RBCLC system. This I

completes an NRC commitment (JPN-97-003-01).

Additionally, NRC Information Notice 96-60 " Potential Common Mode Post-Accident Failure of l

RHR Heat Exchangers" dated November 14,1996, has been linked to Generic Letter 96-06 by the NRC. FitzPatrick's use of an RHR Keepfull System precludes the concems of this NRC l

Information Notice.

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Based on the above, it is concluded that the waier hammer / two phase flow concems of the l

NRC Generic Letter 96-06 are not applicable to FitzPatrick.

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4 ITEM 2: THERMAL PRESSURIZATION l

GENERAL Generic Letter 99-06 states " Thermally induced overpressurization of isolated water filled piping sections in containment could jeopardize the ability of accident mitigating systems to perform their safety functions and could also lead to a breach of containment integrity via i

bypass leakage."

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Att:chmint 1 JPN-97-019 Page 3 of 19 l

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The industry events presented in the generic letter illustrate that the concem not only deals I

with piping system pressure boundary integrity but also the ability of the isolation valves to allow the systems to perform their safety function. In order to analyze this issue the following evaluation process was used" a thorough review of all Drywell and Suppression Pool penetrations was made to identify those which involved liquid systems, valve arrangements for the above systems were reviewed to determine if isolated I

water filled sections exist and if thermal pressurization was possible, a heat transfer model was developed to determine the effect of the post accident ambient area temperatures on the isolated water temperature for selected penetrations from which trapped fluid pressure was calculated, l

valve analysis was performed on the suscepti;ie penetration isolation valves to determine their maximum pressure retaining capability.

the limiting intemal pressure of the isolated pipe section was then determined, containment integrity and safety system function were then evaluated for the DBA thermal pressurization condition.

SCREENING RESULTS A total of 584 containment penetrations were reviewed, of which 314 contained isolated water sections (Control Rod Drive system accounts for 274 of these penetrations).

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Based on the screening process,14 penetrations had isolated water filled sections /potentially susceptible to thermal pressurization.

Two (2) penetrations (X-14 and X-41) were eliminated from further thermal pressurization and containment integrity evaluation since the water temperature during normal operating conditions would be the same or higher than the water temperature under DBA conditions.

One (1) penetration (X-12) was eliminated from thermal pressurization and containment integrity evaluation due to installed valve design features (e.g. bonnet bypass) which limited the pipe line pressunzation to below design pressure.

1 The remaining eleven (11) penetrations (X-8, X-18, X-19, X-39A, X-398, X-210A, X-2108, X-211 A, X-211B, X-224, and X-226) were analyzed at the limiting i

calculated pressure for each penetration.

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JPN-97-019 Page 4 of 19 i

l SYSTEM SAFETY FUNCTION REVIEW GENERAL The generic letter presented an industry experience for Beaver Valley Units 1 and 2 in which a i

motor operated valve associated with an isolated section of thermally pressurized piping would not open during a curveillance test. Since isolated piping sections which penetrate containment will be exposed to thermal pressurization the potential for systems to fail to perform their safety functions needed to be evaluated.

In order to disposition this issue the valves associated with the penetrations identified above were reviewed for post accident operating requirements. The results of this evaluation are nrasented below.

DISCUSSION i

Penetrations X-8, X-18, X-19 and X-224 (systems associ::ted with the penetrations are described later in the Results Section) do not have safety functions. The associated CIVs are either normally closed or receive a signal to close for DBA conditions and remain closed for the duration of the event.

Penetrations X-39 A&B, X-210 A&B and X-211 A&B employ inboard globe valves with isolated fluid ursder the seats. An increase in isoleted fluid pressure due to thermal conditions would assist valve opening. Therefore, valve operability and consequently system safety function would not be jeopardized.

The valvas associated with penetration X-226 (23MOV-57 & 58) provide a transfer function for the HPCi pump suction source from the Condensate Storage Tank (CST) to the Torus. HPCI pump suction is normally lined up to the CST as the preferential source. Potential thermal pressurization of this containment penetration configuration is due to elevated Torus temperature. A review of the plant transient and accident analyses indicates transfer function is not required except for either initiating events which would render the CSTs unavailable or i

non mechanistic failure of non safety related sections of the system. No accident sequences have been identified which would require opening of these valves coincident with elevated Torus temperature. Break size, rate of Reactor depressurization, quantity of outflow from the CST, and rate of Torus temperature rise are all factors whch affect this evaluation. Since l

. HPCI has a wide range of operating scenarios, all of these factors can very significantly. The available data cufirms that the HPCI systam safety function will not be jeopardized for design basis accident conditions. To ensure availability for the transfer function for the HPCI suction l

Une, NYPA plans to eliminate the susceptibility of this penetration to thermal pressurization by completion of a modification.

Although penetration X-224, which provides a transfer function for the RCIC pump suction source from the CST to the Torus, does not have a safety related system function, it is also h

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Attachm2nt 1 JPN-97-019 Page 5 of 19 I

susceptible to thermal pressurization during operating transients due to potential elevation of Torus temperature. To ensure availability for the tranc.fer function for the RCIC suction line, NYPA plans to eliminate the susceptibility of this penetration to thermal pressurization by 4

completion of a modification.

i CONTAINMENT INTEGRITY REVIEW GENERAL During the initial stages of this evaluation, it was recognized that penetrations X-8, X-12, X-18, X-39A, X-398, X-210A, X-2108, X-211 A, and X-211B would have their pressure conditions i

limited by valve design features. Such features included: a) double disk gate valve designs l

with either bonnet bypass or a drilled disk, and b) globe valve designs with isolated fluid under the seat. For these penetrations, the Authority performed calculations (Ref. 4 and 5) to -

determine the limiting isolated fluid pressure that the penetration would experience due to thermal pressurization.

Stone & Webster Engineering Corp. (SWEC) performed detailed analyses for penetratinns X-i=

19, X-224, and X-226 (Ref. 6,7, and 8) using a zero valve leakage assumption since the initial screening did not reveal a valve design feature which would limit the isolated fluid pressure.

The analyses included: a) a heat transfer model to evaluate the effects of DBA condition area temperatures on the isolated fluid temperatura, and b) a pressure calculation to determine the resultant pressure of the isolated fluid due to thermal pressurization.

SWEC then performed stress evaluations (Ref 9 through 16) to determine the acceptability of the stresses in the penetration section of piping and its associated components due to the thermal pressurization. The analyses performed by SWEC considered the containment isolation valve (CIV) bodies, however, it did not evaluate the valve intemals.

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Altran Corporation was contracted to evaluate the integrity' of the intemals of the CIVs, due to their extensive involvement in the GL 89-10 program. The results of their evaluation (Ref. 3) provided limiting pressure data for penetrations X-19, X-224, and X-226. The Altran analysis showed that these three penetrations would develop a leak path at the CIVs due to thermal pressurization. The stress evaluation for these three penetrations was subsequently re-evaluated by SWEC based on the Altran pressure results.

A.ECEPTANCE CRITERIA As part of the GL 96-06 evaluation, a review of the FitzPatrick's design code of record, ANSI B31.1 Power Piping Code 1967 Edition, and its Licensing basis was performed. Neither of these two documents provide acceptance criteria for evaluation of fluid expansion effects due to accident conditions. Therefore, stress-based plant specific acceptance criteria using elastic

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Attachmint 1 JPN-97-019 i

Page 6 of 19 l

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methodology were developed for this evaluation. The criteria developed is consistent with that l

approved by the NRC for resolution of the torus attached piping issue (References 21,22,23).

Maximum allowable intamal pressures are calculated in accordance with the ANSI B31.1 Code equations (Or minimum wall thickness and general pipe stress (which includes a longitudinal pressure term) using criteria allowed by ASME Section lil, Appendix F and IP-PS-04, Pipe Support Analysis Procedure. Maximum allowable intamal pressures for integrally welded pipe attachment (lWA) pipe supports are calculated using the appropriate load combinations defined in Ref. 20 (Table 4.8.1-1). These " maximum allowable intamal pressures" for each penetration are then compared to the maximum attainable pressure caused by thermal pressurization between the isolation valves. The integrity of the piping is assured when J

maximum allowable intemal pressure values envelope the thermal pressurization levels. The maximum allowable pressures are calculated as shown below.

l Minimum Wall: (For Evaluation of Pioing) 2 Sh (tm -A)

P=2 l

Do - 2 y (t. -A)

Normal / Upset Primary Stress: (For Evaluation of Piping & IWA Pipe Supports)

Sr ~ __. - + Swe

+ SRSS(Soa, Som)5 2.4Sh j

" Do - d'

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2 P = (2.4xSh - SToTR UPSET MY) X

+ PoeSion d'

Faulted Primary Stress: (For Evaluation of Piping & IWA Pipe Supports) 1 Sr -

_ + Se f _ % + SRSS(So ey, Sow)$ 2.4Sh 2

Do - d P = (2.4xSh - Stora rareo pamy) X

+ PDESIGN d'

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JPN-97-019 Page 7 of 19 I

Where:

P - maximum allowed intemal pipe pressure (psi)

Sh - allowable pipe stress at temperature (psi) l

t. - minimum wall thickness, including manufacturers allowed tolerance (inches)

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A - additional thickness to compensate for material removed in threading, etc. of pipe and to provide for mechanical strength (inches) t Do - outside diameter of pipe (inches) d -inside diameter of pipe (inches) y - a coefficient to account for pipe material and temperature SRSS - Square Root of the Sum of the Squares l

Sr

- Pipe stress due to the deadweight (the existing value will not change due to l

thermal overpressurization).

Su ; -_ p,

- Pipe stress due to intemal pressure computed in accordance with the BOP Piping Stress and Supports Design Criteria (Ref. 20).

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So er and Soner - Pipe stress due to operating basis and design basis earthquakes respectively (the existing value will not change due to thermal overpressurization).

j Sw - Pipe stress due to hydrodynamic events such as SRV discharge, chugging, etc. are short duration events. Loads from these events will dissipate long before the Intemal pressure can increase significantly. Therefore, they need not be included in the plant operability pipe stress checks for thermal pressurization concems.

i PDeStGN - Existing design pressure of pipe RESULTS l

The following is a summary of the stress evaluations results from Ref. 2 for the eleven (11) l

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penetrations determined to be susceptible to thermal pressurization. The su~mmaries only show the pressures associated with the limiting stress component.

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- Attachment 1 JPN-97-019 Page 8 of 19 Penetration X-8 The following components are subjected to thermal overpressurization and are addressed in Reference 9:

Pipe Line :

3"-SHP-902-3A, 3"-SHP-902-6 Pipe Supports:

None

' Penetration:

X-8 System:

Main Steam Line Drain

. An evaluation of the piping and the penetration listed above determined that the maximum intemal pressure for which the acceptance criteria defined on page 5-7 of this report can be met is 3,546 psi.

Calc alation JAF-CALC-MULTI-02591 (Ref. 4) calculates, for containment isolation valve 29-MO'/-74, a differential pressure across the valve seat which will lift the valve, allowing the intomal pressure to be relieved. The table below shows the differential pressure to lift the valve, the design pressure of the outboard pipe (relative to the isolated section) and total pressure. The design pressure is added to the differential pressure to bound a spectrum of DBA analysis cases.

Differential Pipe Outboard of Design Total Valve Pressure Valve, Relative to Pressure Pressure (psi)

Isolated Section (psi)

(psi) 29MOV-74 2091 3"-SHP-902-3A

-1146 3237 The maximum intemal pressure due to thermal overpressurization (3,237 psi) is within the maximum allowable intemal pressure (3,546 psi). Therefore, the acceptance criteria has been met for the affected piping and piping components associated with Drywell Penetration X-8.

Penetration X-18 The following components are subjected to thermal overpressurization and are addressed in.

Reference 19:

Pipe Line :

3"-WL-151-2 and %" Vent Pipe Supports:

None Penetration:

X-18 and S-263 System:

Drywell Floor Drain Sump Discharge l

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Att: chm:nt 1 JPN-97-019 Page 9 of 19 l

An evaluation of the piping and the penetrations listed above determined that the maximum intemal pressure for which the acceptance criteria defined on page 5-7 of this report can be i

met is 2,188 psi.

l Calculation JAF-CALC-MULTI-02591 (Ref. 4) calculates, for containment isolation valve l

20MOV-82, a differential pressure which will cause the valve to leak, allowing the intamal l

pressure to be relieved. The table below shows the differential pressure, the design pressure i

I of the outboard (relative to the isolated section) pipe and total pressure.

Valve Differential Pipe Outboard of Design Total l

l Pressure (psi)

Valve, Relative to Pressure Pressure l

Isolated Section (psi)

(psi) 20MOV-82 264 3"-WL-151-1 A 150 414 l

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l The maximum intamal pressure due to thermal overpressurization (414 psi) is within the maximum allowable intemal pressure (2,188 psi). Therefore, the acceptance criteria has been met for the affected piping and piping components associated with Drywell Penetration X-18.

Penetration X-19 The following compments are subjected to thermal overpressurization and are addressed in i

Reference 10:

y l

Pipe Line :

3"-WH-151-7 and %" Vent Pipe Supports:

None Penetration:

X-19 and S-258 System:

Drywell Equipment Drain Sump Discharge I

An evaluation of the piping and the penetrations listed above determined that the maximum intamal pressure for which the acceptance criteria defined on page 5-7 of this report can be L

met is 2,188 psi.

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' Altran calculation 96254-TR-01 (Ref. 3) states, for containment isolation valve 20AOV-95 (the outboard ball valve), a differential pressure which will cause the valve to leak, allowing the intomal pressure to be relieved. The following table shows the differential pressure, the design pressure of the outooard (relative to the isolated section) pipe and total pressure.

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Valve Differential Pipe Outboard of Design Total Pressure (psi)

Valve, Relative to Pressure Pressure l

Isolated Secten (psi)

(psi) i i

20AOV-95 1467 3"-WH-151-7 150 1617 l

The maximum intamal pressure due to thermal overpressurization (1617 psi) is within the maximum allowable intemal pressure (2,188 psi). Therefore, the acceptance criteria has been l

met for the affected piping and piping components associated with Drywell Penetration X-19.

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The leak path for 20AOV-95 is considered to be a non-resealing leak path due to the seat l

materials used in the valve. Therefore, it is considered prudent to provide a long term l

corrective action for this penetration. A hole will be drilled in the containment side disk of the inboard Anchor Darling valve 20MOV-94 to provide a relief path. This modification will effectively reduce the thermal pressurization to approximately 265 psid via a resealing leak path. At this value, the leak path past the inboard valve would maintain the sealing integrity of the outboard ball valve.

Penetration X-39A l

The following components are subjected to thermal overpressurization and are addressed in Ref.17:

Pipe Line :

10"-W20-307-12A Pipe Supports:

H10-397, PFSK-1928, and PFSK-1917 Branch:

2"-AS-302-55A,1%"-AS-302-55A, and 1"-AS-302-55A Penetrations:

None System:

RHR Containment Spray An evaluation of the piping and the pipe supports listed above determined that the maximum intemal pressure for which the acceptance criteria defined on page 5-7 of this report can be l

met is 1,684 psi.

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Calculation JAF-CALC-RHR-02589 (Ref. 5) calculates, for containment isolation valve 10MOV-31 A, a differential pressure across the valve seat which will lift the valves, allowing the intemal pressure to be relieved. The table below shows the Gferent!al pressure to lift the valye, the design pressure of the outboard (relative to the isolated section) pipe and total pressure.

I Valve Differential Pipe Outboard of Design Total Pressure (psi)

Valve, Relative to Pressure Pressure isolated Section (psi)

(psi) i 10MOV-31A 703 10"-W20-302-12A 325 1028 i

l The maximum intemal pressure due to thermal overpressurization (1028 psi) is within the maximum allowable intemal pressure (1,684 psi). Therefore, the acceptance criteria has been met for the affected piping and piping components associated with Drywell Penetration X-39A.

Penetration X-39B l

The following components are subjected to thermal overpressurization and are addressed in Ref.18:

Pipe Line :

10"-W20-302-12B Pipe Supports:

H10-521 and PFSK-2393 l

Branch:

1%"-AS-302-558, and 2"-AS-302-55B l

Penetrations:

None l

System:

RHR Containment Spray l

An evaluation of the piping and the pipe supports listed above determined that the maximum intamal pressure for which the acceptance criteria defined on page 5-7 of this report can be met is 1,684 psi.

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~ Calculation JAF-CALC-RHR-02589 (Ref. 5) calculates, for containment isolation valve 10MOV-318, a differential pressure across the valve seat which will lift the valves, allowing the intemal pressure to be relieved. The table below shows the differential pressure to lift the valve, the design pressure of the outboard (relative to the isolated section) pipe and total pressure.

Valve Differential Pipe Outboard of Design Total 1

Pressure (psi) -

Valve, Relative to

. Pressure Pressure Isolated Section (psi)

(psi)

I 10MOV-318 703-10"-W20-302-12B 325 1028

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The maximum intemal pressure due to thermal overpressurization (1,028 psi) is within the i

maximum allowable intomal pressure (1684 psi). Therefore, the acceptance criteria has been i'

met for the affected piping and piping components associated with Drywell Penetration X-39B.

i Penetration X-210A s!

The following components are subjected to thermal overpressurization and are addressed in Reference 11:

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Pipe Line :

16"-W20-302-15A

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Pipe Supports:

PFSK-1940, PFSK-1944 & PFSK-1984 i

Penetration:

None l

System:

RHR to Suppression Pool An evaluation of the piping and the pipe supports listed above determined that the maximum infomal pressure for which the acceptance criteria defined on page 5-7 of this report can be met is 1,565 psi.

NYPA calculation JAF-CALC-RHR-02589 (Ref. 5) calculates, for isolation valves 10MOV-34A l

and 10MOV-38A, a differential pressure across the valve seat which will lift the valves, allowing i

the intamal pressure to be relieved. The table below shows the differential pressure to lift the j

valve, the design pressure of the outboard pipe (relative to the isolated section) and total j

pressure. The design pressure is added to the differential pressure to bound potential

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operating pressures.

Differential Pipe Outboard of Design Total L

Valve Pressure Valve, Relative to Pressure Pressure (psi)

Isolated Section (psi)

(psi) 10MOV-34A -

943 16"-W20-152-5A 150' 1093 10MOV-38A 1132 6"-W20-152-44A 150 1282 Maximum Total Pressure (limited by 10MOV-34A) 1093

- The maximum intamal pressure due to thermal overpressurization (1,093 psi) is within the maximum allowable intemal pressure (1,565 psi). Therefore, the acceptance criteria has been met for the affected piping and piping components associated with Drywell Penetration X-210A.

Att* chm:nt 1

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JPN-97-019 Page 13 of 19 1

i Penetration X-2108 The following components are subjected to thermal overpressurization and are addressed in Reference 12:

Pipe Line :

16"-W20-302-15B Pipe Supports:

PFSK-2042, PFSK-2477 Penetration:

None i

System:

RHR to Suppression Pool An evaluation of the piping and the pipe supports listed above determined that the maximum intemal pressure for which the acceptance criteria defined on page 5-7 of this report can be met is 1,565 psi.

NYPA calculation JAF-CALC-RHR-02589 (Ref. 5) calculates, for isolation valves 10MOV-343 and 10MOV-388, a differential pressure across the valve seat which will lift the valves, allowing the intamal pressure to be relieved. The table below shows the differential pressure to lift the valve, the design pressure of the outboard (relative to the isolated section) pipe and total pressure. The design pressure is added to the differential pressure to bound potcatial operating pressures.

i Differential Pipe Outboard of Design Total Valve Pressure Valve, Relative to Pressure Pressure j

(psi)

Isolated Section (psi)

(psi) 10MOV-34B 943 16"-W20-152-5B 150 1093 10MOV-38B 1361 6"-W20-152-44B 150 1511 Maximum Total Pressure (limited by 10MOV-34B) 1093

)

The maximum intemal pressure due to thermal overpressurization (1,093 psi) is within the maximum allowable intemal pressure (1,565 psi). Therefore, the acceptance criteria has been met for the affected piping and piping components associated with Drywell Penetration X-210B.

Penetration X-211 A The following components are subjected to thermal overpressurization and are addressed in Reference 13:

Pipe Line :

6"-W20-302-16A Pipe Supports:

None Penetration:

None p

Att: chm:nt 1

~

JPN-97-019 Page 14 of 19 System:

RHR Torus Spray 1

l An evaluation of the piping listed above determined that the maximum intemal pressure for which the acceptance criteria defined on page 5-7 of this report can be met is 1,889 psi.

NYPA calculation JAF-CALC-RHR-02589 (Ref. 5) calculates, for isolation valves 10MOV-34A and 10MOV-38A, a differential pressure across the valve seat which will lift the valves, allowing l

the intemal pressure to be relieved. The table below shows the differential pressure to lift the valve, the design pressure of the outboard pipe (relative to the isolated section) and total pressure. The design pressure is added to the differential pressure to bound potential l

operating pressures.

I Differential Pipe Outboard of Design Total Valve Pressure Valve, Relative to Pressure (p Pressure l

(psi)

Isolated Section si)

(psi) l 10MOV-34A 943 16"-W20-152-5A 150 1093 10MOV-38A 1132 6"-W20-152-44A 150 1282 Maximum Total Pressure (limited by 10MOV-34A) 1093 The maximum intemal pressure due to thermal overpressurization (1,093 psi) is within the l

maximum allowable intemal pressure (1,889 psi). Therefore, the acceptance criteria has been l

met for the affected piping and piping components associated with Drywell Penetration X-211A.

Penetration X-211B The following components are subjected to thermal overpressurization and are addressed in l

Reference 14:

Pipe Line :

6"-W20-302-16B l

Pipe Supports:

None l

Penetration:

None System:

RHR Torus Spray 1

a

I Attachm:nt 1 l

JPN-97-019 l

Page 15 of 19 l

An evaluation of the piping listed above, determined that the maximum intemal pressure for l

l which the acceptance criteria defined on page 5-7 of this report can be met is 1,889 psi.

l l

NYPA calculation JAF-CALC-RHR-02589 (Ref. 5) calculates, for isolation valves 10MOV-348 and 10MOV-36B, a differential pressure across the valve seat which will lift the valves, allowing the intemal pressure to be relieved. The table below shows the differential pressure to lift the 1

valve, the design pressure of the outboard (relative to the isolated section) pipe and total pressure. The design pressure is added to the differential pressure to bound potential operating pressures.

1 Differential Pipe Outboard of Design Total l

Valve Pressure Valve, Relative to Pressure Pressure (psi)

Isolated Section (psi)

(psi) j 10MOV-34B 943 16"-W20-152-5B 150 1093 l

10MOV-38B 1361 6"-W20-152-448 150 1511 Maximum Total Pressure (limited by 10MOV-348) 1093 The maximum intemal pressure due to thermal overpressurization (1,093 psi) is within the l

maximum allowable intemal pressure (1,889 psi). Therefore, the acceptance criteria has been i

met for the affected piping and piping components associated with Drywell Penetration X-211B.

l Penetration X-224 The following components are subjected to thermal overpressurization and are addressed in Reference 15:

l Pipe Line :

6"-W22-152-16 Pipe Supports:

H13-1, H13-2, H13-3 & H13-48 Penetration:

None System:

RCIC Torus Suction An evaluation of the piping and the piping supports listed above determined that the maximum l

intemal pressure for which the acceptance criteria defined on page 5 of this report can be met l

is 1,789 psi.

f Altran calculation 96254-TR-01 (Ref. 3) states, for isolation valves 13MOV-41 and 13MOV-39, that the valves willleak at 1,026 psi, allowing the intemal pressure to be relieved. For these valves the leak path is at the body to bonnet flange. Therefore, the downstream pressure was taken to be 14.7 psi.

I

JPN-97-019 Page 16 of 19 The maximum intemal pressure due to thermal overpressurization (1026 psi + 14.7 psi = 1041

- psi) is within the maximum allowable intomal pressure (1,789 psi). Therefore, the acceptance criteria has been met for the affected piping and piping components associated with Torus Penetration X-224.

Penetration X-226 The following components are subjected to thermal overpressurization and are addressed in Reference 16:

Pipe Une :

16"-W25-152-17 Pipe Supports:

PFSK-983 & PFSK-2248 Penetration:

None System:

HPCI Torus Suction An evaluation of the piping and the pipe supports listed above determined that the maximum intemal pressure for which the acceptance criteria defined on page 5-7 of this report can be met is 1,251 psi.

Altran caiculation 96254-TR-01 (Ref. 3) states, for isolation valves 23MOV-58 and 23MOV-57, that the valves will leak at 1,080 psi, allowing the intamal pressure to be relieved. For these valves the leak path is at the body to bonnet flange. Therefore, the downstream pressure was

- taken to be 14.7 psi.

The maximum intemal pressure due to thermal overpressurization (1080 psi + 14.7 psi = 1095 psi) is within the maximum allowable intamal pressure (1,251 psi). Therefore, the acceptance criteria has been met for the affected piping and piping components associated with Torus

. Penetration X-226.

~ CONCLUSIONS 1)

For penetrations susceptible to thermal pressurization, it has been shown that the maximum trapped fluid pressure was limited by valve relief features / mechanisms.

2)

Based on the acceptance criteria contained herein, FitzPatrick systems are not susceptible to the safety concems of GL 96-06.

~

AttachmInt 1 JPN-97-019 Page 17 of 19 I

CORRECTIVE ACTIONS The following corrective actions require implementation prior to start up from the cycle RO,

refueling outage:

1)

The containment side disk of valve 20MOV-94 will be drilled to provide a relief path for thermal pressurization of penetration X-19.

2)

A modification to penetrations X-224 and X-226 will be completed to eliminate the susceptibility of these penetrations to thermal pressurization.

The acceptance criteria contained in this report which was developed for evaluation of the

)

thermal pressurization analysis will be incorporated into the FSAR.

REFERENCES 1)

NRC Generic Letter 96-06 " Assurance of Equipment Operability and Containment integrity During Design Basis Accident Conditions" dated September 30,1996 2)

SWEC report J.O. 02268.5068 "Results of Containment Penetration Screening Analysis and Containment Integrity Evaluation for Response to NRC Generic Latter 96-06" dated 3/21/97 3)

Altran report 96254-TR-01 Rev.1 " Qualification of MOV's Subjected to increased Line Pressure Due to Isolated Fluid Expansion from LOCA Temperatures" dated January 1997 4)

NYPA Calculation JAF-CALC-MULTI-02591 Rev. O " Maximum Pressure for Thermal Pressurization of Penetrations X-8, X-12, and X-18 (GL 96-06)" dated 12/6/96 5)

NYPA Calculatic,n JAF-CALC-RHR-02589 Rev. O " Maximum Pressure for Thermal Pressurization of Penetrations X-39A,B; X-210A,B; and X-211 A,B (GL 96-06)" dated 12/9/96 6)

SWEC Calculation JAF-CALC-RCIC-02015 Rev.1 " Temperature and Pressure Transient Analysis of Trapped Water Volume for Containment Penetration X-224" dated 1/15/97 7).

SWEC Calculation JAF-CALC-HPCI-02616 Rev.1 " Temperature and Pressure Transient Analysis of Trapped Water Volume for Containment Penetration X-226" dated 1/15/97 t

1-

Att: chm:nt 1 JPN-97-019 Page 18 of 19 8)

SWEC Calculation JAF-CALC-RADW-02617 Rev.1 " Temperature and Pressure Analysis of Trapped Water Volume for Containment Penetration X-19" dated 1/15/97 9)

SWEC Calculation JAF-CALC-MST-02618 Rev.1 " Generic Letter 96-06 Operability Assessment for Affected Piping and Piping Components Between Containment Isolation Valve 29MOV-74 and 29MOV-77 (Drywell Penetration X-8)" dated 1/15/97 10)

SWEC Calculation JAF-CALC-RADW-02619 Rev. 3 " Generic Letter 96-06 Operability Assessment for Affected Piping and Piping Components Between Containment Isolation Valve 20MOV-94 and 20MOV-95 (Drywell Penetration X-19)" dated 2/28/97 11)

SWEC Calculation JAF-CALC-RHR-02620 Rev.1 " Generic Letter 96-06 Operability Assessment for Affected Piping and Piping Components Between Containment isolation Valve 10MOV-34A and 10MOV-39A (Drywell Penetration X-210A)" dated 1/15/97 12)

SWEC Calculation JAF-CALC-RHR-02621 Rev.1 " Generic Letter 96-06 Operability Assessment for Affected Piping and Piping Components Between Containment isolation Valve 10MOV-34B and 10MOV-39B (Drywell Penetration X-2108)" dated 1/15/97 13)

SWEC Calculation JAF-CALC-RHR-02622 Rev.1 " Generic Letter 96-06 Operability Assessment for Affected Piping and Piping Components Between Containment isolation Valve 10MOV-38A and Line 16"-W20-302-15A (Torus Penetration X-211A)"

dated 1/15/97 14)

SWEC Calculation JAF-CALC-RHR-02623 Rev.1 " Generic Letter 96-06 Operability Assessment for Affected Piping and Piping Components Between Containment isolation Valve 10MOV-388 and Line 16"-W20-302-15B (Torus Penetration X-211B)"

dated 1/15/97 15)

SWEC Calculation JAF-CALC-RCIC-02624 Rev.1 " Generic Letter 96-06 Operability Assessment for Affected Piping and Piping Components Between Containment isolation Valve 13MOV-41 and 13MOV-39 (Torus Penetration X-224)" dated 2/21/97 16)

SWEC Calculation JAF-CALC-HPCI-02625 Rev.1 " Generic Letter 96-06 Operability Assessment for Affected Piping and Piping Components Between Containment isolation Valve 23MOV-58 and 23MOV-57 (Torus Penetration X-226)" dated 2/21/97 17)

SWEC Calculation JAF-CALC-RHR-02663, Rev.1, "G_eneric Letter 96-06 Operability Assessment for Affected Piping and Piping Components Between Containment isolation Valves 10MOV-31 A and 10MOV-26A (Drywell Penetration X-39A)" dated 2/25/97.

1 l

JPN-97-019 Page 19 of 19 4

t h.

18)

SWEC Calculation JAF-CALC-RHR-02664, Rev.1, " Generic Letter 96-06 Operability i

Assessment for Affected Piping and Piping Components Between Containment isolation Valves 10MOV-318 and 10MOV-268 (Drywell Penetration X-398)" dated 2/25/97.

19)

SWEC Calculation JAF-CALC-RADW-02670, Rev.1, " Generic i.etter 96-06 Operability Assessment for Affected Piping and Piping Components Between Containment Isolation Valves 20MOV-82 and 20AOV-83 (Drywell Penetration X-18)" dated 2/27/97.

1 20)

Design Criteria for Balance of Plant (BOP) Piping Stress and Supports, Rev. O, dated April 2,1991.

i 21)

General Electric Report NEDO-24583-1, " Mark 1 Containment Program Structural Acceptance Criteria Plant Unique Analysis application Guide" dated October 1979.

22)

NRC " Safety Evaluation Report, Mark 1 Containment Long-Term Program," NUREG-0661 dated July 1980.

23)

Telodyne Engineering Service Report, TR-5321-2, " Plant-Unique Analysis Report of the Torus Attached Piping for James A. FitzPatrick Nuclear Power Station" dated November 1984.

1 h

4 ATTACHMENT 2 TO JPN-97-019 James A. FitzPatrick Nuclear Power Plant List of Commitments Number Commitment Due Date JPN-97-019-01 Modify Drywell Equipment Drain Sump Discharge Prior to Start-Up Valve 20MOV-94 to provide a pressure relief path.

from RFO-13.

JPN-97-019-02 incorporate the acceptance criteria for evaluation First Updated of thermal overpressurization during accident FSAR submittal conditions into the FSAR.

following the 1997 UFSAR submittal.

JPN-97-019-03 Modify penetrations X-224 and X-226 to eliminate Prior to Start-Up susceptability to thermal pressurization.

from RFO-13.

--