ML20078A782

Purge & Vent Containment Isolation Valves Resilient Seal Maintenance/Surveillance Program & Technical Basis

ML20078A782
Person / Time
Site:	Oyster Creek
Issue date:	06/24/1994
From:	GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20078A754	List: ML20078A752 ML20078A782
References
PROC-940624, NUDOCS 9407010279
Download: ML20078A782 (29)

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ATTACHMENT I OYSTER CREEK PURGE AND VENT CONTAINMENT ISOLATION VALVES RESILIENT SEAL MAINTENANCE / SURVEILLANCE PROGRAM AND TECHNICAL BASIS l

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TABLE OF CONTENTS

1.0 INTRODUCTION

2.0 MAINTENANCE / SURVEILLANCE PROGRAM 3.0 SUPPORTING TECHNICAL BASIS 3.1 Centerline Butterfly Valves V-27-1/2/3/4 3.2 Fisher Controls Butterfly Valves V-23-13/14/15/16 3.3 Fisher Controls Butterfly Valves V-28-17/18 4.0 TABLES

5.0 REFERENCES

6.0 APPENDICES Appendix A - Manufacturer's Seat Hardness Data Appendix B - Oyster Creek Test Hardness Data for New Warehoused Seats 2

1,0 INTRQDUCTION The NRC identified the need to replace, at least every 5 years, the resilient seals (seats) on Oyster Creek's purge and vent containment isolation valves. GPU Nuclear has conducted a technical review of these butterfly valves' resilient seals (seats) to determine their expected service life. Table 1 (Section 4.0 of this Attachment) summarizes the subject valves and seat type. This review determined that, for the specific anticipated seat / valve environment, EPDM seats have an expected service life of 40 years and the Nitrile and Viton seats have an expected life of 20 years. Based upon these findings of expected long life and the fact that when valve seats are replaced the seal of the new seat may not be as leaktight as the previous seat due to difficulties in valve seat alignment and the need for a break-in" period, GPU Nuclear has concluded that replacement of the subject valve seats should be based on their physical condition as monitored by a maintenance / surveillance program rather than a fixed,5 year, seat replacement schedule. This proposed program would be in addition to the existing seat leakage testing performed during each refueling outage. GPU Nuclear intends to implement the maintenance / surveillance program during the upcoming 15R Refueling Outage.

Section 2.0 of this Attachment provides the details of the maintenance / surveillance program and Section 3.0 provides a summary of the technical basis supporting the expected service life for the purge and vent valve seats.

2.0 MAINTENANCE /SilRVEILLANCE PROGRAM The maintenance / surveillance program for Oyster Creek's purge and vent containment isolation valves seats is intended to commence in the upcoming 15R Refueling Outage scheduled for September,1994. During the 15R and 16R Outages one of the valves with an EPDM seat, one with a Viton seat and one with a Nitrile seat will be inspected. These two outage inspections will establish a trend for the as-found condition of the seats for each type of material. The maintenance / surveillance program will consist of inspecting the resilient seats with a portable hardness tester and making a visual inspection for hairline cracks. A sign of cracks or an increase shift in the durometer hardness average readings (disregarding readings near the disc shaft where the seat is re-enforced) in the range of 70 to 78 Shore A Durometer reading for the EPDM seat, or an increase shift in the range of 65 to 75 Shore A Durorneter reading for the Viton seat, or an  ;

increase shift in the range of 70 to 80 Shore A Durometer reading for the Nitrile seat would indicate j that material degradation is occurring and that the seats should be removed (at the existing or next )

outage of sufficient duration) for laboratory analysli The frequency of these inspections after the 16R Outage will be based upon trending of the as-found condition of the seats.

3.0 SUPPORTING TECHNICAL BASIS All of the subject valves, listed in Taole 1, are located in the Reactor Building, outside the Drywell.

They are required to mitigate all containment events. Valves V-27-1 through 4, V-23-13 through 16 and V-28-17 and 18 are normally closed They are only operated in the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period preceding a shutdown, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following a startup, when a leak test is performed and for drywell entries. They may also be operated in accordance with procedure following accident conditions.

All valves are also full stroke exercised for operability every 3 months.

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l An analysis of the environmental degradation on the resilient seat for each valve was performed to ensure that the valves will perform their intended functions during their normal service, accident and 48 day post-accident life. For conservatism, the valves were assumed to be exposed to the drywell level gamma radiation and not the Reactor Building gamma to which they are actually exposed.

Those valves are closed during accident conditions. Therefore, there is no flow path for Beta radiation. Only the seat edge facing the drywell atmosphere (of the closest valve) will be exposed to Beta radiation. Deterioration of this seat edge is of no concern as it will not extend into the seating surface. Therefore, the effects of LOCA Beta radiation r,eed not be addressed in this analysis.

Opening of the valves to vent the drywell and/or torus only occur during "beyond design basis" conditions. UE&C Calculation #7450.116 (GPUN DRF 069398) documents a 1x10' Rad /hr Beta release. GPU Nuclear calculation (C1302-243-5450-062) has determined that valves V-23-13/14 and V-23-15/16 will be opened a maximum of six times per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for a loss-of-air event to decrease the drywell pressure at 55 psig. These valves can be open for 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> before they are exposed to the Viton hardness threshold of 5x10* Rads or 6000 hours0.0694 days <br />1.667 hours <br />0.00992 weeks <br />0.00228 months <br /> before reaching the 25% hardness degradation as noted in paragraph 3.2, Radiation Qualification.

EPRI NP-3877 Reference 1 describes the methodology used for qualifying active mechanical equipment for safety related service in nuclear power plants. Mechanical equipment has been traditionally designed for replacement of parts that are subject to aging. The parts are usually those that are most subject to wear and to environmental degradation. The following sections provide the technical basis which demonstrates a long predicted seal service life (20 year minimum) and details GPU Nuclear's proposed maintenance / surveillance program for each resilient seal material type.

3.1 CENTERLINE BUTTERFLY VALVES V-27-1/2/3/4 The Centerline 18" butterfly valves EPDM seats (SSN 721-312-3300-1) properly stored, with a durometer hardness in the range of 70 to 78 Shore A Durometer Reading (see Appendix A), and visually inspected for ozone cracking are considered new parts.

Drywell ventilation exhaust valves V-27-~l and V-27 2 and drywell ventilation intake valves V-27-3 and V 27-4 are normally closed containment isolation valves which are only operated in the 24- ,

hour period preceding a shutdown, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following a startup, when a leak test is performed, and for drywell entries and during quarterly operability tests. They may also be operated in accordance with procedures following accident conditions.

GPU Nuclear has reviewed the drywell ventilation intake and exhaust valves installed location, and service and predicted environment exposure during normal and design based accident conditions. Table 2 summarizes the conditions evaluated.

RADIATION OUALIFICATION EPDM has a threshold radiation of 1x10" rads gamma. Numerous tests have shown that mild to moderate damage will occur up to a radiation of 2x10' rads gamma. However, Table 2 documents that the seats will only be exposed to a normal service of 1.23x10d rads where no degradation would have occurred. EPRI NP-4172SP (Reference 2, pgs. B-34 thru B-44) documents that EPDM hardness is not affected below 1x10' rads and 25% degradation at 6x10' to 2.2x10' rads. Therefore, the 3x10' rads OCNGS LOCA would net affect the operability of the 4

valve. Table 3 summarlies data from the OCNGS EO Cable files.

l Radiation Normal & Accident Aging Table 2 documents that these are ventilation inlet and exhaust valves which are normally i closed. Table 3 documents that EPDM insulation was subjected to IEEE 323 74 test '

sequence which included a radiation exposure of 2x10' rads TID gamma. During a Drywell LOCA, these valves would only be exposed to the Drywell LOCA gamma effects in the Reactor Building and the Reactor Building temperature. Table 2 documents a Reactor Building of 2.7x10' rads TID (worst case) during the LOCA.

Margin of Safety Test-Required ,3999 ,

Required 2x10* - 2.7x10 , 3 99g , g49g 2.7x10' However, GPU Nuclear assumed that the valve seats will be exposed to the full drywell LOCA 3x10' rads gamma and 9.5x10* rads beta. The closed valve seat working surface would be protected from beta radiation except for the seat edge on the drywell side of V-27-1 and V-27-4 (redundant series valves V-27-2 and V-27-3 are protected). The beta will be attenuated so long as the valves stay closed protecting the seat working surface.

Margin of Safety , Test - Required ,399g ,

Required 2X10* - 5X10' y 399g , 399y 5X10' THERMAL AGING ICEA publication No. S-68-516, Part 3 documents that EPR has 90^C rating. Table 3 documents the material, activation energy and thermal aging that was conducted on cable. For conservatism, an aging test of 7 days (168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />) at 12t*C with an activation energy of 1.59eV was used to demonstrate the thermal aging normal, accident and post-accident life for the EPDM seats.

Normal Service Life Valves V-27-1 and V-27-2 are exposed to a norrnal ambient temperature of 85'F Use of the following Arrhenius equation to predict the ventilation valve seal service life:

t, t2 exp -

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where:

t, = service life at 85*F unknown t, = 7 days test duration T, = 85*F service temp (302*K)

T, = 121*C test temp (394*K) 249.8"'F e = 1.59eV K = Boltzmann's Constant = 8.617E -SeV/*K 59 1 1 t'= 7 days exp _

K ,302 K 394K, ti = 10,048,117 days = 27529 years at 85*F or 150*F for 40 years The ventilation intake and exhaust valves are located in the Reactor Building and will therefore only be exposed to the 85*F reactor building service temperature during a LOCA and post-LOCA. They also will be exposed to HELB. The 124.6*F for 40 years service life (Table 2) can be shown to envelope the 40 years service at 85 F and the LOCA or HELB plus the 48-day post-accident.

V-27-1/2/3/4 CONCLUSION There is a significant margin to predict a 40 year service life, LOCA and 48-day post-LOCA for the worst harsh zone the ventilation valves are located in. However, GPU Nuclear is proposing that a maintenance / surveillance program be established to monitor one of the ventilation valve resilient seats with a portable hardness tectu and make a visual inspection for hairline cracks.

A test of the resiliert seat was condu;ted on spares located in the OCNGS warehouse (see Appendix B). A sign of cracks or an increase shift in durometer hardness in the range of 70 to 78 Shore A Durometer reading would indicate that material degradation is occurring and that the seats should be removed (at the existing or next outage of sufficient duration) for laboratory analysis. The frequency of these inspections will be based upon trending of the as-found condition of the seats.

3.2 FISHER CONTROLS BUTTERFLY VALVES V-23-13/14/15/16 The Fisher 8" butterfly valves Viton seats (SSN 000-456-5830-1) properly stored, with a durometer ,

hardness in the range of 65 to 75 Shore A Durometer reading (Appendix A), and visually inspected for ozone cracking are considered new parts.

Nitrogen purge valves V-23-13/14/15/16 will be subjected to a normal service life temperature of 81*F. These are normally closed containment isolation valves which are operated in the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period preceding a shutdown, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following a startup, when a leak test is performed, for drywell entries and during quarterly operability testing. They may also be operated in accordance with procedures following accident conditions.

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GPU Nuclear has reviewed the nitrogen purge valves installed location, service and predicted environment exposure during normal and design based accident conditions. Table 4 summarizes the conditions evaluated.

RADIATION OUALIFICATION Viton has a hardness threshold of 5x10* rads and 25% hardness degradation at 6x10* rads (Reference 2, pgs. B-151 and 15?). Table 4 documents that the valves will be exposed to a maximum of 1.93x10' rads gamma during their 40 years service life. This level of radiation is of no concern. During the LOCA accident and 48-day post-accident operation, these valves will be exposed to 3x10' rads gamma. EPRI NP-4172 SP, Reference 2, documents that very little change in seat hardness will occur and the valves will perform their intended function during the LOCA and 48 day post-LOCA period.

THERMAL AGING Dupont Cafdog E-46315 documents that Viton compounds will retain usefully elastic indefinitely when exposed to laboratory air oven aging up to 400 #F. Continuous service limits are generally considered to be:

>3000 hours at 450"F 10C': hours at 500'F 240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br /> at 550'F 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> at 600'F The Arrhenius equation was used to predict the valve seal activation energy as follows:

,i In ' K t,

Ea = - '

1 1 Y, ' 7, where:

Ea = Activation energy eV/ K 1, = 3000 hour0.0347 days <br />0.833 hours <br />0.00496 weeks <br />0.00114 months <br /> service life at 450"F t, = 1000 hour0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> service hfe at 500'F T, = 450^F (505"K) for 3000 hours0.0347 days <br />0.833 hours <br />0.00496 weeks <br />0.00114 months <br /> T., = 500'F (533'K) for 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> K = Boltzmannt Constant = 8.617E - 5 eV/ K In3000' g Ea - ' '000' - 0.91 eVl'K 1 ~

1 505'K 533'K 7

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r Normal Service Life The valve seat normal service life was determined using the following Arrhenius equation:

t,= 12exp -

where:

Ea - 0.91 eV/*K 1, = service life at 81*F (300"K) t, =

1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> at 500"F (533"K)

T, = 81*F (300'K) service temperature T, = 500"F (533"K) for 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> K = Boltzmann's Constant = 8.617E SeVf'K 1 0.91 1 1 t, - 1000 hr exp _

K s 300* K 533' K, t, = > 500000 years at 81'F or 281*F for 40 years The nitrogen purge valves are located in the Reactor Building and will therefore only be exposed to the 81"F reactor building service temperature during LOCA and 48-day post-LOCA. They also will be exposed to HELB. The 281'F for 40 years service life can be shown to envelope the 40 year service at 81*F and the LOCA or HELB plus the 48 day ,

post accident.

v-23-13/14/15/16 CONCLUSION There is a significant margin to predict a 40 year service life plus 48 day post-accident for the worst harsh zone the purge valves are located in. Because of the limited radiation data available on Viton, GPU Nuclear takes the conservative position that a 20 year service life is a reasonable expectation for this type seal. GPU Nuclear is proposing that a maintent.nce/ surveillance i program be established to monitor one of the purge valve resilient seats with a portable hardness tester and make a visual inspection for hair line cracks. A test of the Viton seat was conducted on spares located in the OCNGS warehouse (see Appendix B). A sign of cracks or an increase shift in durometer hardness in the range of 65 to 75 Shore A Durometer reading would indicate that material degradation is occurring and that the seats should be removed (at the existing or next outage of sufficient duration) for laboratory analysis. The frequency of these inspections will be based upon trending of the as-found condition of the seats.

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3.3 FISHER CONTROLS BUTTERFLY VALVES V-28-17/18 The Fisher 12" butterfly valve Nitrile seats (SSN 408-556-1430-1) properly stored, with a durometer hardness in the range of 70 to 80 Shore A Durometer reading (Appendix A), and visually inspected for ozone cracking are considered new parts.

Torus vent exhaust valves V-28-17/18 will be subjected to a normal service temperature of 85 F.

These are normally closed containment isolation valves which are only operated in the 24 hour-period preceding a shutdown, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following a startup, when a leak test is performed, for d.ywell and torus entries and during quarterly operability testing. They may also be operated in accordance with procedures following accident conditions.

GPU Nuclear has reviewed the torus vent exhaust valves installed location, service and predicted environment exposure during normal and design based accident conditions. Table 5 summarizes the conditions evaluated.

RADIATION OUALIFICATION Nitrile has a compression threshold of 1x10' rads and hardness threshold of 1x10' rads with respective 25% degradation at 6x10'and 2x10' rads (Reference 2, pages B-155 to B-158). Table 5 documents that the valves will be exposed to a maximum of 1.23x10' rads gamma during thelr 40 years service life. This level of radiation is of no concern. During the LOCA accident and 48-day post-accident operation, these valves will be exposed to a maximum 3x10' rads gamma.

Based upon EPRI NP-4172 SP, Reference 2, very little change in hardness would have occurred but compression set would be present. Some leakage could occur if the valves are operateo post-accident.

THERMAL AGING Fisher Controls thermally ages their nitrile seat material to 205 F for 70 days. The Arrhenius equation can be used to determine the maximum normal service life as follows:

t, = 12 exP -

where:

Ea = Activation Energy Constant = 0.79 eV/ K t, = Service life t, = 70 days test time at 205"F T, = 85'F (302*K) service temperature T, = 205-F (369'K) test temperature K = Boltzmann's Constant = 8 617E-5 eV/ K 9

0.79' 1 1 t, = 70 days exp ~

K r 302* K 369*K, t, = 16,704 Days = 45.76 years at 85'F.

The torus vont exhaust valves are located in the Reactor Building and will therefore only be '

exposed to the 85^F reactor building service temperature during LOCA and 48 day post-LOCA.

They also will be exposed to HELB. The 45.76 years service at 85 F can be shown to envelope the 40 year service at 85"F and the LOCA or HELB plus the 48-day post-accident.

V-28-17/18 CONCLUSION GPU Nuclear's analysis documents that the 'orus to vent exhaust valve seals will perfonn their intended function through 40 years service, LOCA and 48 day post-LOCA operation. Becat.se of the limited radiation data available on Nitrile, GPU Nuclear takes the conservative position that a 20 year service life is a reasonable expectation for this type seal. GPU Nuclear is proposing that a maintenance surveillance program be established to monitor one of the valves resilient seats with portable hardness tests and make a visual inspection for hair line cracks. A test of the Nitrile seat was conducted on spares located in the OCNGS warehouse (see Appendix B).

A sign of cracks or an increase shift in durometer hardness from range 70 to 80 Shore A Durometer reading would indicate that material degradation is occurring and that the seats should be removed (at the existing or next outage of sufficient duration) for laboratory analysis.

The frequency of these inspections will be based upon trending of ti e as-found condition of the seats.

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3 4.0 TABLEJ TABLE 1 TAG # L.OCATION MANUFACTURER SEAT TYPE V-27-1 R4 RA(18'6") Centerline EPDM V-27-2 R4-RA(18'6") Centerline EPDM V-27-3 R4-RF(84') Centerline EPDM V-27-4 R4-RF(84') Centerline EPDM V-23-13 R4 RF(84') Fisher Controls VITON V-23-14 R4 RF(84') Fisher Controls VITON V-2315 R2-RE(44'9") Fisher Controls VITON ,

V-23-16 R1-RE(44'9") Fisher Controls VITON V-28-17 R4-R A(11'6") Fisher Controls NITRILE V 28-18 R4-RA(11'6") Fisher Controls NITRILE I

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TABLE 2 TAG # V-27-1 V-27-2 V-27-3 V-27-4 Location R4-RA R4 RA R4-RF R4-RF Elevation 18'6" 18'6" 84' 84' Function Drywell Drywell Drywell Drywell Ventilation Ventilation Ventilation Ventilation Exhaust Exhaust Intake intake Valve position Closed Closed Closed Closed RB Service temp 85'F 85'F 81'F 81*F RB HELB temp 124 6"F 124.6*F 124.6*F 124.6*F DW LOCA temp N/A N/A N/A N/A RB Normal Gamma 1.23x10' 1.23x 1')' 1.1 x10' 1.1x10' DW Normal Beta N/A N/A N/A N/A DW LOCA Gamma 3x10' 3x10' 3x10' 3x10' DW LOCA Beta RB TID 2.7x10' 2.7x10' 2.7x10' 2.7x10' RB Beta N/A N/A N/A N/A DW TID 5x10' 5x10' 5x10' 5x10' DW Normal Gamma 2x10' 2x10' 2x10' 2x10' Post-Accident Operation 48 days 48 days 48 days 48 days

• During accident conditions, the valves will be closed. Valves V-27-1 and V-27-2 are in series as are V-27 3 and V-27-4. Valves V-27-1 and V-27-4 are closer to the drywell and will be exposed to Beta radiation at the seat edge facing the drywell atmosphere.

Deterioration of this seat edge is of no concern as the deterioration will not extend into the sealing surface.

Note: TID, Gamma and Beta units are rads for this table.

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....._ ., . - ..- - __ _ = . - . .- . - _ . .. . . - . _ . .

IABLE_3 CABLE TEST THERMAL ACTIVATION TEST TEST EQfAg LN3.U.LATION AGING ENERGY. TEMP R ADI ATi DE EO-OC-311 EPR 168HR@121*C 1.62eV 346*F 2x10' EO OC-341 EPR 1440HR@155"C 1.1380V 340"F 2x10' EO-OC-348 EPR 168HR@150'C 1.21 eV 385"F 2x10' EO OC-370 EPR 168HR@150 C 1.69eV 346'F 2x10'  !

EO-OC-383 EPDM 168HR@121*C 1.59eV 375"F 2x10' EO-OC 385 EPR 504HR@150"C 1.440V 345"F 2x10' Noto; EPR and EPDM belong to the same family and therefore the data can be used to dernonstrato generic values.

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TAG # V-23-13 V-23-14 V-7315 V 23-10 j Location R4 RF R4-RF R2-RE R2-RE Elevation 84' 84' 44'9" 44'9" Function Purge Purge Purge Purge Valvo position Closed Closed Closed Closed  :

RB Service temp 81'F 81*F 81*F 81*F  !

RB HELB temp 233' F 233'F 133F 133*F DW LOCA temp N/A N/A N/A N/A RB Normal Gamma 1.12X10' 1.1 x10' 1.93x10' 1.93x10' DW Normal Beta N/A N/A N/A N/A i DW LOCA Gamma 3x10' 3x10' 3x10' 3x10' DW LOCA Beta i RB TID 6x10' 6x10' 2x10' 2x10' RB Beta N/A N/A N/A N/A DW TID 5x10' 5x10' 5x10' 5x10' DW Normal Gamma 2x10' 2x10' 2x10' 2x10' Post-Accident Operation 48 days 48 days 48 days 48 days

• During accident coclitions the valves will be closed. Valves V 23-13 and V-23-14 are in series as are V-23-15 and V-2316. Valves V 23-14 and V-2316 are closer to the Drywell and will protect V-23-13 and V-2315 from beta radiation Valves V-2314 and V-2316 seat edge facing the Drywell atmosphere will be deteriorated by the beta radiation. This is of no concern as the deterioration will not extend into the sealing surface.

Note: TID, Gamma and Beta units are rads for this table.

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TAG # y_2fLIZ V-28-18 location R4-RA R4-RA -

Elevation 11'6" 11'6" Function Torus to Vent Exhaust Torus to Vent Exhaust Valvc position Closed Closed RB Service temp 85F 85*F RB HELB temp N/A N/A DW LOCA temp N/A N/A RB Nonnal gamma 1.23x10' 1.23x10' DW Normal beta N/A N/A DW LOCA gamma 3x10' 3x10' '

DW LOCA beta

• 9.6X10" 9 GX10" RB TID 2.7X10' 2.7X10' RB Beta N/A N/A DW TID 5x10' 6x10' DW Normal Gamma 2x10' 2x10' Post Accident Operation 48 days 48 days

• Durin0 accklent conditions the valves will be closed. . Valves V 2817 and V-28-18 are in series. Valves V-2817 is closer to the Drywell and will be exposed to bota radiation at its seat edge facing the Drywell atmosphero. Deterioration of this seat edge is of no concern as the deterioration will not extend into the sealing surface.

Note: TID, Gamma and Beta units are rads for this table.

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5.0 REFERENCES

1. EPRI NP-3817, "Oualification of Active Mechanical Equipment for Nuclear Plants",

dated March,1985.

2. EPRI NP-4172SP, " Radiation Data for Design and Qualification of Nuclear Plant Equipment".

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6.0 APPEUDipfS Appendix A - Manufacturer's Seat Hardness Data i

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(201)670-8070 ValleyTechnica1 Sales, Inc. I Hollywood Asc. 2A, Ho Ho Kus, N.J. 07423 In 201)6 70-8778 t

14a rch 2 4 , 1994 GPU Nuclear Corporation One Upper Pond Road Parsippany, New Jersey 07054 Attention: Paul E. Boucher

Reference:

Service Life of Valve Seat Your Letter of 3/22/94 5350-94-056

Dear Paul:

Confirming our phone conversation, Center Line's standard EPDf4 seat durometer is 70 to 78, shore A.

Sincerely, VALLEY TECHNICAL SALES, INC.

O. .

() &'f}l:ui--'<CC%Q' David L. Whittemore

PtSHIR Memorandum to Jon McElhaney From: Paul Gassman C.B. Ives, Inc.

cc: Peg Cline Paul Abens care April 20,1994 sum.ct Butterfly Liner Material Ouestions from GPU Nuclear Jon; in response to GPU Nuclear's questions of March 16 (attached), we can respond as follows.'

1 Fisher publishes a maximum recommended service life of 4 years for elastomeric parts in nuclear service. (See attached

• Elastomeric Parts Service Life" statement ' rom Catalog 11).

This is based on elevated temperature testing that was performed on nitrile parts a number of years back. The actuallife may vary, depending on temperature and radiation levels. It is entirely possible that the customer may experience a longer service life; the results of the testing were considered to be somewhat conservative, and many diaphragms, liners, etc. have lasted longer in mild temperature environments.

2. If hardness checks indicate that the hardness of the elastomeric liners is beginning to fall outside of those specified for the material, Fisher would consider it "out of spec". with the possibility of high break-out torques developing. Additionally the T-seal could begin to crack or even break off in areas.

3. The durometer hardness specified for the viton liners is Shore "A" 70 +/ 5, measured at 70 degrees F(+/ 5 degrees).

Our current nitrile liner vendor specifies a hardness of 60 +/ 5. It should be noted that another liner supplier inat has also been used for nitrile liners has specified 70 +/ 5 for their product.

I hope that this answers the bulk of your customer's questions. If they desire, they can call me at 515-754-2380. or fax me at 515 754-2830 if they need additional info.

Best regards; C ,

Paul Gassman Fisher Nuclear Power Team 300C '330 NIYW N201.W 969C F94 STSG 9F;60 F64tek0

Appendix B - Oyster Creek Test Hardness Data for New Warehoused Seats l

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