Text

Proposed Tech Specs,Consisting of Tech Spec Change 88-06 Re Leak Rate Test Frequency for Containment Purge Supply & Exhaust Isolation Valves
	ML20196D162
Person / Time
Site:	Sequoyah
Issue date:	12/05/1988
From:	; TENNESSEE VALLEY AUTHORITY
To:	;
Shared Package
ML20196D156	List: ML20196D150; ML20196D162;
References
	NUDOCS 8812080229
	Download: ML20196D162 (75)
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ENCLOSURE 1

-- - -- .-. PROPOSED TECHNICAL SPECIFICATION CHANGE

- SEQUOYAH NUCLEAR PLANT UNITS l'AND 2 DOCKET Nos. 50-327 AND 50-328-.

(TVA-SQN-TS-88-06)

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O CONTAINMENT SYSTEMS C CONTAINMENT VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION

3. 6.1. 9 One pair (one purge supply line and one purge exhaust line) of containment purge system lines may be open; the contaiNnent purge supply and exhaust isolation valves in all other containment purge lines shall be closed.

Operation with purge supply or exhaust isolation valves open for either purging or venting shall be limited to less than or equal to 1000 hours0.0116 days 0.278 hours 0.00165 weeks 3.805e-4 months per 365 days. .

The 365 day cumulative time period will begin every April 15. R22 APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a. With a purge supply or exhaust isolation valve open in excess of the above cumulative limit, or with more than one pair of containment purge system lines open, close the Isolation valve (s) in the purge line(s) within one hour or be in atfollowing the least HOT STANCBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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SURVEILLANCE REQUIREMENTS 4.6.1.9.1 The position of the conta'inment purge supply and exhaust isolation valves shall be determined at least once per 31 days.

4.6.1.9.2 The cumulative time that the purge supply and exhaust isolation valves 7 days. are open over a 365 day period shall be determined at least once per R22 y, f, \ 9,3 /]l lenK h encel f'er 3 Men Y 5) O " " ' ^ '" '

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CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

c. Verifying that on a Containment Ventilation isolation test sigral, each Containment Ventilation Isolation valve actuates to its isolation position.

4.6.3.3 The isolation time of each power operated or automatic valve of r.16 Table 3.6-2 shall be determined to be within its limit when tested pursuant to ~~' '

Specification 4.0.5.

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MAR 251982 .

SEQUOYAH - UNIT 1 -

3/4 6-18 Amendment No. 12 4

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CONTAINMENT SYSTEMS CONTAINMENT VENTILATION SYSTEM-LIMITING CONDITION FOR OPERATION 3.6.1.9 One pair (one purge supply line and one purge exhaust line) of containment purge system lines may be open; the containment purge supply and exhaust isolation valves in all other containment purge lines shall be closed Operation with purge supply or exhaust isolation valves open for either purging or venting shall be limited to less than or equal to 1000 hours0.0116 days 0.278 hours 0.00165 weeks 3.805e-4 months per 365 days The 365 day cumulative time period will begin every January 1. R9 APPLICABILITY: HODES 1, 2, 3, and 4.

LCTION:

ct.. Wi'.h a purge supply or exhaust isolation valve open in excess of the above cunulative limit, or with more than one pair of containment purge system lines open, close the isolation valve (s), in the purge line(s) within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ,

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3URVEILLANCE REQUIREMENTS 4 6.1.9.1 The position of the containment purge supply and exhaust isolation valves shall be determined at least once per 31 days.

4.6.1.9.2 The cumulative time that the purge supply and exhaust isolation valves are open over a 365 day period shall he dc;ernined at least once per.

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CONTAINMENT SYSTEMS SURVEILLANCE RE0VIREMENTS (Continueo) 4.6.3.2 Each isolation valve specified in Table 3.6-2 shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once per 18 months by: -

4. Verifying that on a Phase A containment isolation test signal, each Phase A isolation valve actuat,es to its isolation position.

b. Verifying that on a Phase 8 containment isolation test signal, each Phase B isolation valve actuates to its isolation position.

c. Verifying that on a Containment Ventilation isolation test signal, each Containment Ventilation valve actuates to its isolation position.

4.6.3.3 The isolation time of each power operated or automatic valve of Table 3.5-2 shall be determined to ce within its limit when tested pursuant to Specification :.0.5.

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ENCLOSURE 2 4

PROPOSED TECR!!ICAL SPECIFICATION CHAN0E

~SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS.00-327 AND 50-328-(TVA-SQN-TS-88-06)

DESCRIPTION AND JUSTIFICATION FOR RELAXATION OF TEST FREQUENCY FOR CONTAINMENT PURGE SUPPLY AND EXHAUST ISOLATION VALVES T

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ENCLOSURE 2 DESCRIPTION OF CHANGE s

Tennessee Valley Authority proposes to modify the Sequoyah Nuclear Plant Units 1 and 2 Technical Specifications to provide a relaxation in the test frequency for the containment purge supply and exhaust isolation valves.

Surveillance requirement (SR) 4.6.3.4 currently states that each containment purge isolation valve be demonstrated operable (undergo a type C leak test) within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after each closing of the valve except when the valve is being used for multiple cyclings. Any purge valve that has undergone multiple cyclings is required to be tested at least once every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Operability is demonstrated by performance of a type C '

leak test to verify that the measured leakage rate from each purge valve, when added to the leakage rates for all other type B and C penetrations, does not exceed 60 percent of the maximum containment leakage rate (La).

In lieu of SR 4.6.3.4. TVA proposes to add a new surveillance requirement under Containment Ventilation System Specification 3.6.1.9. The new SR 4.6.1.9.3 would relax the 24/72-hour test requirement to include a 3-month test interval and would establish a specific maximum leakage rate of 0.05 La for each purge valve. An action statement from revi.. ion 5 of the NRC standard technical specifications (STS) was included to address operability requirements for leakage in excess of 0.05 La to be consistent with the revised surveillance requirement. These changes are patterned after revision 5 of the NRC STSs.

REASON FOR CHANGE The pecposed change requests a relaxation from the current 24/72-hour test interval for SQN's purge system containment isolation valves. Relaxation of the test interval to once per three months would provide significant benefits and savings to TVA in the areas of ALARA (as low as reasonably achievable), cost, and plant safety.

ALARA Under SQN's current surveillance requirement, test personnel are required

to enter the annulus at least three times each week for a minimum test duration of one hour. This places test personnel in areas of low to intermediate radiation for prolonged periods of time, resulting in additional exposure. By reducing the number of times test personnel enter the annulus s252 entries versus 8 entries) over 1-year time period, the annual saving in net exposure is estimated to be 3 man-rem.

Cost Under the current 24/72-hour test frequency, at least 252 tests are performed each year for both units. Considering that each complete test requires three technicians for a minimum of three hours at an average cost of $40 per hour per technician, the annual cost to TVA in testing alone equals $181,000 for both units. Additional overhead costs needed to support each test, such as planning, scheduling, issuing, and reviewing '

4

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_2 test packages, are approximately $300,000 annually. The proposed 3-month test interval would reduce the total humber of tests to eight times per year, resulting in an annual saving to TVA of $290.000. The net saving projected over SQN's remaining operational life would therefore be very significant.

Plant Safety During the type C leak test of SQN's purge system containment isolation valves, a test connection valve is opened between the inboard and outboard purge isolation valves creating a containment leak path. While the test ,

valve is open, the plant must enter a 1-hour limiting condit!sn for operation (LCO 3.6.1.1). The frequent entering and exiting if an LCO to perform surveillances place an undesirable burden on the Opetations staff to record and status each entry. Relaxing the test interval to'once every three months would alleviate placing the plant into frequent LCOs and would provide more freedom to the operators for monitoring plant conditions.

JUSTIFICATION FOR CHANGE

System Description

Each reactor building purge system (one system per unit) provides mechanical ventilation of the primary containment (upper and lower), the instrument room located within containment, and the annulus located between the containment and shield buildings. The system is designed to supply f resh air for breathing and contamination control to allow personnel access for maintenance and refueling activities. The system consists of two purge air supply fans and two purge air exhaust fans for the containment and annulus areas, two cleanup filter trains, an instrument room supply fan, air supply and exhaust ducts, and a series of containment isolation valves (20 per unit). The SQN purge system containment isolation valves are motor-operated Henry Pratt Mark II butterfly valves with valve sizes ranging from 8 inches to 24 inches (see, figure 1 of this enclosure).

Each purge system containment penetration includes both inboard and outboard isolation valves designed for minimum leakage in their closed position. Each containment purge penetration is also equipped with an additional valve (shield building isolation valve) outside of the outboard isolation valve. A typical purge penetration is shown in figure 2 of this enclosure. As part of SQN's secondary containment, enclosure, leakoff lines are provided on each purge penetration. These leakoffs ensure that the enclosed volume between the outboard valve and the shield building isolation valve remain open to the annulus during isolation. Any outward leakage from the primary containment or inward leakage from inside the shield building will be drawn into the annulus, which remains under a negative pressure. Annulus air is then processed by the emergency Sas treatment system, which is a safety-grade filter system.

Each purge. penetration includes.an. isolation mechanism that is activated i.

by-a-containment isolation signal or upon detection of high radiation in the purge exhaust. It should be noted that, although the shield building isolation valves are not considered containment isolation valves, they also close upon a containment isolation signal. For a complete system description, refer to sections 9.4.7 and 6.2 of the SQN Final Safety Analysis Report (FSAR).

The proposed change is needed to help bring current test requirements into conformance with NRC STSs. The STSs recognize two leak-test frequencies for purge supply and exhaust isolation valves a 6-month test frequency for 42-inch purge valves and a 3-month test frequency for 8-inch purge valves. The STSs also state that the measured leakage rate be less than or equal to 0.05 La for both size valves. The test requirements for the 42-inch purge valves would not be applicable to the proposed change because SQN, by design, does not have 42-inch purge valves. SQN's purge valves range in size from 8 inches .o 24 inches. Because the STS leak rate criterion of 0.05 La is being applied to all of the SQN purge valves, the proposed change is considered to be conservative.

It is important to note that the SQN valves are equipped with qualified resilient seats that lead to improved leakage performance. TVA conducted an environmental material degradation analysis for the soft-seat material within these valves (trademark name is Resiloseal "W", which is a rubber compound known as ethylene propylene diene monomer). A copy of the calculations used to demonstrate the acceptable service life is included in enclosure 4. This analysis was performed to ensure that the seating material would not degrade under worst-case accident conditions (high temperatures and high radiation). The analysis was divided into two parts:

1. Thermal Analysic

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The thermal analysis evaluated the Resiloseal W against the most conservative activation energy for that material and calculated an expected life based on required service temperatures. An accident ,

equivalency calculation was then performed to determine the life expea.tancy during a worst-case postulated accident. The accident equivalency time is then subtracted from the expected life, which provides a thermal life for the material.

2. Radiation Analysis The radiation analysis evaluated the Resiloseal W against a worst-case design basis accident. A radiation life was determined by subtracting the postulated accident dose from the materials radiation damage level and then dividing by the yearly normal operating dose rate.

The results of the thermal and radiation analysis indicate that the service life of the Resiloseal W material is well above the 40-year design life of the plant. Based on these analyses, the Resiloseal W seats would function properly (remain resilient) during normal operation and 6

4

-postaccident operating conditions. The proposed change to relax the leak-test frequency from 24/72 hours to 90 days would therefore not reduce the valve's ability to provide an adequate seal for containment isolation.

Experience has shown that SQN's purge valves have a high degree of reliability with regard to seat leakage. Historical leak rate test data collected over a 3-year period indicate that these valves have consistently low leak rates. Over 99 percent of the tests indicate a leak rate less than 1.44 standard cubic feet per hour or 0.0064 La. Because the proposed change only affects the leak rate test frequency and does not physically affect the valves or their design, the level of confidence in

l the valve's ability to remain leak-tight is not significantly reduced.

TVA conducted a review of the Henry Pratt Installation and Service Manual for Nuclear Class Valves to determine if any additional maintenance would be required or recommended for SQN's purge valves. The vendor manual stated that the best method for detecting valve degradation would be to l perform a leak test by pressurizing the area between the valves. The I

vendor manual gives no frequency other than at periodic intervals at the l owner's discretion. In light of SQN's excellent leak-test history and the l known information on the sof t-seat material, TVA considers the proposed 3-month leak-test frequency to be conservative for detecting leakage past the seat.

TVA proposes a specific maximum leakage rate of 0.05 La, which is consistent with the NRC STSs and NRC requirements for maintaining specific leakage paths. SQN's present technical specification SR does not provide a specified leakage limit for each individual purge valve. The proposed leakage limit of 0.05 La is considered to be reasonable with regard to the l overall leakage limit of 0.60 La. Additionally, this limit is not so l overly restrictive that it would create an operational problem when l

testing is conducted witb the plant online. The corresponding action statement is taken from revision 5 of the NRC STS. It provides specific timeframes to correct excessive leakage problems. Associated shutdown requirements are also included in the event that ~Se leakage problem cannot be repaired. ,

In conclusion. TVA considers the proposed change to be justified for three reasons. First, the proposed change would standardise the current SQN technical specifications. Second, the relaxation in test frequency would continue to maintain valve reliability with regard to seat leakage.

Third, the proposed change would establish a specific maximum leskage limit for each purge isolation valve.

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ENCLOSURE 3 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-32S (TVA-SQN-TS-88-06)

DETER.MINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS .

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ENCLOSURE 3 Significant Hazards Evaluation TVA has evaluated the proposed technical specification change and has determined that it does not repreLent a significant hazards consideration based on criteria established in 10 CFR 50.92(c). Operation of SQN in accordance with the proposed amendment will nott (1) Involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed change conforms to

the current NRC STSs for establishing a specific maximum leakage limit (0.05 La) (nd associated action statements for each containment purge supply and exhaust isolation valve. This leakage limit is a conservative limit with regard to the overall leakage limit of 0.60 La. TVA has. evaluated the leakage characteristics of the subject purge valves under normal operation and under worst-case postaccident operation to ensure that the proposed relaxation in leak-test frequency from the current 24/72 hours to 90 days would not reduce valve reliability for containment isolation between tests. No new hardware or operating practices are introduced by these changes.

Consequently, the probability or consequences of accidents previously evaluated are unchanged.

(2) Create the possibility of a new or different kind of accident from any previously analyzed. The proposed change establishes a specific maximum leakage rate of 0.05 La and associated action statements for SQN's containment purge supply and exhaust isolation valves, which is consistent with the NRC STSs and NRC requirements for maintaining specific leakage paths. TVA conducted an environmental material degradation analysis of the soft-seat material within these valves to ensure the seating material wculd not degrade under normal operation or worst-ease accident conditions. Based on TVA's analysis, the valve seats would continue to function properly (remain resilient)t and, therefore, the proposed relaxation in test frequency from 24/72 hours to 90 days would not reduce the valve's ability to seal for contain=ent isolation. The proposed change does not physically affect these valves or their design consequently, the possibility of a new or different kind of accident from any previously analyzed is not created.

(3) Involve a significant reduction in a margin of safety. The proposed change deletes the existing SR (SR 4.6.3.4) and functionally groups it under the containment ventilation system specification as a new SR *

(SR 4.6.1.9.3). An action statement from revision 5 of the NRC STS ,

was included to address operability requirements for leakage in excess of 0.05 La to be consistent with the revised surveillance requirement. Because the new SR conforms to the current NRC STSs and establishes a specific maximum leakage rate for the SQN containment purge supply and exhaust isolation valves, the proposed change is considered to be a technical specification improvement. TVA's -

evaluation indicates that the soft-seat material within SQN's h

purge valves would remain resiliert under normal or postaccident operating conditions and that the service life of this seat material is well above the 40-year design life of SQN. Based on TVA's analysis, the proposed change to rielax the leak-test frequency from 24/72 hours to the proposed 90-day frequency would not rec.uce the valve's ability to sesi for containment isolation nor wouit the valve endergo catastrophic failure between tests. Consequently, the proposed change will not significantly reduce the margin of safety.

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PROPOSED TECHNICAL SPECIFICATION CHANGE ,

SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 :

DOCKET NOS. 50-327 AND 50-328 (TVA-SQN-TS-88-06) i Calculation Entitled "Post-LOCA Location Specific Beta Doses

for Containment Isolation Valves " SQN-APS3-091 (825 880927 803) ;

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l l The beta dose due to etternal sources was investlgated using the pipe wallI thic the betas present to detenalne if there is a significant contribution toroi)the llThesedoseswerefoundtobenegligible. of intern urces.,,l l

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l The valves were of three different sites: 8",12", and 24" ,1he musinun beta dose calculated for the 8" I

\ valves was 2.7E+2 Rad, for the 12' valves the sustaan beta dose was 6.1E+2 R lmaalemsacalculatedbetadosewas5.6E+3 Rad.

l I- l !

l, l t ) Micmfila and store calculations in RIMS 5ervice CenterMicrofilm and destroy. ()

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cc RIMS, 5L 26 C-K t Wu -_ LtA0 Mutt AL t+&t. Address m0 c;;; M D5t A SQ4p' yl W.7.oM %/lcht-N' 4488Q r.A.ddleA p Dst-t, sN

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Attact ment No. 1

.' ,/ Pcg3 1 cf 2 GhjJ,U, LATICN CLASSIFICAT!CN

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PLANT /uM:T .. S O N lE2 tCENTtrtEn StW 4csA-09i RIM 5 NO. ,. ISSUE CATE 4 ' 2 7* bb TITLE mom - I. I.Y. A L. 0c A non 5 me 8it @> i' b 'Di% F i 30 COOtts i blMPFd 7 '3 o c 4"h0 Ai \/ A Lvf 6 REVISICN LEVEI. O

.rrrmre sysernes3 PLANT FEATURE: SYSTEM /CCMPCilENT CESC?.:PT:CN:

k SAFETY SYSTEM SYSTEM No. 30 Co4J+n iNM e el u a +. . o n , . . , n

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[ eLANT ENVIRCNMENT CEQ, ETC.)

O "c"-sarETv SYSTEn sysTrn "o-O areE.io:x R C Czurt. STRUCTURES

[,INSTRUMENTATICN C 1. S7, P At1. ETC)

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c] CE rRaStE E' SueERCECEo suSnzrTED C 2, d /,. d CaTE 9

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@LCULATION CLASSIFICATICM ,

IDENTIFIE . bbN SN>l d4/ '

MRELIt1INAR'Y CLASSITICATICN: -

s E ESSDITIAL FILE CbLY C DESIRABLE Q SUPERCIDED CALCULATICN CLASSIFICATION JUSTIFICATION: -

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SUBMITTER: '4 At l eo /9 40 ad///SRS Ori s 7% .4 ruo1

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NKP-3.1 At t a chtherit Page 1 of 1 CAI.CULATION DESIGN VERIFICATION (INDEPENDENT REVIEW) FORM ,

SotJ- A p53 - 69 i RO.

Calculation No. Revision

Method of design verification (independent review) used (check method used):

1. Design Review .

2. Alternate Calculation

3. Qualification Test Justification (emplain below):

Method _11 In the design review method, justify the technical adequacy of che calculation and explain how the adequacy was verified (calculation is similar to another, based on accepted handbook methods, appropriate sensitivity studies included for confidence, etc.).

Method 7: In the alternate calculation method. identify the pages where the alternate calculation has been included in the calculation package and explain why this method is adequate.

Method 3: In the qualification test method, identify the QA documented source (s) where testing adequately demonstrates the adequacy of this calculation and.esplain.

~This tMleida b1v is i}m% A D9JC/NER Ctic GEU L//tL 3-ntt

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AG 64 25 EM) '% artuwalthA edtd a tt NAh<nhf> eruA 4G ealk en im bo n vid e w k/ d tM/nn. OJrci us.A wkw twet. ~

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ddpendent Reviewer) !

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. ...%J,- w' sat , , . *. 3 ' *;.?? (#. AUM

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TENNESSEE VALLEY AUTHORITY S, ~

,'1 w 9 .

SUBJECT Post-LOCA Location Soweific Beta Doses for PROJECT SON

_ Containment isolation Velves BCML"53-091 RS ;

.~.. .. ...

q _ u - es '""'" g a " << h s %-

i l

Purnose ;

- f The purpose of this calculation is to determine the I post-LOCA beta radiation doses to internal components of the t containment isolation valves listed in the Mechanical Engt- !

neering Branch reouest O!R MEB SON 88032 RQ. This calcula- '

tion is being done in support of Qualification Aging Analysis Report, AA-19. l I

Introduction !

DNE/NTB/APS3 roccived a request (OIR MEB SON E803: RO, [

Ref. 1) to calculate beta radiation donas to internal compo-l nents of system 30 valves an contatoment. These valves are -

butterfiv valves on the containment ventilation system.

The methodology for calculating the beta dose due to internal sources will be to first calculate an internal ;

source volume for each energy group based on the range, in air, of beta particles with these energies. The dose to the !

valves from internal sources will then be calculated by :

scaling the source from containment to a source instdo the i valve using ttie internal source volumes, the free volume of containment, and the doses provided in reference 5. This

( '

methodology is the same as that tled in reference 7.

1

. l The dose due to external sources will be investigated using the pipe wall thickness and the range (in iron) of the l 1

betas present to determine if there is a signtftcant contri- ;

l bution to the internal dose from external sources. l l i 1

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l e i e b y

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TENNESSEE VALLEY AUTHORITY S.,u , 2 ., Q !

f SUBJECT Post-LOCA Leestion soveidte BeQA Deseg der PROJECT SQN t

_ Containment Isolation Valves SC.40rG3-091 R8 g g_ gg m o.. . g g .... g Assumotiong l i

1.) It in issumed that dry air at standard temocrature '

and Dressure (1. QUE-3 gm/cc. Ref5 4) can be used to model the gasses inside the valve. This is a conservative assumption since the air normally in ,

containment would be "wet" and this motst air l would increase the attenuation of Oetas.

l i

[

2.) Since the valves aro actu.\ted on a containment ventil ation isol ation signal. it is assumed that these valves will close during a LOCA.

3.) The internal source to the valv' * .an be modeled !

as a evlinder with a clameter ecual to the piping -

diameter. The length will be ecual to the range of the betas in air or the available distance in ;

the otoe (the distance from the valve to the first ;

bend or grille in the espe) whschever as smaller.

Cal cul ationn '

t

(

The beta accident cose ( D.s ta . anunt ) to the valves are given in reference 5. Dee,a, uu.n, is the containment free field beta dose from the environmental drawings which were ,

calculated in reference using the volume of contLinment and considering the source as a semi-infinite cloud. to i determine the beta done. .

l l

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TENNESSEE VALLEY AUTHOR!TY 0.

e, 9 ;

t SUSJECT Post-LOCA Location Goecidic PeQ3 Deses #ce_, J ROJECT ,

5QN - l Contatnment_!selatie>n se vee e, e.,e Valves - - -

SQNAPLU 091 R9 f

$ - l 't,. - % s..cn e er[p d,; e.,e g11/4 ,

Internal sources L

The internal source for a valve wo.uld havn a volume :

(assumed to be a cylinder) calculated from the valve otoe -

radius and the beta range in air or length of ospe (see ajaumotion O). The source volume in the otos can oc a function of the beta energy since eacn energy has a '

different f*ange in air. +

i

(

The ratto of the internal source volume for a particu- !

lar beta energy to the volume of containment. ti ta s s the ;

probacility that a beta particle has that enorgy gives a '

rnultiplier that will scale Dee,...cian , to an internal valve r dose for the beta energy under consideration. The sum of ,

the multioliers over all energte, gtvos a multiplier to de- [

l termine the total internal beta vose. Ecuation (1) in used

! to find the internal dose to the valvos under consiceration. '

i i

as (Vole .cs:) I l Dn ,.. i .,,. = I: rP.) Dn ,. . usin., (1)

'"' l 1 ( Vol e .. i....,)

l .

l I !

j Wheren l l

l Dee . . .es .* = The dose to any point inside the valve 1

I *- The l'" ene gy group Vol me =

The source <olumo (see aseumo-l tion ').

3 t c reiur , a C o. sanment free volume from reference 2. ++

= 1. 041E+6 f ta . 3, g t 4g, go emi i

I l

l . 3 W

Q '

. . . . . .. . .2 '

J

, TENNESSEE VALLEY AUTHORITY S,.., 4 ,, 9 SUBJECT _ Post-LOCA Location Soecific Beta Doses for PROJECT SON

_ Cont ai nment - Isol ati on Val ves

. ~ ,.. .. EQNAPS3-091 RO g ..u...

.g ....

g P = The probability of the 1" beta energy group. The hard-est soectrum occurs one day post-LOCA, therefore this is the spectrum used for this calculation. The one day beta spectrum probabilities are taken from reference 3.

The valves listed in reference 1 are of three different pipe sizes: 8", 12". and 24". thus the volume of the source must be calculated for three different cases.

The source volume i s calculated as below:

Vol s.uace : = nC (Radius) (2. 54) ]: (Ri) (Dens. )-t , (2.1) where the radius is in inches, the factor 2.54 converts the inches to centimeters. F is the beta range (in air) o t' the I" energy group in gm/cm . The density of air converts the units of the range to centimeters. The beta ranges and the density of air (1.205E-3 gm/cc) are given in reference 4.

Values 20 pipes.

for the radii are teken f rom ref erence 6 f or schedule The value for the range is bound by an upper limit, that i s, when the range of the betas is greater than the distance fecm the valve to the first bend or grille in the pipe. These distances were found on the drawings listed in reference 8. The l argest distance from the valves to the first bend or grille for each valve si:e, Lnas, is 108

(274.32 cm) for the 24" valves and 24" (60.96 cm) for the 12" and 8" valves. Therefore. equation 2.1 is used when the range is less then Lui and equation 2.2 is used wnen the range is greater th an Lnas.

Vol ....ce : = nt (Radius) (2.54) J2 Lui (2.2) i I

e4 4

5 1

'k s .; w .* . u s sue _3

TENNE _TSEE VALLEY AUTHORITY s,,,,, 5 , 9 SUBJECT Post-LOCA Location 90eci f i e Beta Doses for PROJECT SON Containment Isolation Valves SQNAPS3-091 R0

.......,/ gA ..,.

9-2.S-68 t /u/ee Substituting ecuation (2.1). the density of air. and the free volume of containment into equation (1) results inn if R: is less than Ln. , then:

as /

M= E (2.362E-8) (n) (Radius *2.54) 2 (R ) (P:) ,

-s else by substituting ecuation (2.2):

.s /

M= E (2.846E-11) (n) (Radius *2.54) 8 (Ln. ) (P ) ,

-s and Deer.. nien..e = M

D. , . . .c e s ... , (3)

Where M is the dose multiplier. The dose multiplier is calculated for each valve st:e in Table 1.

4 e

t t

,w t

._ . .- _ . . - : .2

i_, . . - -

TENNESSEE VALLEY AUTHORITY '. i ser 6 o 9 SUBJECT Post-LOCA Location Soecifie Beta Doses for iROJECT SON Containment Isolation Valves SQNAP?T-091 RO

..se senevtse av

, , csenseevj o.re g

Table 1 '

Calculation of Beta Dose Multioliers Energy 1 Day Range B' 12' 24' Rage ,

Group (MeV1 frobability Ecol Multiplier Multiplier Multiplier Ige /catl NTA:

1 2.5ME-01' 4.592E41' 5.985E+01 / 2.673E-0 6.077E-0 A.189E46 /7.212E-02' l.20$E43 dir density 2 5.000E41' l.637E-Ol' !.656E+02 9.500E-0 V 2.160E-07# 2.!!3E-06. / 1.995E-01 ' 6.096E+011 12' & 8' L 3 7.500E41 - 5.8ME42' . 2.846E+02 3.366E-0V. 7.65tE-08 1.240E46 3.429E41' 2.743E+02 C24' L...

4 1.M0E+00 - 3.790E-02' 4.076E+02' 2.199E-08/ 5.M0E-08 8.105E-07 4.912E41 "1.032E+0lT8' radius (cal

- 1.556E+0!T12' radius (ce 5 1.250E+00' 3.170E-02' 5 '!3E+02/ !.840E-08 4.!BIE-08' 6.779E-07 6.408E41' 2.953E+0r = 24' radius (ca 6

1.500E+00 - 3.020E42 6.556E+02 !' 1.753E48 3.984E4B 6.456E-07 7. 9ME-0! '

7 1.750E*00' 2.9ME-02 7.786E42 1.682E48 3.826E48 6.201E-07 9.382E-Ol' 8 2.000E+00' ?.660E-02 ' 9.004E+02 1.55 E-08 3.53 E-08 5.731E-07 1.06 E+00 "

9 2.250E+00 ' 2.390E42 ' l.02tE43 1.387E4B 3.153E-08 5.lllE-07 1.230E 40 '

10 2.500E+0F 2.090E-02' l.140E43 1.I!3E-08 2.757E-08 4.469E47 1.374E+00-

!! 2.750E40< 1.8GE42' l.2:3E+03 1.06BE4B 2.427E-08 3.93!!-07 1.516E+00 '

12 3.000E40.- 1.660E42 ' !.376E43 9.634E49 2.190E48 3.5:0E47 1.S!SE40 '

13 3.250E40' l.540E42 ' l.4?tE+03 8.137E-09 2.032E48 3.292!-07 1.797E 40 '

14 3.500E+M l.410E42 ' l.606E+03 8.lB3E-09 1.860E48 3.0l!E 07 1.93:E40 '

15 3.750E 4 V 1.260E42 ' l.720E+03 7.312E-09 ..l.662E-03 2.694E47 2.072E+00 '

16 4.M0E40 ' l.0ECE42 ' l.832E+03 6.266E4V 1.42:E-08 2.310E47 2.20!!+M -

17 4.25CE+00' 8.7ME-03'- 1.944E+03 5.047E 09 1.146E48 1.860E47 2.342E4 0 '

18 4.500E+M ' 6.300E43 2.05 E*03 3.656E-0? 8.311E49 1.347E 07 2.476E+00 ' .

19 4.750E40 3.9ME43' 2.164E+03 2.263E-09 5.14:E-09 8.3+0E4B 2.60BE*00 '

20 5.0ME 40 ' !.700E-03' 2.274E+03 9.866E-10 2.243E-09 3.63!E48 2.740E40' 21 5.250E+00 ' 3.0ME 04 - 2.382E4/ 1.741E-10 + 3.959E-10 ! .415!

6 09' 2.870E40 Totals: 5.754E47 " 1.303E46 't.21!E45 ~

e9

. t i ,, - c :a m. .. . ~ . . r. c ~.2 >

-- - - - ~ TENNESSEE VALLEY AUTHORITY S n e, el W -

SUBJECT Post-LOCA Location 90ecific Eiet a Doees for PROJECT SON Containment Isolation Valves SQNAPS3-091 RO conevene av o v.

q,g, caeciso sv 4 e.,e gg Once the dose multiplier is known, then the internal dose to any point in tne valve can be calculated by multipling the multiolier by the containment free field dose as in equation 3. This is done for the three valve si: es s

$,h)72hb for 8 " va l ves D.ev.. in, s

= 5.8E-7'* 4.7E+8 rad = 2.7E+2 rad,

...t

/ ,

g,p"/p for 12" valves D ,.. .rs...t = 1.3E-6

4.7E+8 rad = 6.1E+2' rad, 3< '

/

for 24" valves Dee,.,s.ren.6 = 1.2E-5 '* 4.7E+8 rad = 5.6E+3 rad. frX8"#I, '

%A7g h [yf;t. k External Sources The valves are exposed to e::ternal sources as well as internal sources. however the valve internals are shielded by the steel pipe walls. The 8" pipes have the thinnest pipe walls (0.022") of the lines considered, thus

./

0.022 inches = 0.8179 cm Using the density of iron (7.674 gm/cm2 Ref. 4). the den-sity thickness is found to be:

0.8179 cm

7.974 gm/cc = 6.44 gm of Fe/cm2' Only botas /

with energies greater t' n 10.0 MeV can pass through this amount of iron (Ref, . Since the maximum beta energy is 5.20 MeV. the pipe ra al l ts provide enough shielding to protect the internal components of the valves from the post accident external beta source. The valveu on the 8" lines are protected from external betas. thus the valves on larger lines (and thicker pipe walls) will be pro-tected as well /

Resultn The highest calculated beta doses are received by the 24" valvet, in containment. These valves receive 5.6C+3 rad 1 due to internal sources only./ 4

1 4

1

. . . . - , . ~. . . - .....L-

.J j

g.__._--.- - - - -

TENNESSEE VALLEY AUTHORITY S ... ~ 8 o, 9

. ~

' ' SUBJECT Post-LOCA Locati on 90ec4. fi c Beta Doses f or PROJECT SON Containment Isolation Valves SQNAPS3-091 R9 p ....

, . u _ ,, _ .. .. a w References '

1. OIR T1EB SON 88032 RO. RIl1 Git B44 880518 823.

2. TI-RPS-48 R2. "Integrated Accident Dose Inside Primary Containment". RIt154 B45 551105 205. page 3.1.

3. SGNAPS3-041 RO. "Beta Dose to Rubber Expansion ;

Joint". RIMS # B45 870123 235 (Attachment 1). l l

4 "Stopping Powers and Ranges of Electrons and i Positrons", IJational Bureau of Standards, NBSIR 82-2550.

5. TVA Drawings 47E235-44 R3 47E235-45 R3 47E235-47 R3 47E205-48 R3.

6. Crane, Flow of Fluids. Technical Paper No. 410, 1976.

7. GENNAL3-013 RO, "Beta Dose Reduction From Finite Vol ume" . RIMSN B45 860624 235.

8. TVA Drawings 47W915-2 R14 47W915-3 R31 47W915-4 R14 47W915-6 R22.

l

.. _. a.... . ~. . . .a . . A

l -

bQ RA PSS- oq l .

y/q

'A';% chi m f-f SHEET _ V fF +4"

{ sufJECT Beta Dose to"Rubber Ettcansien Joint PROJECT SON OC N NE'; 8%t d( ) 60\)AP61-Od1 A TTAC HME AW 1 !

CCFfUTED BY :/f2. DATE 7 4.8/ C}iECKED BY M < O '

DATElj7/p,*f

, . Table 1 Beta Socctra for ,iHerent D Times

1 min 1 day ~0 days 1 vear .

Group Energy (MeV) Prob Prob Prob Prob 1 0.053 - 0.25 0.4859 0.4692 0.7585 0.7297 2 0.25 - 0.50 0.2337 0.1637 0.1187 0.1579 3 0,50 - 0.75 O.0685 . O.0580 0.0244 0.0303 4 0.75 - 1.00 0.0462 0.0379 0.0135 0.0114 5 1.00 - 1.25 0.0375 0.0317 0.0101 0.0086 6 1.25 - 1.50 0.0320 0.0002 0,.0094 0.0080 .

7 1.50 - 1.75 0.0272 0.0290 0.0090 0.0076 8 1.75 - 2.00 0.0200 0.0268" 0.0084 0,0071 9 2.00 - 2.25 0.0164 0.0239 0.0075 0.0063 10 2.25 - 2.5C 0.0111 0.0009 0.0065 0.0055 11 2.50 - 2.75 0.0067 0.0184 0.0057 0.0048 12 2.75 - 3.00 0.0039 0.0166 0.0052' O.0044 13 3.00 - 3.25 0.0028 0.0154 0.0048 0.0041 14 3.25 - 3.50 - _ 0.0020 0.0141 0.0044 0.0037 15 3.50 - 3.75 0.0015 0.0126 0.0039 0.0033

_ 16 3.75 - 4.00 0.CLO10 0.0108 0.0004 0.0029 17 4.00 - 4.25 0.0007 -

0.0087 0.0027 0.0023 18 4.25 - 4.50 0.0004 0.0063 0.0000 0.0017 19 4.50 - 4.75 0.0002 0.0039 0.0012 0.0010 20 4.75 - 5.00 0.0001 0.0017 0.0005 0.0005 -

21 5.00 - 5.25 0.0000 0.000~ Q.0001 0.0001 22 5.25 - 5.50 0.0000 .. 0.0000 0.0000 0,0000 -

O

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l' .

1 B44 '880616 800 QUALIFICATION AGING ANALYSIS REPORT

( TRANSMITTAL DATE: June 13, 1988 TO: 14 E. Daniels g-FROM: Thomas R. Witmer DINDER NUMBER:

gg AGING ANALYSIS REPORT NO.: AA-18 using The attached Material Aging Calculation Report the System'1000. was performed and verified computer software programs were compiled,The written System and 1000 material a Program and in c.7eordance with Digital Engineering Quality Assurance to be accer able for TVA use by Thomas R.'Jperating Procedures and were ve the Envirt Witmer's memorandum to emental Qualification Project files dated September 25, 1985 (B7fd50925013).

Performed By: /20% -

Dates aan Verified By: IMA Date: _, h .

j 4 .

FORM NUMBER: ED TF-005-85 (3-86)

L .

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.. . .- . ~a

_ __~ . -

1 i

B44 '88 0616 80 0 ENVIRONMENTAL QUALIFICATION PROJECT CALCULATIONS s

PERFORMED BY THE SYSTEM 1000 All are material aging calculations performed using the System 1000 based on the Arrhenius Model. Derivations of the Arrhenium Equations, along with definitions of the various parameters are Expected Life in (life) =

(Ea/k )(1/T) + Constant '

(1)

. B Qualified Life (Sinole Service Temperature) t = t /exp((Ea/k )(1/T - 1/T )) (2) 2 1 B 1 2 Qua.lified Life (Multiple Service Temperatures) n+1

(' t 2

= t / I P exp((Ea/k )(1/T - 1/T )) (3) 1 x=2 x B 1 x Deoradation Equivalency Analysis n

t =

E t /exp((Ea/k )(1/T - 1/T ))

A y=1 y (4)

B y A

vheret in (life) = Expected Life (hours)

Ea/k B

= Slope (Activation energy /Boltzmans' Constant)

T = Temperature (Degrees Kelvin)

Constant = Intercept t

2

= Qualified Life (houro) t = Aging Time (hours) 1 e

1 i

'~

- __ , - __ _, , ,__ ._. --. . _ - - . ' ' L -l

_)

exp =

B44 exponent to base e '88 0616 8 0 0 T =

Aging temperature (K) 1 7 =

Service temperature (K) 2 P

X

=

Fraction of 40-year life at T X

T =

Service Temperatures (K) -

X t = Equivalent Time at T A

(houra)

A t =

Time at Temperature T (hours)

Y Y T = Accident Test Temperature (s) (K)

Y T = Baseline Temperature (K)

A i

W e

I

\ .

, . . .~.

B44 '880616 800 Qualification Aging Analysis Report AA-18 problem The 64A form Mark Bur:ynski to Rick Daniels dated April 19, 1988 requested the Aging Section (E.L. Daugherty) to provide an environmental material degradation analysis for the soft seat material in forty purge system containment isolation valves in System 30 (See Appendix A). This request is intended to aid SGN Regulatory Licensing Staff as they pursue a change to the Technical Specification which would relax the test frequency for the subject valves.

The valses in question are Henry pratt Mark II butterfly valves and are identified by Appendix-B. The valve seats need to maintain their integrity following a worse case accident condition which requires containment isolation. These seats are to remain functional for a minimum of 100 days after the accident condition. The valves are located in a harsh environment (Upper Compartment, Lower Compartment-Accumulator Rooms 1, 3&4, Annulus and Instrument Room). The solenoid valves and the limit switch for these valves are qualified under the Electrical Equipment Qua.'fication program. The remaining items of these valves (incluoxng the soft seat) are not covered under any type of Environmental program.

pratt drawings E-3346 and E-3348 describe the nonmetallic material for the soft seat as Resiloseal "ll" (EpT). Ms. Anita Reichling of Henry pratt Co., Marketing Services further defined the material as EpDM with a temperature rating of 300 degrees F (See Appendix C).

The material properties for the EpDM used in this evaluation are defined on Watts Bar Mechanical Equipment Qualiffmation Material Type Listing sheet for Material Type ID M-3L (See Appendix D).

Assumotions In order to perform this calculation, several assumptions were mades (1) Without knowinD the location specific radiation dosos, the published information on the Environmental Drawings was used.

(2) The only material that was requested to be evaluated was the Rosiloscal "W". No other material was evaluated.

(3) The DETA radiation in to be considered as the same as the GAMMA for material degradation analysis.

f ..

k-w$

Quclification Aging Analycio Rsport AA-18 B44 '880616 800 [

b Quality Assurance h1.

The software for this calcu*ation report was performed using MEB's MEQ Computer Program and the Digital Engineering's System 1000 computer' system. The MEQ program and System 1000 computer software programs vere compiled, written and verified in accordance with Digital Engineering Quality Assurance Program and Operating Procedures and were verified and determined to be acceptable for TVA use by Thomas R. Witmer's memorandum to the Environmental Qualification Project files dated September 25, 1985 (B70850925013). The MEQ program requires all input data to -

be prepared, checked, and verified prior to issuance of the final program document. The Aging Section within the Mechanical Engineering Branch has prepared and checked the Watts Bar input data, but has not verified the data within the Aging Section's Quality Assurance Program.

Overview of Mechanical Ecuipment Qualification Proaram The MEQ (Mechanical Equipment Qualification) program provides a comprehensive program for the implementation of equipment qualification analysis and an integrated system that facilitates storing, manipulating, and reporting of components, subcomponent, material, and environmental information. The program was developed to provide many printouts to give such

(- information as the equipment description, subcomponent description, environmental definition, and analysis information.

The program also serves as an library source for documentation of mechanical components and subcomponent part breakdown, and Digital Engineering's System 1000.

In addition to the function of providing pri.itouts sorted on a number of fields, the main purpose of the MEQ program is to analyze each component (including all subcomponents) against each appropriate environment. This analysis reviews each subcomponent against the most conservative activation energy for that materia"1 (as found in System 1000) and performs an expected life calculation based on the required service temperatures. It then performs an -

Accident Equivalency calculation relating the subcomponent to the environment during a postulated accident. The Accident Equivalency la then subtracted from the Expected Life and provides a Thermal Life of the subcomponent. A Radiation Life is then calculated. The Radiation Life is determined by subtracting the postulated accident dose from the material's radiation damage level and then dividing by the yearly normal operating dose rate.

A Service Life is determined for each subcomponent by taking the :

lesser of the Thermn1 Life and Radiation Life. This Service Life is added to the installation date and the program provides a replacement date for each subcomponent. The program will also

.. provide summary information containing the limiting service life for each component.

(

i

B44 Qualification Aging Analycio Rap 3rt AA-18 '88 0616 8 0 0 r The program is also capable of providing postulated accident

( conditions comparison to several components or a defined combination of components.

Methodolooy The above described problem was analyzed'by using the Aging Sections' System 1000 computer software programs and Watts Bar Mechanical Equipment Qualification program which is in accordance with NEP 5.10.

This analysis was divided into four Sections.

The first section (Section I - Thermal Analysis) uses expected life with multi-service temperatures, and accident degradation equivalency calculations. This analysis evaluates the material (using material aging data found in System 1000 for Appendix D - WBN MEQ Material Type Listing Sheet M J' 8) and the environmental temperatures on the respective environmental drawing. This analysis, using Arrhenius methodology, determines the expected life and the accident life. The thermal life is determined life.

by subtracting the accident life from the expected The second section (Section II Radiation Analysis)

( determines the effects of the radiation. The radiation life calculated in accordance with the methodology in NEP-5.10, Appendix B, Attachment B-2, is Page B-40. The radiation life is datermined by subtracting the postulated accident dose from the material's radiation damage level and then dividing by the yearly normal operating dose rate. This analysis takes into account the full BETA and GAMMA radiation values as shown on the environmental drawings.

l The third section (Section III Vender Manual Review)

! documents the results of a review of the Henry Pratt Installation and Service Manual for Nuclear Class Valves for suggested, recommended, and required maintenance for the subject valves.

l Thio manual is in the 73C38-83573 contract documentation.

l Th'e forth section (Section IV - Radiation Analysis, Location Specific BETA Dose) determines the effects of the radiation uaing the location specific BETA radiation doses as determined by HTB Radiation Protection Section. The methodology used in this section to determine the radiation 11f'. is the same as that in Section II.

Resulte The analysea have not evaluated any other non-metallic items f

(,

in this valve other than the soft seat. No other qualification documentation is known to exist for other subcomponento in these valves. HovcVer, the soft conto are the only subcomponento that

,sre requented desired to be analyzed, t j 1 i

B44 '880616 800 Qualification Aging Anclycio Rcptrt AA-18 ,

a Table 1 provides a cross-index of Valve ID to Mark h'o. to Pratt Item No. to Pratt Drawing No.

The environmental conditions are summarized in Table 2.

The results of the analyses performed in Sections I and'II are summarized in Tab *e 3.

The Post-LOCA location specific BETA doses for the valves located inside containment and the Instrument Room are summarized in Table IV. ,

The results of the analyses using the Post-LOCA location specific BETA doses are summarized in Table V.

The following is the results of the analyses performed:

Section I The results of the thermal analysis (as shown in Appendix E) reveals the thermal life of the soft mest installed in the Upper Compartment, Lower Compartment-Accumulator Rooms 1, 3&4, Annulus and Instrument Room to be greater than 40 years and sufficient for its installed application.

( Section II The results of the radiation analyses are shown in Appendix F and summarized below:

A.

The calculated in the Annulus is greater radiation life of the soft seat installed than 40 years.

B.

Compartment, The calculated radiation life for the valves in the Upper Lower Compartment-Accumulator Rooms 1, 3 & 4 and Instrument Room was determined to be insufficient to withstand a'n accident and perform its' inteno'ed safety function and provides no radiation life.

Section III The results of the review of the Henry Pratt Installation and Service Manual are summarized below:

A. Apperdix G is the section out of the manual that deals '

primarily with Nuclear Rubber Seat Butterfly Valves.

B. Page 0-5 of Appendix 0 deals specifically with the rubber seat.

(L This section states that Res11oseal W has an excellent service life. However, not enough data has been coJ1ected to accurately predict its expected life. ,

C. It is known that higher radiation levels and temperature 9 l

- __ = , _ - _ _

B44 '880616 800 Quolificction Aging Anolycio R:psrt AA-10 shorten the life expectancy of the seav.

D. Failure mode is characterized by a shrinking along with a hardening of the rubber and eventual cracking.

E. Henry Pratt recommends the following semi-annual checks:

1. A visual inspection of the seat be made to check for dimensional shrinking and cracking.

2. A Durometer reading be taken. (If a reading of over 75 is indicated, we recommend that the seat

be replaced.)

The normal operating plant conditions do not severely impact th'e life of the soft seats. The Res11oseal W seats would function properly during normal plant operation. Plant test fraquency should be at a minimum at that specified in the Henry F. tt Installation and Service Manual for Nuclear Class Valvas.

However, based on the material evaluation and acceptance criteria set forth in NEP 5.10, assumption (3), and on the known radiation data as described on the SQN environmental drawings, the use of the soft seat valves in the Upper compartment, Lower Compartment-Accumulator Rooms 1, not acceptable.

3, & 4, and Instrument Room is '

This is due to the fact that the seats can not C perform their intended safety function for the 100 day time period following an accident. This failure is due to the high BETA dose during an accident, treating BETA degradation equal to GAMMA and the criteria in HEP 5.10 that a 50% property change is considered material failure.

Sectisn IV Due to the pottintial impact of the results of the analyses in Section III on the operation of Sequoyah HP, it was decided te write OIRMEBSQN88032 (see Appendix H) to determine the actual

post-LQCf_ location specific BETA dose for each valve. The decisjon to write the QIR was made after verbal assurances were given that the BETA doses could be reduced by a factor 10egL QIRNTBSQN88193 (see Appendix I) responded to the request for location specific BETA doses. The doses are summarized in Table IV.

The results of the radiation analyses are shown in Appendix J and summarized below:

A.

The calculated radiation life of the soft seats installed in the Upper Compartment, Lower Compartment-Accumulator Rooma 1, 3 & 4 and Instrument Room is greater than 40 years.

(_ :

d

. _ _ _ _ ~ .

Quclificction Aging Analyc1o Raport 1 Summary C Based on the thermal analyses and the inclusion of the post-

}

LOCA location opecific BETA doses in the radiation analyses, the Res11oseal W seats will function properly during normal plant operation and post accident operating conditions. The soft seats of Res11oseal W in the subject valves are considered to be environmentally qualified for their intended application.

Plant test frequency should be, at a sinimum, at that specified in the Henry Pratt Installation and Service Manual for Nuclear Class Valves to insure the ability of the soft seats to -

provide an adequate seal.

This evaluation in no way takes into consideration any detrimental effects of wear on the soft seats. This must be a consideration in the choosing the test frequency.

The results of this evaluation will be formally documented in a calculation when the BETA doses are formalized in calculation SQNAPS3-091, as stated in QIRNTBSQN88193.

0

(_

B44 QualificOtien Aging An21yc10 rop 3rt AA-18 '880616 800 C

TABLE 1 Yalve Size Pratt ID NO. Mark No. (in.) Item No. Pratt 3revino No.

FCV-30-7 47W915-54 24 1 E-3344 E-3346 FCV-30-8 47W915-64 24 5 E-3345 E-3348 -

FCV-30-9 47W915-55 24 1 E-3344 E-3346 FCV-30-10 47W915-65 24 5 E-3345 E-3348 FCV-30-14 47W915-36 24 1 E-3344 E-3346 FCV-30-15 47W915-66 24 5 E-3345 E-3348 FCV-30-16 47W915-57 24 1 E-3344 E-3346 FCV-30-17 47W915-67 24 5 E-3345 E-3348 FCV-30-19 47W915-59 12 2 E-3344 E-3346 FCV-30-20 47W915-69 12 6 E-3345 E-3348 FCV-30-50 47W915-62 24 5 E-3345 E-3348 .

FCV-30-51 47W915-52 24 1 E-3344 E-3346 FCV-30-52 47W915-63 24 5 E-3345 E-3348 FCV-30-53 47W915-53 24 1 E-3344 E-3346 FCV-30-37 47W915-60 8 3 E-3344 E-3346 FCV-30-40 47W915-70 8 7 E-3345 E-3348

( FCV-30-56 FCV-30-57 47W915-61 47W915-51 24 24 5

1 E-3345 E-3344 E-3348 E-3346 FCV-30-58 47W915-68 12 6 E-3345 E-3348 FCV-30-59 47W915-58 12 2 E-3344 E-3346 O

e f

4

B44 Cu211tication Aging An31yalo Report AA-18 '880616 800 ,

TABLE 2 GAMMA MAX.

BETA GAMMA 40 YR NORMAL /

Valve TVA ID #

Accident Accident INTEG M4X. Environment

_ Room Dose Dome DOSE ABNORMAL Drawina -

8 U-C 4.7 E8 3.80 E7 10 U-C 1.0 E6 110/120F 47E235-44 R3 4.7 ES 3.80 E7 1.0 E6 110/120F 47E235-44 R3 56 C-AC1 4.7 E8 1.00 E7 2.0 E7 120/130F 47E235-45 R3 17 C-AC3 4.7 E8 1.00 E7 52 C-AC3 2.0 E7 120/130F 47E235-45 R3 4.7 E8 1.00 E7 2.0 E7 120/130F 47E235-45 R3 15 C-AC4 4.7 EB 1.00 E7 40 C-AC4 2.0 E7 120/130F 47E235-45 R3 4.7 E8 1.00 E7 2.0 E7 120/130F 50 C-AC4 47E235-45 El 4.7 28 1.00 E7 2.0 E7 120/130F 47E235-45 R3 7 AHN 5.0 E5 1.25 E7 1.0 E6 110/120F

( 14 16 9 AHH AHH AHH 5.0 5.0 5.0 E5 E5 55 1.25 E7 1.25 1.25 E7 E7 1.0 1.0 E6 E6 110/120F 110/120F 47E235-47 R3 47E235-47 R3 47E235-47 R3 19 ANN 1.0 E6 110/120F 47E235-47 R3 5.0 E5 1.25 E7 1.0 E6 110/120F 37 AHH 5.0 E5 47E235-47 R3 1.25 E7 1.0 E6 110/120F 47E235-47 R3 51 AHH 5.0 E5 1.25 E7 53 1.0 E6 110/1207 47E235-47 R3 AHH 5.0 E5 1.25 E7 1.0 E6 110/120F 47E235-47 R3 57 AHH 5.0 E5 1.25 E7 59 1.0 E6 110/120F 47E235-47 R3 AHH 5.0 ES 1.25 E7 1.0 E6 110/120F 47E235-47 R3 20 C-IIR 4.7 E8 1.00 E7 3.5 ES 75/120F 58 C-IIR 4.7 EB 47E235-48 R3 1.00 E7 3.5 E5 75/120F 47E235-48 R3 (NOTE - The postulated accident temperature profile ic included in the thermal analysis by use of the Accident Degradation Equivalency calculation) i r

(

f (

=D J

i B44 '880616 800 I Qualificction Aging Analycio Raport AA-18 j

.s I

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l TABLE 3 The. mal Life Radiation Life Room (Years) (Years)

U-C 14368.57 -18320 -

C-AC1 6285.18 -860 C-AC3 6285.18 -860 C-AC4 6285.18 -860 ANN 14368.32 1480 C-IIR 220812.45 -49142.9 l

i L !

4

844 '880616 80(

Qualificatien Aging An 1ysic Rsport AA-18 C

TABLE 4 Post-LOCA Location Specific BETA Domes BETA Valve Accident ID # Room Dose .

8 U-C 6.4 E3 10 U-C 6.4 E3 56 C-AC1 6.4 E3 17 C-AC3 6.4 E3 52 C-AC3 6.4 E3 15 C-AC4 6.4 E3 40 C-AC4 7.1 E2 50 C-AC4 6.4 E3 20 C-IIR 1.6 E3 58 C-IIR 1.6 E3 t

l l

e l

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, . . :. : . . 4

I B44 '880616 800 Quc11ficction Aging Analycic R p3rt AA-10 ..

l TABLE 5 Post-LOCA Location Specific BETA Doses Valve Radiation Life ID # Room (Years) -

8 U-C 479.74 10 U-C 479.74 56 C-AC1 79.99 17 C-AC3 79.99 62 C-AC3 79.99 15 C-AC4 79.99 40 C-AC4 80 50 C-AC4 79.99 20 C-IIR 4571.25

( 58 C-IIR 4571.25 e

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Qualification Aging Analycio Raport AA-10 B44 '880616 800 g

sA APPENDICES Appendix A 4/29/88 Telecopy from Don Goodin to Tom Witmer forwarding Mark Burzynski's 4/19/88 45D to Rick Daniels B 4/22/88 Telecopy from Bob Poole to Tom Witmer listing valve ids, location, function and 50.49 categories and operating times C C. A. Chandley's letter to Henry Pratt dated January 29,

1987 (244 87 0303 503)

D WBN MEQ Mhterial Type Listing sheet for Material Type ID M-3ti E System 1000 Calculations for Degradation Equivalency and Expected / Service Life F System 1000 Calculations for Radiation Life G Henry Pratt Installation and Service Manual for Nuclear Class Valves, Pages 0-1 thru O-14 H QIRMEBSQN8dO32,RO (B44 88 0518 823)

GIRNTBSQB88193 (B45 88 0609 262)

I J System 1000 Calculations for Radiation Life which incorporate GIRNTBSQN88193

(

4 s'

?

J

B44 '880616 800 Qualification Aging An31yo13 Rspart AA-18 e

APPENDIX A C

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NU?.S E.T. OF FULLOWING PAGES: . !

TH:: u.ESSAGE IS BEING SENT FRCP.: (6*5) 870-7139 PANAFAX UF-600 THIS NUMBER LICENSING ONLY FOR USE BY SITE D SFE :.'L INSTRUCTIONS: --

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B44 '880616 800 Quclification Aging Analyoic Ropsrt AA-18 G

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0J011fication Aging Anclyaic !!:p3rf AA-18 B44 '880616 800 4

APPENDIX C C

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B44 '88 0616 8 0 0

I'ENN ECC EE VALLEY AUTH3MITY KNOXVILLE. TENNESSEE 37902 E Cii V E P' V7 C126, 400 West Sumnit Hill Drive hh g3) g//

{ January 29, 1987 s -

M RkETING 5TMG

Henry Pratt Company 401 South Highland Avenue Aurora. Illinois 60507 B44 '87 0303 503 :

Attention: Anita Reich 11ng

, s

Gentlemen: '

Thank yo'u fow the information you supplied TVA. From my notes, the telephone documentation on the following page is an accurate account of our

~ conversation. If.you agree, please sigri and date the spaces provided at the ,',J bottom of each page of this letter and return it Jn the envelope provided. .D If there should be any discrepancies.' please contact John Sams at

~;,

(615) 632-4626. A quick response would be greatly appreciated.

,* i ,

Very truly yours. '

TENNESSEE VALLET AUTHORITY 44 '

f 0 n' '

R hC.A.Chandley,ChiefMechanicalEngineer

_ MES MAR 3 ~1987 CAC (SEG HRC.

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- EU AE_r1 1983-TVA SOTH ANNIVERSARY '

An Equal Opportunity Employer

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B44 '88 0616 8 0 0 ; ,

DOCUMENTION OF ,

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. S' TELEPHONE CONVERSATIONS .s F., A' i ' M

.s DATE: January, 26, 1987

PARTY CALLING: John A. Dams PARTY ANSWERING: Ani,ta Reichling .

COMPANY: Henry Pratt Company s ..'

ADDRESS: 401 South Highland Avenue Aurora,IL 60507 .

- - PHONE NUMBER: (312) 844-4000 x *- '.

. ., . . ....,i,-

Temperature rating of R'esilossal C. Temperature ratini

SUBJECT:

, (,

and material composition of Resiloseal W. (M-38 & 95) ..i.- -

...................................................................... gi TELECON NOTES (Include questions, decisions, and comments) .

0. ) What is 'the temperature rating of Restloseal C7 ,

A. ) *r.

The temperature rating of Resiloseal C iw 200 degrees F. Ji

Q. ) What in theW7 mater'ial composition and temperature rating of the Resiloseal ' i A. ) The Res11oneal W is made from EpDM and has a ter,tprrrature

t'ating of 300 decrees F.

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B44 '880616 800 Cuclific ticn Aging AnolyOic R port AA-18 3" 7

h APPENDIX D

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EDmCa.EwExi Mrbm'88 0 616 8 0 0 MTERIAL TYE LISTIE Natorial Type ID: M-38 Nate*ialType: EPDR Trade Names ES! LOSE 4. W Ref. No.

Threshold: 1 K6 MOS 51146 Activation Erergy: 1.31 eV 20643A Teeperature Rating: 148C M-36-1 Dwge Level 495 DMGE at 5.IE7 MDS 31744 *

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k. sial used in EQ ID:

171, 172, 359, 364, 3L3, 364, 365, Et, 634, 635, 636, LT7, 638, 639 Subcesponent Listing Sheet ho.:

6,7,7A,11,51 Referefces:

LIBERYCODE 4 .. 511-86 DOCUOff TIM.....MD!AT!W DATA FOR DES!GN/QUR.lFICAfim 0F lt0ER IUNT ERJIPEXT -

41 THOR. . . . . . . . . . . . . F. 30UOLET, J. WINSLOW FIlJR2. . . . . . . . . . . . . EFR I DOC 1)DT K). . . . . . . 1717 7 LIBERY CDDE E...ite-83A DOCUOG TITLI.....NDRXI. EN61EERING PROPERTIES Ape APR.!CATICNS C AJT@R.............M/A EX)R2. . . . . . . . . . . . . DUPmi DOCLKXT E.......E 13193 LIDAARY C00C M)....R-381 D0370(f TITLE.....Brl caTT- CDiTACT EPORT TO mlTA E10t!E m 1/26/87 m MT'L CDIP. 8 TDG MTIE RfDDt. . . . . . . . . . . . . J0W ' SMS 50UR2. . . . . . . . . . . . . & $TT CDGM D0C200 E...... 84487aism LIBmRY CDCC NO... 31744A 80CLIOT TIM.....CDGILAf!m 0F AADIATim DWGE TEST DATA, PART 1: CABLE IGLATIE MTER!4.S AID 0t.............H. SOOGAOER MD A. STCLARI-llYCXA F(K2.............EW, EA.TH PC SWITT DIVISim LO1K.NT E.......Em 7144 9

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Revi M % M d d D.t. ' L 4 J /8/

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B44 '880616 800 Cuolificatien Aging Analycio R; port AA-18 1

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APPENDIX E C

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B44 '88 0616 8 0 0 .

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BASE TEMPERATURE (F): 110.00 OPERABILITY (DAYS): 100.00 ACTIVATION ENERGY (EV): 1.31 TEMPERATURE (F) DURATION (SECONDS) 102.C3 2300.00 160.00 850 00 147.00 25800.00 130.00 170000.00 117.00 400000.00 109.00 400000.00 105.00 1592000.00 100.00 6048000.00 .

OPERABILITY VS ACCIDENT: .00 SECONDS

.00 DAYS EQUIVALENT TIME: 72.61 DAYS

( 1694.67 HOURS

.19 YEARS ROOM U-C AA-18 1

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m _

I B44 '88 0616 8 0 0 1 i

BASE TEMPERATURE (F): 120.00 OPERABILITY (DAYS): 100.00 i

ACTIVATION ENERGY (EV): 1.31 .

TEMPERATURE (F) DURATION (SECONDS) 180.80 60. Se 199.30 {

689.40 259.20 949.90 !

200.00 28299.80 l 175.00 170000.00 150.00 1000000.00 125.00 1392000.00 115.00 6040000.00 OPERABILITY VS ACCIDENT: .00 SECONDS l

.00 DAYS ,

EQUIVALENT TIME: 513.64 DAYS

( 12327.42 HOURS 1.41 YEARS ROOM: C-AC1, C-AC3 & C-AC4 An-10 2

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B44 '880616 800

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BASE TEMPERATURE (F): 110.00 OPERABILITY (DAYS): 100.00 ,

ACTIVATION ENERGY (EV): 1.31 TEMPERATURE (F) DURATION (SECONDS) 133.68 4500.00 131.14 95500.00 126.43 159200.00 125.45 772500.00 121.59 772500.00 117.73 772500.00 113.86 772500.00 110.00 5290800.00 OPERABILITY VS ACCIDENT: .00 SECONDS

.00 DAYS EQUIVALENT TIME:

( 159.38 DAYS 3825.10 HOURS

.44 YEARS

, ROOM: ANN AA-18 3

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B44 '880616 800 (I

BASE TEMPERATURE (F): 75.00 OPERABILITY (DAYS): 100.00 ,

ACTIVATIUN ENERGY (EV): 1.31 TEMPERATURE (F) DURATION (SECONDS) 180.80 60.90 199.30 689.40 259.20 949.90 200.00 28299.80 175.00 170000.00 150.00 1000000.00 125.00 1392000.00 115.00 6048000.00 OPERABILITY VS ACCIDENT: .00 SECONDS

.00 DAYS

( EQUIVALENT TIME: 27406.20 DAYS 657748.84 HOURS 75.09 YEARS ROOM: C-!!R AA-18 4

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B44 '88 0616 8 0 0 SUPERCA'.C CROGRAM: EXPMULTI V1.1 6/9/87 EP THIS PRLCRAM IS DESIGNED TO PERFORM EXPECTED LIFE CALCULATIONS FOR MUL1IPLE SERVICE TEMPERATURES SLOPE ..................... 15150.732372901 C INTERCEPT ................. -29.23178619665 SERVICE TEMPERATURE (C) #1. 43.33 PERCENT OF TIME AT #1 ..... 99.00 SERVICE TEMPERATURE (C) #2. 48.89 PERCENT OF TIME AT #2 ..... 1.00 t

SERVICE TEMPERATURE (C) #3. .00 PERCENT OF TIME AT #3 ..... .00 SERVICE TEMPERATURE (C) #4. .00 PERCENT OF TIME AT #4 ..... .00 SERVICE TEMPERATURE (C) #5. .00 *

, PERCENT OF TIME AT #5 ..... . t'0 EXPECTED LIFE ............. 14368.76 <

ACCIDENT EQUIVALENCY ...... .19 d

e SERVICE LIFE .............. 14368.57 ROOM: U-C -

AA-18 5

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. . _ _ . _ - . l B44 8 06" SUPERCALC PROGR2M EXPMULTI V1.1 G'd /87 E#. 6 8 0 0 i THIS PROGRAM !S DESIGNED TO PERFORM EXPECTED LIFE CA.LCULATIONS FOR MULTIPLE SERVICE TEMPERATURES

( SLOPE .....................

INTERCEPT .................

15150.732372901

-29.23178619665 SERVICE TEMPERATURE (C) #1. 48.89 PERCENT OF TIME AT #1 ..... 99.00 SERVICE TEMPERATURE (C) #2. 54.44 PERCENT OF TIME AT #2 ..... 1.00 SERVICE TEMPERATURE (C) #3. .00 PERCENT OF TIME AT #3 ..... .00 STRV!CE TEMPERATURE (C) #4. .00 PERCENT OF TIME AT G4 ..... .00 ,

SERVICE TEMPERATURE (C) #5. .00 PERCENT OF TIME AT #5 ..... .00 EXPECTED LIFE ............. 6286.59 ACCIDENT EQUIVALENCY ...... 1.41 SERVICE LIFE .............. 6285.18 ROOM: C-AC1, C-AC3, & C-AC4 -

AA-18 6

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844 '88 0 6 P6- 8 0 0 SUPERCALC PROGRAM: EXPMULTI V1.1 6/9/87 EP THIS PROGRAM IS DESIGNED TO PERFORM EXPECTED LIFE CALCULATIONS FOR MULTIPLE SERVICE TEMPERATURES SLOPE ..................... 15150.732372901 C INTERCEPT ................. -29.23178619665 SERVICE TEMPERATURE (C) #1. 43.33 PERCENT OF TIME AT #1 ..... 99.00 SERVICE TEMPERATURE (C) #2. 48.89 PERCENT OF TIME AT #2 ..... 1.00 SERVICE T2MPERATURE (C) #3. .00 PERCENT OF TIME AT #3 ..... .00 SERVICE TEMPERATURE (C) #4. .00

PERCENT OF TIME AT #4 ..... .00 SERVICE TEMPERATURE (C) #5. .00 PERCENT OF TIME AT #5 ..... .00 EXPECTED LIFE ............. 14368.76 ACCIDENT EQUIVALENCY ...... .44 SERVICE LIFE .............. 14368.32 , ,

ROOMS ANN AA-18 7

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'88 06T6 8 0 0 SUPERCALC PROGRAM: EXPMULTI V1.1 6/9/87 EP THIS PROGRAM IS DESIGNED TO PERFORM EXPECTED LIFE CALCULATIONS FOR MULTIPLE SERVICE TEMPERATURES C SLOPE ..................... 15250.732372901 INTERCEPT ................. -29.23178619665 SERVICE TEMPERATURE (C) #1. 23.89 PERCENT OF TIME AT #1 ..... 99.00 SERVICE TEMPERATURE (C) #2. 48.89 PERCENT OF TIME AT #C ..... 1.00 SERVICE TEMPERATURE (C) #3. .00 PERCENT OF TIME AT #3 ..... .00 SERVICE TEMPERATURE (C) #4. .00 PERCENT OF TIME AT #4 ..... .00 SERVICE TEMPERATURE (C) #5. .00 * '

PERCENT OF TIME AT #5 ..... .00 EXPECTED LIFE ............. 220887.54 ACCIDENT EQUIVALENCY ...... 75.09 SERV!CE LIFE .............. 220812.45 ROOM C-IIR

t AA-18 8

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1 cuotincatien Aging Ano101, 4 ,,1QQ 0 616 8 0 0

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844 '880616 800 THE PURPOSE OF THIS CALCULOTION IS TO DETERMINE THE RADIATION LIFE OF MATERIALS BASED ON RADIATION LIFE METHODOLOGY PER NEP 5.10 MATERIAL RADIATION DAMAGE LEVEL (RADS)......: 5e7 C MATERIAL DAMAGE LEVEL (%).....: 49 MATERIAL PROPERTY............. ELONGATION ACCIDENT RADIATION DOSE (GAMMA).............: 3.8e7 ACCIDENT RADIATION DOSE (BETA)..............: 4.7e8 40 YEAR TID (RADS)..........................: 1e6

_-------~~-----------------

RADIATION LIFE (YEARS)......................: -18320 ENVIRONMENTAL DRAWING NO.: 47E235-44, R3 AA-18

9 C

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B44 '88 0616 8 0 0 THE PURPOSE OF THIS CALCULGTION IS TO DETERMINE THE RZDIGTION LIFE OF MATERIALS BASED ON R2 DICTION LIFE METHODC JGY PER NEP 5.10 MATERIAL RADIATION DAMAGE LEVEL (RADS)......:

C MATERIAL DAMAGE LEVEL (%).....:

MATERIAL PROPERTY............. ELONGATION 49 5e7 ACCIDENT RADIAT!OM DOSE (GAMMA).............: 1e7 ACCIDENT RADIAT!Oh DOSE (BETA)..............: 4.7e8 40 YEAR TID (RADS)..........................:

-__---_.-_--_--__._----_-_-__------__--_ -_--_---__2e7 ___-_____-________-__

RADIATION LIFE (YEARS)......................: -860 ENVIRONMENTAL DRAWING NO. 47E235-45, R3 AA-18

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844 '880616 800 .

THE PURPOSE OF THIS CALCULCTION IS TO DETERMINE THE RDI ATION LIFE OF MATERIALS BASED ON RT,DIQTION LIFE METHODOLOGY PER NEP 5.10 3

.rc. e MATERIAL RADIATION DAMAGE LEVEL (RADS)......: 5e7 il

' C MATERIAL DAMAGE LEVEL (%).....:

MATERIAL PROPERTY............. ELONGATION 49 A ACCIDENT RADIATION DOSE (GAMMA).............: 4 1.25e7

ACCIDENT RADIATION DOSE (BETA)..............: 5e5

40 YEAR T!D (RADS)..........................: 1e6

_____________________________________________.....__....._____._ji RADIATION LIFE (YEARS)......................: 1480 i ENVIRONMENTAL DRAWING NO. 47E235-47, R3 AA-18 ,

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THE B44 '88 6616 8 0 0 PU.~. POSE OF THIS CALCULGTION IS TO DETERMINE THE RIDIATION LIFE OF MATERIALS BASED ON RADICTION LIFE METHODOLOGY PER NEP 5.10 C MATERIAL RADIATION DAMAGE LEVEL (RADS)......: 5e7 MATERIAL DAMAGE LEVEL (%).....: 49 MATERIAL PROPERTY............. ELONGATION h' ACCIDENT RADIATION DOSE (GAMMA)..... .......: 1e7 L ACCIDENT RADIATION DOSE (BETA)..............: 4.7e8 40 YEAR TID (RADS)...... ...................t 3.5e5 '

RADIATION LIFE (YEARS)....................... -49142.9 .

ENVIRONMENTAL DRAWING NO. 47E230-43. R3 AA-18

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B44 '88 b6 [6 x Cuo11ficaticn Aging Analycio R; port AA-18 8 0 0 .fi.

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] i Main Elccents of ...

I j Butterfly Valves T Top Stub Shaft A ' q, .d

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B44 '88 0616 8 0 0 INSTALLATION AND SERVICE MANUAL NUCLEAR RUBBER SEAT BUTTERFLY VALVES Pratt Rubber Seat But t er f ly Valves are designed for long service life and durability under elevated radiation levels for nuclear service. Their construction is extremely rugged, but a reasonable amount of care in installation and handling is suggested to insure the realization of its normally long service life. For proper operation, the NRIA air purge valves must be installed and oriented as shown in Fig. 1.

Although your valve will require little maintenance, it is desirable that its operation and construction be thoroughly understood. This manual has been prepared to assist you in learning all tbout your valve.

The only recommended maintenance operations that are normally required on your valve are ;1) the inspection procedures on the rubber seat, and (2) the lubrication of top a n d b o t t o'a trunnions.

l Although any part of the valve is field r e pl a c e ab le uhe seat is ,

the only component whose replacement would normally be necessary.

However, it is not recommended that seats be stocked since they will age unless stored under ideal conditions.

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HENRY PRATT COMPANY % MATERIAL SPFCIFICATIONS M

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/openAyous woT swowu A A k m me r & & -

i e... ., e.c. s . _ cm ., "" om " *- "" om "1 "- 3' gt h,kChaE D4ir IQN,1LN % APPtoyrn !

! LT I B44 '880616 800 It should be noted that Petroleum based greases cause this seat material to swell while silicone based greases cause

) no effect.

I LUBRICATING top AND BOTTOM TRUNNIONS:

+

The applicati'on of grease to the top and bottom trunnion is not for purposes of lubrication. It is strictly used as a corrosion inhibitor by creating an air barrier and to keep the bearing free from abrasive particles that could get into the area.

J Due to the adverse affects of petroleum grease on the rubber seat and the possibility of elevated temperatures, we require that Dow Corning 111 silicone grease be used to flush the trunnions g once every 12 to 18 months. This type was chosen because of its' ability to withstand high tenperatures and still retain its' E ability to adhere to the working parts without affecting the rubber seat.

I To perform the operation, pump one pint of grease into g the trunnion grease fitting. Repeat this procedure for the opposite trunnion and wipe excess grease from the inside of.the l valve with a clean cloth if possible.

I DISCUSSION ON OPTIONAL MAINTENANCE At per! odic intervals of your discretion, you may want to leak test your Pratt valve. There are two methods that can be used to check this. One is a quick method with questionable accuracy and the other is involved, but has a high degree of

.I accuracy. 4 o

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B44 '880616 800 a

The simple method involves the use of a feeler gauge as a

( probe that can be inserted between the disc and the seat as a 4 P

measure of the amount of interference. While inserting the gauge note the amount of force needed and compare that with the force 7 1

needed to insert and extract it at various intervals around the q disc.

The force involved snould be fairly high and approximately i 1

equal all the say around the valve. If not, there are two ad- j justments that can be made to correct this problem (assuming that '

'the seat is in good condition and does not need replacement). '

4 If the force is greater near one of the disc hubs than it !

is at the other it could be that the thrust bearing needs to be adjusted. If this is the case, follow the instructions enclosed that explain the thrust bearing adjustment procedure. -b ,

If the force varies in only one spot or at intervals 1 C other than previously described, the retaining ring segment next J.

J to the area in question should be tightened down until the force '

involved in that area equals that elsewhere around the disc. If I grease has fallen onto the seat it will make a difference in the pressure needed to insert the feeler gauge so be careful not to make the mistake of assuming that this automatically indicates that the retainer segment '.eeds tightening -- inspect the seat first.

The second method of testing your valve is to employ a halogen leak test by pressuri:ing the area between the valves.

If this method shows excessive leakage, the same methods of correction as described above should be used.

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l B44 '880616 800 DISCUSSION ON MAINTENANCE PROCEDURES <

RUBBER SEAT:

j The rubber seat is made of RESILOSEAL W* which is an ethylene propylene compounded specifically for nuclear service under elevated background radiation and temperature levels.

Although indications'show that RESILOSEAL W has an excellent

~

service life, not enough data has been collected to accurately predict its expected life under varying conditions of use and P

environment. It is known that hicher radiation levels and

_ temperature shorten the life expectancy of the seat. Failure mode is characteri:ed by a shrinking along with a hardening of the rubber and eventual cracking, g For this reason we recommend the following semi-annual checks:

1. A visual inspection of the seat be made to check l(' for dimensional shrinking and cracking.

2. A Duremeter reading be taken. (If a reading of over I 75 is indicated, we recommend that the seat be g replaced.)

These are strictly precautionary measures taken due to the -

I accelerated aging caused by radiation and heat and the fact that relative life expectancy can change from installation to I installation.

I l *Resiloseal is a trademark of Henry Pratt Co., Inc.

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1 B44 '880616 800 i

l

( For information about Pratt Rubber Seat Butterfly Valve 3 j

! not contained in this manual, contact our local Sales Represent-

ative, or write direct to:

3 .

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i HENRY PRATT COMPANY l 401 South Highland Avenue 1

l i Aurora, Illinois 60507 f

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REPLACING RUBBER SEAT IN NUCLEAhNTERhhhIMlE6 800 9 A. DISASSEMBLY B44 '88 0616 8 0 0

1. Turn disc in open position.

f 2. Remove adjusting screws in segments.

3.

Number all retaining segments and mark proper 1ccation on valve bot /. '

4 Pry segments out of body.

l S. Remove old rubber seat. .

6. Clean seat groove with clean dry rag.

B. INSTALLATION

1. Apply. light coat of silicone grease to seat groove.*

2.

Insert new liner seat into place by pushing rubber snug.

3. Insert segments back into body and install screws. ;

4. Close valve disc.

5. Tighten all segment adjusting screws to assure reason-able interference between seat and disc.

6. Bubble or leak test valve and tighten adjusting screws as needed t stop all apparent leaks.

t

We recommend Dow Corning valve seal or Dow Corning III Compound.

Please note that the use of Petroleum grease negates all guarantees.

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B44 '88 0616 8 0 0 3 DOUBLE- ACTD'G THRUST BEARING ADJUSTMENT ',

l

( NUCLEAR RUBBER SEAT BUTTERFLY VALVES 3

l The tight seal of your Pratt Rubber Seat Butterfly Valve is .

I "I

achieved by making the disc diameter somewhat larger than the inside diameter of the rubber liner. To insure equal interference of disc and rubber in the areas adjacent to the disc hubs, an .

adjustable thrust bearing is provided at the lower end of the l . shaft. The thrust bearing has been properly set at the factory, and further adjustments will not usually be necessary. If the valve has been dismantled, or if interference at the disc hubs is not equal (leaky valve), the thrust bearing must be adjusted.

The following procedures should be employed: :

1. Determine the interference between disc and liner adjacent to the dise hubs. Interference will be l .

apparent by the effort required to insert and remove a - :

l feeler gage between the disc and rubber. This inter-ference should be equal near each nub.

2. Loosen set screw next to adjusting screw (see sketch).

If the section of disc nearest the thrust bearing had l the greatest interference, turn thrust screw "in" slightly. If this section had less interference, turn

! thrust screw "out" slightly.

l

3. Tighten the set screw and again check interference I

adjacent to each disc hub. Repeat adjustment procedura until the interferences adjacent to tne opposing dise -

l hubs are as nearly equal as possible. It may be necessary !

to drill and tap a new hole for the set screw, making I sure that there is a full diameter in both the adjusting screw and cover plate (See Sketch),

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B44 ESSENTI AL FEATURES OF

'880616 800 THAUST BEARING ASSEMBLY 4

I VALVE BODY

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7 B44 '88 0616 8 0 0 BUTTERFLY VALVE STRUCTURE, 5'

l For your convenience, we are enclosing a drawing that shows ,j the General Construction of your Pratt Valve and a copy of the g exact Bill of Materials from which each of your valves was built. <

J The part and drawing numbers will be valuable if a replacement {

part should ever be necessary. Although contained in our per- k 3

manent records, receipt of this information would speed the handling of your request. ,

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B44 '88 0616 8 0 0 BUTTERFLY VALVE OPERATOR AND PERIPHERY COMPONENT

( PART INSTRUCTIONS In this section of your maintenance manual you will find all the appropriate manuals for the component parts associated with your Pratt Valves, and any other pertinent data from our l Suppliers. Inquiries may be handled either through your local , ;

Henry Pratt Representative or directly to the Supplier. ,

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4 Inicrmctlan about B44 ,88 0616 8 0 0 h E@ O Q U S[ Qy R..M EO ft'E O aiGG S

- _ . . _ _ . _ _ __._ . . _ . . _ _ . , 9 DOW CORNING

111 Compound N .

Dow Corning ill compound is a stiff, non. melting it is resistant to most aqueous w!utions ofinnrganie silicone lubricant used primarily as a valve lubricant salts and to dilute ocids and alkalics. It is unaffected

'[f and sealer This compound is similar to Dow Corning by most vegetable and mineral oils, many oryanic /1 val've seal and is used in applications where a heavier compounds and most common gases. The tables e n consistency material with more tack is desired. the back of this sheet outline the perfortnance of Dow Corning Ill compound eshibits the following Dow Corning Ill compound in valves handling a .[~,

variety of chemicals at different tempesatures and qualities:

concentrations. :

e serviceable from -40 to $00 F E

cacellent water and oil resistance Dow Corning ill compound will not carbonize at

good rc>istance to most chemicals high temperatures and has a scry low vapor pressure.

low volatility even at 400 F, It will neither melt not harden on long 1

has no swelling effect on rubber or plastic esposure to clevated tempenturet Thus, this com. -

Dow Corning ill compound is insoluNe in watei, pound will keep salves clear and limit contamination j rnethanol, cthanot,. acetone, glycol anil glycerine. of products. .

FDA $!ATUS All of the components of Dow Corning 1iI compound are inc!aded in the list of )

substan:cs set forth in FDA Regulation No, 121.2514 Although the regulation does not Jeal with valve lubricants as 5:.sh,it does relate in polymeric contin;s supplied as a continuous film over a m:tal snbstrate, it would appear that such a q

( use does have a reasonable degree of peitincnee to valve lubiicant,at least insefar as tosicolo;i:al safety is concerned. The user should consider whether fuither FDA approval for use of these prodnets in his particular system or applicationis a

y appropriate. J i 6NDUST t!At USC5 Chemical h uc cuine: P.-cause of chemicai resistance, binceulars and telescopes, and as a dnnpin; mett ium 1i lonc service bre.nnd temperature stabihty, Dow Corn. in hydraulic pisots, as a direct. contact transducer medium in extruders, and as a miniature besiing seal.

ing ill compound has found use ns a vahe, "O"

ring, and plug sahe lubricant. Sir.ce it does not C4"II*#"' M"""I##'"'I^'r: Applications include oxidire, gum. or sotatik:c, it is an effective lubri.

yams operating over a wide temperature range and ,

cant for contre,t valves, piessure plug valves, and as a atrnospheres. It prevents sticking of sealant for vach m and pressure systems. It is also '"*"'"..O, valves and rings at high temperatures or in 'g used as a chemical bariier coating. freeting conditions. Dow Corning lli compound aho q Merhonic al Derice Munnforturing: An excellent acts as an excellent seater between synthetic rubber .

plastic ai.d rubbci lubricant. Dow Corning ill com- or plastic parts in contact with metal in vacuum or

pound is used on devices which requite special con- pressure systems. {s sideration. Finw meter bearings. fire estin;uisher vahes, water softener unt! faucet vahes, and lock Horsli Operating Conditioni: Specialized properties such as low volatility and chemical resistance allow ,

rnechanisms, are esamples of uses dependent upon usage in high vacuum or pressure systems, or in high temperatute and racchanical stabiht). Dow Corning lli compound is nho usal to lubricate "O" rings on temperature atmospheret (Continued on nest pour) Q ,

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Da eMeematiom et tel est eth.aP %.e 5 % %'1%c' twgh'y itgt emp agrht e t .e ^ e8 41 **8stendtoup m.. ..'.. . ... - - ==-- DOW CORNtNG COR POR ATION ttACYMe betetle:10*y esetPP W t (017e te m ' j $AjC H l G A N 4gC40 j SAlD(,A N D, 8 64hJ 4t een $we pagties of sm , g*a 4f eN the (seca j e e, i g%s g. () t. s* e,s es .ai ta it e ' . h e%e 4% b e

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CHEMICAL RE5157ANCE OF DOW CCRNING 111 Compownd The following tables are divided ints inorg:nic chem. TABLE 11 ic:1 service, oiganic chemical service tnd gas service. ORGANIC CHEMICAL SERVICE

""*~"Y " ~~ ** "'T r- *l. .ar ' M.v3gsc3. =- . . g The r.iting or comment m, each case m, dicates the y,,,,,

arparent chemseal resistance of Dow Corning ill rice s, ia r 7 o/ C.

cornpound in contact with these various chemicals Neuer , O reme. ou,. ,,

N in laboratory tetIs or in field use. 4,,ianaa, /,,, o 73 ,,a 4,in p Cy dg Acet N' atut. 73 and .t441l > 6eW (atual These lab!cs should be ticated as guides only, since it Accue ac4.irrIl.itLilaquean CO Rmm iemn ie oao is impractical lo secure or include comple te informatio n '

^ '/i,',7 !*o'[ ll[P l ",$ (*,,"

1 on sariables such as picssure, new velocity.relubrica. Aa.hne od o

$i o 1.$ a,*l anu i Fww cao tion schedules, and vmbe construelion. For this rea. $'p$, d* ' " * ,.4 ". .

F",d $",dJ son, the suitability of Dow Coining t il compound tieniene Romim ie mp. t ah Ping should alnys be tested on a small scale before it in ["'["ly,non,, ,,yg adopted for regular use.

n, ig M 'r$'r ih avi> l aicos.i ammieme P6mv

, But)lphenol Q.... . FwW Cnid Carten tetr.whkir.Je g Rmen teme i ob paw TABLE I Cd'h ud

WNm "*? I*> Ga4 ChlonnaieJ phemds llish fic LI G.3 4 INORGANIC CHLMICAL SERVICE sodium m. dis or. t ieu Cau smriss em.- . ,=w r. sus e -mmer- - C Noniten. ine . . . . Room grne. lA P.w I'" '" Creo%ow Rcem :mp iA ( h = =J I sen n r I)f' of (,om. 2 4 0.chhuvphenoi n v.

'Q' 140 l' amm genip Iwu GawJ qh g n%I Wrue, ns Temp /4ra mi ne s.ihanolim.ee 9u ) > ic W Gasi

~ ~ ~ ---- - --

Eih> l akohol.14"4 Nmm hee tb bad Amnsm.um sNonite War m F wid (nuJ Eth>l nicoNt Wrf kmmwat ib (nul E t hgl O kvede .. twu l'mr Amammami h drowle, 3

eone Nim.m ie.rt Iab Good Eih> kne Si gol - au in :tmi t F .e ki (nul Itrine. utuiased .

Fieu Ga4 F eron lid *aru Noom te'"o Ih dast B one is.unin i Briunme 0.wn.Ji Hot Rmm temp.

heLJ Ga=J l..b Pm' f,'[r '1 1 1.gsnJ km m kirr l af p g G. net.ne Rmm temr 1h tww Calcin.n s hkinde h+== . 7tr4 13 anJ Wil' F.c y Gawl Glycenne Nau temo lab 6.* ml C.nicium t hl.whk. ,ikdi6 n 327 i Fiekt Gutd MJ I W'I N#* ""P 0# O'"

Ces hon Jasilhik jl'.4 feel N 6s .n k nm i sh I'ms 427 l' FwlJ >au A e.m e ne kron u es i ah Pin e Chhwine in H 0 koom Wmp ib Gmd t.anseed imi Room krf. I d' G md Chnun e .nn! pl t'ng m do km mtemp Lab GaQ l ubricating oils . ~ . . . . F ielJ (but

[ ( beommm mullatt. b.mc pu s ad 264 F F seq Ga4 d

,g p l.

) ,i,w .aism. mr, iso r t.is o.i ,,,y ,mt,J, , ., o . . ,

n, J n,hi .,nu a rieu Gad mih,i a. ,i.iw ....kwn. uns,.. u, c.au it)JrogMw e ..eJ F eeW F air Atoniwhkaaccur as.J l oclJ Osm4 .

Il>J.oern perouJe. Nf6 ktwn icmr i uh Fair H,Jm s n re ooJe. +r4 R 00, ,,- p iu Noi use NdPh{hf ' df*'{ gg c,1,,, ,o.s.,ne I mie heu Good A M 11 No i I sh GaQ Nmm wmr stapwunm s hk rde. .W4 1to F F eelJ Gad l'hermd. Inr4 Room hmp I ob ein Nans .J. mne Roorn wmp. L.6 Fan Ph'*"3" " * .*('J ' k bad Neo n edJ. Jiri Phenothunne Fwkl G.ul Fieu Fer pg,,gs,g,,,,,,, F .e y gag PNwrNu w .%.J. come kmm icene i ah Grw4 o. phen,tehenol t wlJ Ga.J PNwpNw.e cI 44h F F ielJ Ganj pron tbenreme F AlJ P eir l'ota%%twni sNomate Fie u Good N %'; g* g ,g gj p P ...um >>r n.<o. m km, ump t# em, s e ,,_',', ,o,_,,,,'d",,Q',U . . . , ,,.a uma Pouumni am ate, we'J km mtent Ia GooJ Near.t .n m1 ReL'* 24 =11' > wiJ f .ut I %Qam.Je 404 > Fie y bmJ %'ce"8 d(d (ht' h*'> l ah l'a 4 NQ wm cetatc. vi'J Raim ttmr. l ..h Ga4 f,', y ,'[,'".I' ,"'

n, l{ Q l ,")

N st.wm carbon..te 2*i kmm kmp 1sb Gad T ncihani*acine kmm terup i ah Gad bd.um a hkw hk. ufd kmm hnt th ( wuJ s ar. = ri w www-am .- r a re- a s -s LV.wm cNewikite tu D 6 w k' Ga4 IAN..

kmJ.um J.Nonde i14 i ewu GmJ GAS SERVICE 50.um 1.pl.osdi;. Dre " ~ ' " " " ' ~ - ~ ~ ' " ' " ' " ~ ' ' " " *

kmmwet 1 ..h t air 1rit en 7, ire kiJa.ni h,J ouJe. 4 % to 7"*i %ial i i eclJ t av 3,n , , pf cy,,,.

NWwm h3J.ouJe. medien ini1 i w Li ) au g ,, g ,, y ,,,y, p,,,, ,,,,

SV un rh.nph.ac. wa J kaen kn e i .ib P6mv

$Q.wm ul .de kman Wnt I ah (k s.J A cin kne - l aclJ (Lad l eekt Paw A mam *u il +isli i bal.om wiltsk. rrm4Nn

$atan , mm.lu n

%.it u. J hk w .J, 34 mi l

,to I i A IJ ( u.nd Amn=sau Ovi h k.im p hompem.h . . . . .

l *G W I ab Iwu ( .a 4 h*"J Gad

.ol Iwu l .ar ( u w ine y n ( ou os hot Iwy (nmiJ tanwg s.J.loto bri kan Wmp i ah b ul U.n. n dvul .. twW (nad k L.liune .ol ?k to *W. Riun A nt l % ki (Lud U ' d" % '"' * ' S' h"C ') 0" . . .04III IWU I Saiwa n .. .J e...e liti i s tJ b.ul ll>Jngastui rn w n% e = 1:0 hi *(si l I wkl I(w"". mJ g g,, , , p ,4); p g,y gag

\- %.n o ..n . .,t kmm h et I sh I +4 %.U n. g.. v.t. 1A kiun h nt. (had

u .a . , lion e sol.l 1%U (nut %icam 24n 1 I Ekl h^d

,, , ,;, . ,, . ., .. .m .u. et- . -i * . 6 0-13 h

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TyrtcAt rcortitt::s m q w.,- ,, B 4 4

'830..616 8001 m .,..a.,,y Coio r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . whi t e . i ra n ,iu c e n i 3

( Serviceable Temperatur Range, degrees F .................'....... -40 to 500 Pc n e t rat io n, u n wo r L e d :.. ..... . . .... . ... . .. . . .. ..... . . ... . ...... . . . . . . .. . 2 2 0 work e d 60 s t ro k e s.. . . . . . .. .. . . .... . .. . . . .. .. . ... . . . . ... . 2 4 0 g

q N . L G . I . Con sis t e n c y . . . . . . . . .. . .. . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . .3. . . . .

Bleed 8. percer.t. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at 3 90 F. . ................................ 0.2 Evaporation'. percent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at 390 F. .................. ....... 1.0 1 8

Wa t e r Wa s hou t . pe rc e n t .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.0 A ppare nt De nsit y, grams pe r c c ... .. . ... ..... ...... .............. ... l .0 Solidification Point. degrees F less than ............................ -45 M e!!ing Point ............. .. . ............................................None '

8 ASTM D 217451

Tested occcad.ng to htlL.l.a66n a4 Amen.imeni 1. .$

'ASThtD1:W4\ .

,i HOW TO USE Bef' ore u;!ng Dow Corning ill compound a thorough cleaning to remove n!1 7 contaminants is recommended. Although Dow Corning 111 compound in very '

tacky, normal grease dispensin; equipment can be used. A Imge pecuure drop does occur. howescr. when pumping this material. Therefore. pumpint over a long distance is not recommended. A gu:de for adopted dispen,'r.; equipment is available upon request.

Dow Corning 111 compound can be dispersed in Lerosene. Stoddarst solvent, benzene, toluene, ethyl ether. petroleum ether or chlorinated solventi. This is an advisable application method if the lubticant must penetrate to gis c adequate J

(

protection. However. Dow Corning 111 compound is insoluble in wMer. rr.eth. ',

anol, ethanol, acetone, glycol, or glycerine.

PACKAGING 2 ounce tubes 10 pound cans ,

6 ounce tubes $0 pour.d rails I

CAUTION Dow Corning 111 compound ma) cause temporaty discomfort if rubbed inte the eye. It is essentially non.iriitating to skin. '

ce. c.~r es e eentms i.eseme+

ei to. cows cnnt es C

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4 APPENDIX H s.

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QUALITY INFORMATION REQUEST / RE E E(QI)

DIVISION CF NUCLEAR ENGINEERING

4 idTERN AL List ONLY I '

( '

rim 5 ACCNO B44 DOCUMENT NUMBE R

'880518 823 1 k,

p, A. Koow r7 i wiot ud C-r_ otn MEB Sou 8t503'2. RO

'"f..L.b6.Lr73 un Bit o c-K PAGE1oF 7O .s TYPE OF DOCUMENT . Daft B REQUEST NEE 0 0 ATE 5/19 / 8 8 N Y IOi l936

___________________________________ ,L ANT A~D UNif I l RELEASE R E F. OIR $QM , OdtT $ ( f 7,,,

REFERENCED DOCUMENTS AVAILABLE IN QNE OF THE RIUS SYSTEMS ATT ACHMENT TO Twis Qim DOCUMENT IDENTIF YING NUMSER DOCUMENT ATT ACHMENT NUMBt m site. ( I

%6 ie 2. 2.

T< t e t e p y h e m Ps-~ i C im >* N 1 4e '7"o % 6J.% **

~ -

SUBJECT "6 G.T A R A'4)t ATs o *J 4 0 C A'r6 0 W SPEC.t Ftc b o% E 5 f o =7 C0 0TA t eJ m G 4 T tsoL.A raceJ V A LV G. S SYSTEMS AFFECTED UNiD l 5 YSTE M ID QU ALITY INFORM ATION Rt OVESTE D I RE LL A$t D

( Pl e a.s c. Se. e ,8cz Q g (Q }, l l 1 l l l 1 l l PREP RED RtyttWED(RELLASESONLil

                    .             No                -

APPROvtD (BR ANCH CMitF/ PROJECT ENGINt t R) k P hl i TVAtot:3 tcN t 444) A. A. 0 %% < ts - U#A.C w etsy.wiciL6 t.o.A.O. c apv.,a gN.J. ' ,w.s U 4wC.W.cc If A, 4 .s bt 56 t.g. f-MM.6GO

                   <t ( Atta;*waint RIUS. 5L 24 C K                                                                                                                 ,

J , p, t, 4 4 g g. , q g ) ( g . v. _ N ___ ___._.___._____ . . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _

8 f 844 '88 0616 8 0 0 M' ~' ~ QIR MEB SGN 86032RO J 5/6/88. This Q!R confirms your conversation with Earl Daugherty on 'l please provide accident BETA radiation specific doses for the System 30 valves in containment and in the Instrument Room. The valves are listed on attachments 1,& 2. Attachment 3 is a telecopy from paul Simmons to Tom Witmer dated 5/18/88 which depicts the installation configuration for each valve in question. This telecopy should provide you with the information you requested to a13ew you to perform the BETA

radiation location specific calculations, Should you require any f urt her t riformat ion, please contact paul Simmons at SGN 3490.

please contact William Estes at SON 5486 for work activity number. (_ ; O J

         *     .e             Gus11fi20 tion Rging Cnalycio R: port AA-18                                                                  ;(

M 4 h u:.b w e d I , B44 '88 0 616 8 0 0" em'"o"S""f ' ( - ss m % cv '* - 0 10 ELL L Valve Deatt - ID NO._ Mark No. Item No. . Erit t Drawlsn No. FCV-30-7 47W915-54 1 E-3344 E-3346 !

                     ,   FCV-30-8  47W915-64                  5          E-3345       E-3348

FCV-30-9 47W915-55 1 E-3344 E-3346 ' FCV-30-10 47W915-65 5 E-3345 E-3348 FCV-30-14 47W915-56 1 E-3344 E-3346

                       , FCV-30-15 47W910-66                  5          E-3345       E-3348 FCV-30-16 47W910-57                  5 E-3344       E-3346 FCV-30-17 47W910-67                  5          E-3345       E-3348 FCV-30-19 47W910-59                  2          E-3344       E-3346 FCV-30-20 47W910-69                  6          E-3345       E-3348 FCV-30-50 47W910-62                  5          E-3345       E-3348                                                    '

FCV-30-51 47W910-52 1 E-3344 E-3346 FCV-30-52 47W910-63 5 E-3345 E-3348 FCV-30-53 47W91!!-53 1 E-3344 E-3346 FCV-30-37 47W91!$-60 3 E-3344 E-3346 FCV-30-40 47W91!$-70 7 E-3345 E-3348 FCV-30-56 47W91!:5-61 5 E-3345 E-3348 FCV-30-57 47W91t5-51 1 E-3344 E-3346 FCV-30-58 47W913-68 6 E-3345 E-3348 FCV-30-59 47W915-58 2 E-3344 E-3346 , O P 6 P t i 1

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Qualificction Aging An31y310 R;ptet AA-18 B44 '88 0616 8 0 0,,dhC h%<M

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TABLE 1 GAMMA MAX. , PETA GAMMA 40 YR NORMAL / TVA Valva Accident Accident INTEG ID # Room Dese MAX. Environment Dose QDiL__ ABNORMAL Drawino 8 U-O 4.7 E8 1.00 L7 1.0 E6 110/120F 10 U-C 47E235-44 R3 ~ 4.7 E8 1.00 E7 1.0 E6 110/120F 47E235-44 R3 56 C-AC1 4.7 E8 1.00 E7 2.0 E7 120/130F 47E235-45 R3 17 C-ACJ 4.7 E8 1.00 E7 2.0 E7 120/130F 52 C-AC3 47E235-45 R3 4.7'E8 1.00 E7 2.0 E7 120/130F 47E235-4b ..'3 15 C-AC4 4.7 E8 1.00 E7 2.0 E7 120/130F 40 C-AC4 47E235-45 R3 4.7 EB 1.00 E7 2.0 E7 120/130F 47E235-45 R3 50 C -AC4 4.7 E8 1.00 E7 2.0 E7 120/130F 47E235-45 R3 3 7 ANN 5.0 E5 1.25 E7 9 1.0 E6 110/120F 47E235-47 R3 ANN 5.0 E5 1.25 E7 1.0 E6 110/120F 14 ANN 47ES35-47 R3 5.0 E5 1.25 E7 1.0 E6 110/120F 47E235-47 R3 16 ANN 5.0 E5 1.25 E7 1.0 E6 110/120F C 19 ANN 47E235-47 R3 5.0 E5 1.25 E7 1.0 Eb 110/120F 47E235-47 R3 37 ANN 5.0 E5 1.25 E7 1.0 E6 110/120F 51 ANN 5. 0 E5 47E235-47 R3 1.25 E7 1.0 E6 110/120F 47E235-47 R3 53 ANN 5.0 E5 1.25 E7 1. 0 E8, 110/120F 57 ANN 47E235-47 R3 5.0 E5 1.25 F7 1. 0 F 4 110/120F 47E235-47 R3 59 ANN 5.0 E5 1.25 E7 1. 0 i6 110/120F 47E235-47 R3 20 C-I!R 4. 7 EB 1.00 E7 3.5 E5 75/120F 58 C-!!R 47E235-48 R3 ; 4.7 EB 1.00 E7 3.5 E5 75/120F 47E235-48 R3 ' (NOTE - The postulated accident temperature profile is included :' in the thermal analysis by use of the Accident Degradation Equivalency calculation) i l

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Qualification Aging Analyata R port AA-18 Bd' '880616 800 ( . %!i

                                                                   ;\nf 1

4 APPENDIX I I i e a L i i 8 4 i

              ,                                 -               -. . .. . t

- -~ ~ 844 '880616 800.[ QUALITY INFORMATION REQUEST / RELEASE (QIR) DIVISION OF NUCLEAR ENGINEERING (INTERNAL USE ONLY) y1 n' QA Record l"'"$ = B45 W 0609 26 2 1 (i l f0 R. E. Daniels, DNE, DSC-H039, Sequoyah lDOCUENT NUPBER l01R NTESQN88193 y4 I lFADM V. A. Blanco, DNE, DSC-A030, Sequoyeh l l l lPAGE I 0F I l l' " ' C"*"' I l( ) REQUEST NEED DATE Il ^ JUN 091988 1 l - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - - - - _ _ - - - I rtAxT AND UN i r I (X) RELEASE REF. CIR EBSON88032RO l Seouoveh units I and 2

    - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _RE_FERE,NCED _DOCU Q TS,_ _ _ _ _ _ _ _ _ _ _ ,, _ _ _ _ _ _ _ _ ,, _ _ _ _

l Avellable in One of the RIMS Systems l Attachrent to This 0IR l l Doewent identifyino Neber l Doewent Attachmet Ntsder ( l i 1 lQlREBSQN88032R0 B44 880518 823 l I I l l l l l l l l __ 1 l l SUBJECT l l Post-LOCA Location Specific Beta Doses for Containment isolatico Valves l l -._ l lSYSTEMSAFFECTED lUNID/SYSTEMID l l Contalewmnt Ventilation Systnm l 50 l lQUALITYINFORMATIONREQUESTED/ RELEASED l l l to provide the postaccident beta doses for those Systern 30 l C' DhE/NTB/APS3 received talves listed in reference 1. Aacalculation request(SQUAPS3-091) (reference 1)is in progress that will calculate these doses. l lPrelimicarydosevalueshavebeencalculatedandareswmarlzedbelow. l l l l The valves are of three different dlareters: 8-inch, 12-inch, and 24-inch. The maxirrun beta dose calculated l l for the 8-inch valves was 7.lE+2 rad, for the 12-inch valves the maxirnun beta dose was 1.6t'+3 rad and for the l l 24-Inch valves the noxirrun beta calculated dose was 6.4E+3 rad. l 1 I lThisQlRwillberevisedwhenSQNAPS3-091isissued. l l l l . I I l l l l l l l l l 1 - 1 I I I I I I I l PREPARED lREVIEhtD(RELEASESONLY) l l

                  -   )"                                         !   .2. I. M                                                           l l APPROVED (BRANCH OllEF/ PROJECT EkGlhEER)                                                                                           l uf            f     1  fW h W. WoYJH1 TVA 10829 (DNE-CA-6-86)        //                                                                                  DhS4-4118Q cc (Attachrents): RlRS, SL 26 C4                J. B. Hosner, DN,E, DSC-E, Sequoyah          D. W. Wilson, WID Cl26 C4 V. C. Dishman, W6 0221 C4                    C. A. Chandley, v7 Cl26 C 4         ,

I

B44 '880616 800 $ Qualification Aging Analycio Rsport AA-10 A 4 c p. ( .k:,) s e APPENDIX J ( e L i i i , s n i'

                                                                                   ~g n

                                                                                           .i B44      '880616 800 THE PURPOSE OF TH8S CALCULAT!3N IS TO DETERMINE THE RADIATION LIFE         c.

OF MATERIALS BASED ON RADIATI]N LIFE METHODOLOGY PER NEP 5.10 j

     -----------------------------. ----..------------------------------   . of)

C MATERIAL RADIATION DAMAGE LEVEL (RFDS)......: MATERIAL DAMAGE LEVEL (%).....: 49 5e7 O

                                                                                    !h MATERIAL PROPERTY............. ELONGATION ACCIDENT RADIATION DOSE (GAMMA).............:          3.8e7                    .ef ACCIDENT RADIATION DOSE (BETA)..............:          6.4e3                          3 40 YEAR TID (RADS)..........................:       ,

1e6 7,f{;!f RADIATION LIFE (YEARS)......................: 479.74 ENVIRONMENTAL DRAWING NO.: 47F235-44, R3 . . BETA DOSE PER GIRNTBSON88193 (B45 88 0609262) FOR 24 INCH VALVE AA-18 13 , ( 1 l 1 , 1 A l $ L .l

B44 '88 0616 8 0 0 THE PURPOSE OF THIS CALCULATION IS TO DETERMINE THE RADIATION LIFE 'N OF MATERIALS BASED ON RADIATION LIFE METHODOLOGY PER NEP 5.10 p C MATERIAL RADIATION DAMAGE LEVEL (RADS)......: 5e7 y) MATERIAL DAMADE LEVEL (%).....: 49 MATERIAL PROPERTY............. ELONGATION d) ACCIDENT RADIATION DOSE (GAMMA).............: le7 ACCIDENT RADIATION DOSE (BETA)..............: 6.4e3 40 YEAR TID (RADS)..........................: 2e7 RADIATION LIFE (YEARS)......................: 79.99 ENVIRONMENTAL DRAWING NO.: 47E235-45, R3 . , BETA DOSE PER GIRNTBSON88193 (B45 80 0609262) FOR 24 INCH VALVE ' AA-18 14 o ( O I I l (. ~i l A k s k (

                                                                 ,    ,             - :   .J

      . - - - --                                                  .                 :s B44   '880616 800' THE PURPO2E OF THIS CALCULATION IS TO DETERMINE THE RADIATION LIFE       .

OF MATERIALS BASED ON RADIATION LIFE METHODOLOGY PER NEP 5.10 MATERIAL RADIATION DAMAGE LEVEL (RADS)......: C MATERIAL DAMAGE LEVEL (%).....: Se7 }7 49 MATERIAL PROPF.RTY............. ELONGATION 'Nr.:i. ACCIDENT RADIATION DOSE (GAMMA).............: 1e7 ACCIDENT RADIATION DOSE (BETA)..............: 7.1e2 40 YEAR TID (RADS)..........................: 2e7 RADIATION LIFE (YEARS)......................: 80.00 ENVIRONMENTAL DRAWING NO. 47E235-45, R3 - BETA AA-18 DOSE PER GIRNTBSON88193 (B45 88 0609262) FOR 8 INCH VALVE 15 Mr i

B44 '88 0616 8 0 0 4.... THE PURPOSE OF THIS CALCULATION IS TO DETERMINE THE RADIATION LIFE IQ > OF MATERIALS BASED ON RADIATION LIFE METHODOLOGY PER NEP 5.10 '3 ( ------------------------------------------------------------------==. MATERIAL RADIATION DAMAGE LEVEL (RADS)...... MATERIAL DAMAGE LEVEL (%).....: MATERIAL PROPERTY.............: ELONGATION 49 5e7 '!.kN Y'f r ACC7 DENT RADIATION DOSE (GAMMA).............: 1e7 ACCIDENT RADIATION DOSE (BETA)..............: 1.6e3 40 YEAR TID (RADS)..........................: 3.5e5 RADIATION LIFE (YEARS)......................: 4571.25 ENVIRONMENTAL DRAWING NO.: 47E235-48, R3 _ BETA DOSE PER GIRNTBSGN88193 (B45 80 0609262) FOR 12 INCH VALVE AA-18 16 4 4 P l l

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ML20196D162

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