ML20045E153

LER 93-017-00:on 930527,determined That Feedwater Isolation Bypass Valves Inoperable Since Pneumatic Positioner & Solenoids Remained in Svc Beyond Qualified Life.Caused by Incorrect Interpretation of Design Documents.Mod Underway

ML20045E153
Person / Time
Site:	South Texas
Issue date:	06/25/1993
From:	Pinzon J HOUSTON LIGHTING & POWER CO.
To:
Shared Package
ML20045E151	List: ML20045E150 ML20045E153
References
LER-93-017, LER-93-17, NUDOCS 9307010186
Download: ML20045E153 (8)

Text

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NRC EORM 366 U.S. C%CLELD REGULATC3Y COMMISSION APPROVED BY OMB C~3. 3150-0S04 (5-92) EXPIRES 5/31/95 ESilMATED BURDEN PER RESPONSE 10 COMPLY WITH

" " ^ ' LL LICENSEE EVENT REPORT (LER) $AR0 CMMEN 5 R GARDING BURDEN ST MATE b THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLE AR REGUL ATORY COMMISSION, (See reverse for required ruber of digits / characters for each block) WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WAGHINGTON, DC 20503.

FACILITY WAME (1) DOCKET NUMBER (2) PAGE (3)

South Texas, Unit 1 05000 498 1 OF 8 TITLE (4) Extencion of FWIBV Positioner and Solenoid Equipment Beyond Qualif_ication Lifn OTHER FACIL ITIES INVOLVED (8)

EVENT DATE (5) IER NLMBER (6) REPORT DATE (7)

FACIL11Y NAME DOCKEi NUMBER SE N S MONTH DAY YEAR YEAR B

HONTH DAY YEAR South Texas, Unit 2 05000499 FA I M CKET " "

05 27 93 93 -- 017 --

00 06 25 93 0r0O OPERATING THlS um IS SusMimD mm TO THE HMMEUS OF 10 CW D (Check one or w e) W) 5 20.402(b) 20.405(c) 50. 73(a)(2)( i v) 73.71(b)

Ma)E (9)

PourR 2WaWO McW MaMW M e) g LEVEL (10) 20.405(a)(1)(ii) 50.36(c)(2) 50.73(a)(2)(vii) OTHER x 50.73(a)(2)(i) 50. 73(a)(2)( vi t i )( A) (specify in 20.405(a)(1)(iii)

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20.405(a)(1)(iv) 50.73(a)(2)(ii) 50.73(a)(2)(viii)(B) 20.405(a)(1)(v) 50.73(a)(2)(iii) 50. 73(a)(2)( x) NRC Form 366A)

LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER (include Area Code)

Jairo Pinzon - Senior Engineer (512) 972 - 8027 COMPL E TE ONE LINE FOR EACH COMPONENT F AltVRE DESCRIRED IN THIS REPORT (13)

CAUSE SYSTEM COMPONENT MANUFACTURER CAUSE SYSTEM COMPONENT MANUFACTURER g

X SJ CPOS V037 YES X SJ SOL A609 YES SLFPL E ME NT AL REPORT EXPECTfD (14) MONTH DAY VEAR EXPE CT ED YES(if yes, conplete EXPECTED SUBMISSION DATE). X NO ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)

On May 27, 1993, Unit 1 was in Mode 5 at 0% power and Unit 2 was defueled in a refueling outage. Plant personnel determined that the Feedwater Isolation Bypass Valves (FWIBVs) in both units were inoperable since the pneumatic positioner and solenoids had remained in service beyond their qualified life, which had been inappropriately extended to 40 years. STP personnel determined that Technical Specification 3.7.1.7 had been violated during various modes of operation since 1992. The inappropriate extension of the equipment qualification life of the FWIBV positioner and solenoids was the result of incorrectly interpreting design documents. Corrective actions include reviewing other safety-related valves with positioners for similar problems, reviewing other solenoid valves qualified for service in a harsh environment to determine if similar problems could exist, reviewing safety-related components which have been classified as non-safety to determine generic implications, and modifying the pneumatic control scheme of the FWIBV to ensure closure of the valves independent of the positioner upon de-energization of the safety-related solenoid valves.

l NRC m M 366 (5-92) 9307010186 930625 PDR ADOCK 05000499 S PDR ,

LER-93\L93017RD.U2

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NRC FORM 366A U.S. NUCLEAR REQJLAT O Y COMMISSI C APPROVED B7 OMB CD. 3150-0104 -

($.92} EXPIRE 5 $/31/95 ESilMATED BURDEN PER RESPONSE TO COMPLY WITH THl$

INFORMATION COLLECTION REQUEST: 50.0 HRS.

FORWARD COMMENTS RFCARDING BURDEN ESTIMATE TO THE L'ICENSEE EVENT REPORT (LER) INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB TEXT CONTINUATION 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001 AND TO THE PAPERWORK REDUCTION PROJECT (31$0-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAMF (1) DOCKET NUMBER (2) LER NUMBER (61 PACE (3)

YEAR SEQUENTIAL REVISION 2 OF 8 South Texas, Unit 1 05000 498 93 -

017 -- 00 TEXT (if more space is reouired, use additional copies of NRC Form 366Al (17)

DESCRIPTION OF EVENT:

On May 27, 1993, Unit 1 was in B de 5 at 0% power and Unit 2 was defueled in a refueling outage. Plant pe. .el determined that the FWIBVs in both units were inoperable since the pm .ic positioner and solenoids had remained in service beyond their qual! A tife, which had been inappropriately extended to 40 years. STP personne2 determined that Technical Specification 3.7.1.7 had been violated during various modes of operation since 1992. Technical Specification 3.7.1.7 requires that each FWIBV be operable while in Modes 1, 2 and 3.

POSITIONERS The operability determination for the positioner event was based on a review of Seismic / Environment Qualification (EQ) reports and the valve manuf acturer's (Valtek) report. These qualification reports show that the positioner was qualified with the valve and meets the requirements for use as safety-related equipment. These reports also required the positioner sof t parts be replaced periodically. The current EQ document for the positioner indicates that it is a non-safety-related device and, therefore, requires no qualification.

This derating was based on a September 1986, evaluation concluding that the pneumatic positioners were non-safety. Further review of the plant design documents showed that the positioners do perform a safety function and, therefore, the EQ requirements should be maintained.

The EQ report for the positioners requires a replacement of sensitive parts every four years (based on maximum 6 continuous operating temperature of 300 F and a threshold radiation of 10 rads) to achieve a forty-year life. The replacement interval is based on the worst case effect (maximum operating temperature or radiation exposure) to the age-sensitive materials in - the positioner. Design criteria radiation dose for a postulated accident.in the Isolation Valve Cubicle is Jess than the threshold limit, which makes operating temperature the determining factor. Based on an assumed maximum continuous operating temperature less than 212 F, the EQ life for the FWIBV positioners is five years. Based on these design assumptions, the EQ replacement period for Unit 1 expired in 1992, thus making the FWIBV technically inoperable for the last 1.5 years. The Unit 2 FWIBV positioners i have not yet exceeded their five-year life. HL&P is continuing to evalunte the operability of the positioners based on exposure to actual operating conditions.

On May 21, 1993, service requests were written to install a modification to the pneumatic control scheme of the FWIBV to ensure closure of the valves, independent of the positioner, upon de-energization of the safety-related solenoid valves. Since the positioner would no longer be required to perform a safety function, the requirement for EQ was deleted.

l LER 93\L93017RO.U2  !

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NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSICO APPROVED BY OMB NO. 3150-0104 (5-92) EI:PIRES 5/31/95 ESilMATED BURDEN PEC CESPONSE TO COMPLV WITH THl$

INFORMATION COLLECTION REQUEST: 50.0 HRS.

FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE LICENSEE EVENT REPORT (LER) INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB TEXT CONTINUATION 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHlhCTON, DC 20555-0001 AND TO THE PAPERWORK REDUCTION PROJECT (31300104), OFFICE OF MANAGEMENT AND BUDGET, WASHINCTON, DC 20503.

FACILITY NAME (1) DOCKET N1MBER (2) l tFR NUMBER (67 PAGE (3)

TEAR SEQUENTIAL REVISION South Texas, Unit 1 05000 498 3 OF 8 93 -- 017 -- 00 YXT (if more space is reouired, use additional copies of NRC form 366A) (17)

DESCRIPTION OF EVENT: (Cont'd)

A five step plan was established to address the generic implication of the declassification of the FWIBVs safety-related positioners and solenoids. This plan included reviews of the classification of all Valtek valve positioners, positioners provided for safety-related valves, and Heating, Ventilating and Air Conditioning system dampers. This plan also provided for reviews of active parts declassified by use of Technical Evaluations for proper consideration of system / component operation, a sample of EQ packages . for proper consideration of system / component operation, and reviews of EQ for solenoids which are normally energized but were evaluated as normally de-energized.

This review resulted in identifying twelve additional valves per unit where failure of the positioner could cause the valve to be mispositioned. Further review revealed the valves are located in a mild environment and do not require periodic replacement. Other solenoids of the same model used with the FWIBVs were identified and action taken to replace them. These. additional solenoids were on the non-safety related Steam Generator Augmented Blowdown valves, which do not have a qualified life limitation.

SOLENOIDS The Equipment Qualification Calculation Package (EQCP) for the FWIBV safety-related solenoids was reviewed and questions were raised relating to the assumptions used in determining the qualified life. Calculation E ASCO, Revision 3, indicated that the safety-related solenoids are only .

energized during feedwater heatup and not during normal operation.resulting in a qualified life of forty years. Review of the logic drawings for.these s valves showed that there are three solenoids for each FWIBV. One non-Class r 1E solenoid is controlled by a main control board handswitch and is normally de-energized. The other two Class 1E solenoids are each controlled by a combination of the Main Feedwater Isolation Valve control switches and a Feedwater Isolation signal from the Solid State Protection System. A review of the design documents showed that these two safety-related solenoids are energized during normal plant operations. As a result of recalculating the life of the solenoids in the normally energized state using worst case design basis assumptions, the qualified life is 2.5 years. As a result, _the qualified life expired in December 1989 for Unit 1 and June 1991 for Unit 2.

HL&P is continuing to evaluate the operability of the FWIBV solenoids based a on exposure to actual operating conditions.

I LER-93\L93017RO.U1 l t

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NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSI(D APPROVED BY OMB No. 3150-0104 (5-92) .

EXPIDES 5/31/95 ESTIMATED BURDEN PER DESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUES1: 50.0 HRS.

"8" """ EN "'"

L'ICENBEE EVENT REPORT (LER) E $1[o$ ".'A23 Rr'c$ p N AuActM BRANCH MN NT TEXT CONTINUATION 7714), u.S NuttEAR REculAToRY c0MMISsl0N, WASHINGTON, DC 20555-0001 AND TO THE PAPERWORK 8 0104),

(31.,0 OFFICE OF REDUCTION PROJECT MANAGEMENT AND DUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NtMBER (?) LFR NUMBER (61 PAGE (3)

SEQUENTIAL REVISION YEAR

"#8" "#8" South Texas, Unit 1 05000 498 4 OF 8 93 -- 017 -- 00 TEXY (If more space is reovired, use additional copies of NRC Form 366Al (17)

DESCRIPTION OF EVENT: (Cont'd)

Other solenoid valves qualified for service in a harsh environment were reviewed to determine if the same condition could 9xist. The review' included ASCO, Valcor and Target Rock solenoids. This review resulted in a reduction in qualified life for the Preheater Bypass valves and the Main Steam Bypass valves ASCO solenoids. Although their qualified lives have been reduced, they have not yet expired as had the FWIBV solenoids and positioners.

CAUSE OF EVENT:

In both cases, the qualified life extensions of the FWIBV positioners and solenoid valves were the result of incorrectly interpreting design documents.

Plant Operations procedures and discussions between Instrumentation & Control personnel and Design Engineering EQ personnel were used to determine the function of the positioner and solenoids.

ANALYSIS OF EVENT:

The events pertaining to the extension of the FWIBVs positioner and solenoid valve EQ lives are reportable pursuant to 10CFR50.73 (a) (2) (i) (B) . This event represents an operation or condition prohibited by Technical Specification 3.7.1.7. Although the Units were not operated with FWIBVs open, failure of the FWIBV positioners or solenoid could result in t's FWIBVs remaining open.

The following accidents are impacted:

Steam Generator Tube Rupture (SGTR)

The prevalent concern with a SGTR (UFSAR section 15.6.3) is a release of radiation. There is no single failure coincident with a SGTR which involves a dose release path through the FWIBVs. By assuming a single failure of the check valve between a stuck open FWIBV and the Steam Generator, a dose release path could only be provided by a coincident feedwater line break. _However, considering a SGTR coincident with a loss of feedwater piping is beyond the design basis of the plant.

Also of concern for a SGTR is the Steam Generator overfill analysis.-Open FWIBVs may provide additional flow to the Steam Generators in addition to Auxiliary Feedwater (AFW) flow. Failure of the FWIBVs to close has no effect on feedwater system safety function as discussed in the Main Feedwater Failure '

Modes & Effects Analysis (FMEA) (UFSAR Table 10.4-8). The FMEA states that  !

failure of the FWIBVs and Feedwater Isolation Valves to close is backed up by l closure of the feedwater flow control valves. Also, both the Steam Generator Feedwater Pumps (SGFPs) and the Start-up SGFPs trip, ef fectively reducing j feedwater flow to the affected Steam Generator.  ;

l LER-93\L93017RO.u2

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSICD APPROVED BY OMB 110. 3150-0104 (5 92) .

EXPICES 5/31/95

, ESTIMATED BURDEN PER RESPONSE TO COMPLV WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.

FORWARD COMMENTS REGARDING BURDEN ESilMATE TO THE LICENBEE EVENT REPORT (LER) INFORMATION AND REr0RDS MANAGEMENT BRANCH (MNBB TEXT CONTINUATION 7714), U.S. . NUCLEAR REGULATORY COMMISSION, WASHlWGTON, DC 20555-0001 AND TO THE PAPERWORK REDUCTION PROJECT (31$00104), OFFICE OF MANAGEMENT AND BLOCET, WASHINGTON, DC 20503.

FACit1TY NAME (1) DOCKET NLMBER (2) LER NLMBER (61 PAGF (3)

YEAR SEQUENTIAL REVISION South Texas, Unit 1 05000 498 5 OF 8 93 -- 017 -- 00 TEXT (If more space is required. use additional copies of NRC Form 3f4A) (17)

ANALYSIS OF EVENT: (Cont'd) feedwater System Pipe Break A Feedwater System Pipe break (UFSAR section 15.2.8) is defined as a break in a feedwater line large enough to prevent the addition of sufficient feedwater to the Steam Generators to maintain shell-side fluid inventory in the Steam Generators. A break upstream of the feedwater line check valve is bounded by a Loss of Normal Feedwater Flow (UFSAR section 15.2.7). If the break is postulated between the check valve and the Steam Generator, Steam Generator fluid as well as feedwater flow may be discharged through the break. The limiting single failure for the current analysis is loss of safety Train A resulting in a loss of AFW to Steam Generators A'and D. With the failed feedwater line in either Steam Generator B or C, one Steam Generator would be available for Reactor Coolant System cooling.

If the break occurs concurrent with a single failure of the check valve _in an unaffected loop, some AFW flow may be lost to two Steam Generators: (i) the Steam Generator in the loop with the failed check valve (through the check valve and the FWIBV), and (ii) the Steam Generator in the loop in which the break occurs. The flow lost through the FWIBV would be limited by closure of the Feedwater flow control valves as discussed in the Main Feedwater FMEA.

However, there would still be AFW flow to two intact Steam Generators, which is bounded by the current analysis.

Loss of Norma _1 Feedwater Flow For a Loss of Normal Feedwater flow accident (UFSAR section 15.2.7), the limiting case is a single failure loss of Safety Train A resulting in a loss of AFW flow to Steam Generators A and D. In this instance, the check valves between the Steam Generators and the FWIBVs would prevent loss of Steam Generator inventory through the misaligned FWIBVs. As discussed in the SGTR analysis, there is no credible f ailure coincident with the single failure ol' one of the check valves which could result in loss of AFW inventory back through the FWIBVs.

Small BJ_eak Loss of Coolinct Accident The limiting single failure for a Small Break Loss of Cooling Accident (LOCA)

(UFSAR section 15. 6. 5) is a loss of safety Train A resulting in a loss of AFW flow to Steam Generators A and D. The same discussion for the Loss of Normal Feedwater Flow accident above is applicable here; taking the check valve failing as the single f ailure is bounded by the loss of AFW to Steam Generator 1 A and D. l LER 93\L93017RO.U2

NRC FORM 366A U.S. NUCLEAR REGULATOY COMMISSIC APPROVED BY OMB NO. 3150-0104 (5-92) .

ERP!3ES 5/31/95 ESitMATED BURDEN PEQ SESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS, FORWARD COMMENTS REGARDING BURDEN ESilMATE TO THE LICENSEE EVENT REPORT (LER) INFORMATION AND RECORDS MANAGEMENI BRANCH (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, TEXT CONTINUATION WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150 0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503, FACillTY NAME (1) DOCKET NtNBER (2) 1ER WlMBER (61 PAGE (3)

SEQUENTIAL REVISION YEAR

"#"" "#8" South Texas, Unit 1 05000 498 6 OF 8 93 -- 017 -- 00 TEXT (if more space is reauired use additional copies of Nec Form 366A) (17)

ANALYSIS OF EVENT: (Cont'd) liain Steam Line Brea}: (MSLB)

In the event of a MSLB accident, significant feedwater flow through the open FWIBVs would not occur because of feedwater flow control valve closure and tripping of both the SGFPs and the Start-up SGFPs as discussed above.

Based on the above considerations, safety significance from postulated misalignment of FWIBVs, is very low.

CORRECTIVE ACTIONS:

1. The Preventive Maintenance (PM) activities for EQ replacement for the FWIBV solenoids have been reactivated.

2. A review of other safety-related valves with positioners was performed to identify similar concerns. A group of twelve Chilled Water valves was identified with a similar configuration; however, these valves are not located in a harsh environment and no periodic parts replacement is required to maintain the qualification. Further investigation revealed that the EQ of the Chilled Water valve positioners have not been altered.

3. A modification to revise the pneumatic control scheme'of the FWIBVs to ensure closure of the valves independent of the positioner upon de-onergization of the safety-related solenoid valves, has been developed.

This modification has been implemented in Unit 1 and will be implemented by the end of the current refueling outage in Unit 2.

4. A review of other solenoid valves qualified for service in a harsh environment was performed to determine if a similar problem existed. This review included ASCO, Valcor and Target Rock solenoid valves. This review resulted in a reduction in qualified life for the Preheater Bypass valve and Main Steam Bypass valve ASCO solenoids. Although reduced, their qualified lives reduced, have not yet expired as had the FWIBV solenoids and positioners. EQ PM activities will be developed to ensure replacement prior to expiration. These PM activities will be developed by August 18, 1993.

LER-93\L93017RD.U2

NRC FC:n 366A U.S. NUCLEAR REGULA1C2Y COMMISSICC APPROVED BT CMR No. 3150-0104 (5-92) .

EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.

FORWARD COMMENTS REGARDING BURDEN ESilMATE TO THE LICENSEE EVENT REPORT (LER) INFORMAfion Ano REcokDS MANAGEMENT 6 RANCH (MNBB TEXT CONTINUATION 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555 0001 AND TO THE PAPERWORK REDUCYlON PROJECT (31$0-0104), OFFICE OF MANAGEMENT AND BUDGET. WASHINGTON, DC 20503.

FACit!TY NAME (1) DOCKFT NtmarR (?) trR NUMBER (61 PAGE (3)

SEQUENTIAL REVISION YEAR South Texas, Unit 1 05000 498 7 OF 8 93 -- 017 -- 00 TEXT (if more space is rew ired, use additional copies of NRC Form 366A) (17)

CORRECTIVE ACTIONS: (Cont'd)

5. Information regarding the positioner and solenoid events will be provided to Engineering personnel as lessons learned. In addition, Engineering personnel will be sensitized on the need for attention to detail with regard to EQ and verifying assumptions. This action will be completed by July 9, 1993.

6. To address generic implications, a five-step review process was developed to evaluate the part reclassifications and EQ issues indicated as a result of this LER. This review has not resulted in any additional plant safety issues. As a result of this review, the sample scope has been expanded.

This additional review will be completed by July 21, 1993.- Based on this additional review, corrective actions and recurrence control measures will be developed as necessary, bDDITIONAL INFORMATION:

The FWIBV positioners are manufactured by Valtek and are model number 80R.

The FWIBV solenoids are manufactured by ASCO and are model numbers 206-832-4VF, NP831655E, and NP8321A2E/A6E.

During the past two years, two LERs have been submitted to the NRC which were related to Equipment / Environmental Qualification problems. These LERs are as follows:

o Unit 2 LER 93-008 regarding a Technical Specification violation due to a failure to maintain Environmental Qualification of a Residual Heat Removal Motor Operated Valve.

o Unit 1 LER 93-017 regarding a failure of an Essential Cooling Water traveling screen drive coupling.

The following information is with regard to an event that was discovered on  ;

April 21, 1993, in which the FWIBV was thought to be open when a Maintenance ,

technician observed pointer movement on the stem clamp during maintenance work l on the FWIBV. The technician believed that since the pointer movement on the l stem clamp was sudden, the valve stem had moved and the valve had been partially open. This was thought to be reportable because the plant had operated in various modes contrary to Technical Specifications. The valve was

. thought to be open since April 25, 1992. Further investigation determined that the valve, in fact, was closed as required during this time period.

LER-93\L93017RO.U2

o ECFORM366A U.S. NUCLEAR REGJLATORY COMMISSICC APPROVED BY CMB NO. 3150-0104 (5-92) .

EXPIRES 5/31/95 EST! MATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.

FORWARD COMMENTS REGARDING BURDEN ESilMATE TO THE

. $ICENBEE EVENT REPORT (LER) INroRMAT!ow AND RECORDS MANAGEMENT BRANCH (MNBB TEXT CONTINUATION 7714), U.S. NUCLEAR REGULATORY COMMIS$10N, WASHINGTON, DC 20555 0001, AND TO THE PAPERWORK REDUCT10W PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON. DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMR[R (65- PAGE (3)

SEQUENTIAL REVISION YEAR South Texas, Unit 1 05000 498 8 OF 8 93 -- 017 -- 00 TEXT (if more srmee is reo; ired. use additional cordes of NRC Form 366A) (17)

ADDITIONAL INFORMATION: (Cont'd)

.On May 22, 1993, Fisher Service Company performed a series of diagnostic tests to determine the overall operating condition of the valve. The tests verified that the valve stroke, packing friction, and seat load were within specifications, but did indicate minor resistance in the upper portion of the stroke. This resistance did not prevent or hinder the valve from stroking.

Fisher felt that the resistance could be from galling in the stem guide area or packing material buildup on the stem of the valve. The actuator spring rate, total travel, and bench settings were within specifications. On May 25, 1993, the FWIBV was disassembled and an inspection performed by Engineering and Maintenance personnel revealed signs of normal wear except for thread damage on the upper 5/8 in. of the actuator stem and valve plug stem. There were no indications of any sticking or binding in the valve actuator or body.

On May 26, 1993 the Valtek vendor was brought in to inspect the valve.

Valtek's inspection concurred with HL&P's finding.

HL&P has analyzed Fisher's diagnostic test and Valtek's report. The review did not lead to any conclusive evidence that the valve was opened during nornal operations or that binding occurred which prevented closure.

Therefore, HL&P has determined that the valve was not open and this event is not reportable. The basis for the conclusion are as follows:

o The I&C technician could not be certain that he witnessed valve stem movement.

o Valve travel is limited by plug motion between seat and backseat which was measured to be 1~ . 67 2 inches. The limit switch settings were found approximately 1.5 inches apart, indicating that the valve had been stroking fully.

o No physical evidence was found upon inspection of the valve components which could have caused or indicated mechanical binding of the valve, o Fisher's diagnostic testing revealed normal valve _ operation with relatively little packing drag, o The discovered thread damage on the actuator stem and valve plug stem at the point of overlap suggests that when the valve stem clamp was loosened 1 I

for maintenance, the valve actuator stem moved downward while the valve plug stem was stationary due to the valve plug being seated. This also l accounts for the observed pointer movement. I lek-93\L93017RO.U2 l