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{{#Wiki_filter:PS~G Public Service Electric and Gas Comj'.?any P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nuclear Business Unit AUG 2 21996 LR-N96253 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:
* Public Service Electric and Gas Comj'.?any P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nuclear Business Unit AUG 2 21996 LR-N96253 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:
LER 272/96-012-00 SALEM GENERATING STATION - UNIT 1 FACILITY OPERATING LICENSE NO. DPR-70 DOCKET NO. 50-272 This Licensee Event Report entitled "Potential Loss of Residual Heat Removal Capability Due to Inadequate Valve Design" is being submitted pursuant to the requirements of the Code of Federal Regulations 10CFR50.73(a) (2) (ii) (B).
LER 272/96-012-00 SALEM GENERATING STATION -UNIT 1 FACILITY OPERATING LICENSE NO. DPR-70 DOCKET NO. 50-272 This Licensee Event Report entitled "Potential Loss of Residual Heat Removal Capability Due to Inadequate Valve Design" is being submitted pursuant to the requirements of the Code of Federal Regulations 10CFR50.73(a)
rr;e;~
(2) (ii) (B). Attachment DVH/tcp . C Distribution LER File 3.7 ..,...., ' 1 n'-' ,.l t; :-:;. :r.L'..'. .
David F. Garchow General Manager -
David F. Garchow General Manager -Salem Operations 95-2168 REV. 6/94
Salem Operations Attachment DVH/tcp
* Document Control Desk LR-N96253 Attachment A The following represents the commitments that Public Service Electric & Gas (PSE&G) made to the Nuclear Regulatory Commission (NRC) relative to this LER (272/96-012-00).
  .C             Distribution LER File 3.7 1   n'-' ~Y*m?r i~ ~n ,.l t; :- :;. :r.L'..'. .
The commitments are as follows: 1. The RHR flow control valves will be replaced prior to Mode 6 in Salem Units 1 and 2. 2. A review is in progress to identify other Fisher Model 7600 valves that may be subject to the same conditions of design inadequacies and/or increased torque demands resulting from packing changes. The review will be completed prior to Mode 6. Corrective actions, if any, will be completed to support system operability in accordance with Technical Specifications requirements.
95-2168 REV. 6/94
*
 
* NRCFORM388 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 (4-95) EXPIRES CM/30198 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS MANDATORY INFORMATION COLLECTION REQUEST: 50.0 HRS. LICENSEE EVENT REPORT (LER) REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE LICENSING PROCESS AND FED BACK TO INDUSTRY.
Document Control Desk LR-N96253 Attachment A The following represents the commitments that Public Service Electric
FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE.INFORMATION (See reverse for required number of AND RECORDS MANAGEMENT BRANCH (T.e REGULATORY COMMISSION, WASHINGTON, DC , AND TO digits/characters for each block) THE PAPERWORK REDUCTION PROJECT OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, 20l503.
& Gas (PSE&G) made to the Nuclear Regulatory Commission (NRC) relative to this LER (272/96-012-00). The commitments are as follows:
* FACILITY NAME (1) DOCKET NUMBER (2) PAGE(3) SALEM GENERATING STATION, UNIT 1 05000272 1 OF4 TITLE (4) Potential Loss of Residual Heat Removal Capability Due To Inadequate Valve Design EVENT DATE (5) LER NUMBER (8) REPORT DATE , 1) OTHER FACILITIES INVOLVED (8) YEAR I I FACILITY NAME DOCKET NUMBER llONTH DAY YEAR SEQUENTIAL  
: 1. The RHR flow control valves will be replaced prior to Mode 6 in Salem Units 1 and 2.
'REVISION MONTH DAY YEAR NUllBER NUMBER SALEM UNIT2 05000311 07 23 96 96 012 00 08 22 96 FACILITY NAME DOCKET NUMBER --05000 OPERATING N THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check one or more) (11) MODE(9) 20.2201(b) 20.2203(a)(2)(v) so. 73(*)(2)(i)
: 2. A review is in progress to identify other Fisher Model 7600 valves that may be subject to the same conditions of design inadequacies and/or increased torque demands resulting from packing changes.
SO. 73(a)(2)(viii)
The review will be completed prior to Mode 6. Corrective actions, if any, will be completed to support system operability in accordance with Technical Specifications requirements.
POWER 000 20.2203(a)(1) 20.2203(a)(3)(i) x SO. 73(a)(2)(ii)
 
SO. 73(a)(2)(x)
NRCFORM388 (4-95)
LEVEL (10) 20.2203(a)(2)(i) 20.2203(a)(3)(ii)
U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)
SO. 73(a)(2)(iii) 73.71 -20.2203(a)(2)(ii) 20.2203(*)(4)
* APPROVED BY OMB NO. 3150-0104 EXPIRES CM/30198 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS MANDATORY INFORMATION COLLECTION REQUEST: 50.0 HRS.
SO. 73(a)(2)(iv) x OTHER 20.2203(a)(2)(iii)
REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE LICENSING PROCESS AND FED BACK TO INDUSTRY.               FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE.INFORMATION AND RECORDS MANAGEMENT BRANCH (T.e ~NUCLEAR (See reverse for required number of                                  REGULATORY COMMISSION, WASHINGTON, DC                       , AND TO THE PAPERWORK REDUCTION PROJECT ~160-0104), OFFICE OF digits/characters for each block)                                   MANAGEMENT AND BUDGET, WASHINGTON,             20l503.
S0.38(c)(1)
* FACILITY NAME (1)                                                                             DOCKET NUMBER (2)                                   PAGE(3)
SO. 73(a)(2)(v)
SALEM GENERATING STATION, UNIT 1                                                             05000272                             1 OF4 TITLE (4)
Abetract below or In C Form 366A 20.2203(a)(2)(iv)
Potential Loss of Residual Heat Removal Capability Due To Inadequate Valve Design EVENT DATE (5)                 LER NUMBER (8)               REPORT DATE , 1)                         OTHER FACILITIES INVOLVED (8)
S0.38(c)(2)
IFACILITY NAME                             DOCKET NUMBER llONTH       DAY     YEAR             SEQUENTIAL   'REVISION   MONTH       DAY     YEAR YEAR I    NUllBER        NUMBER                                         SALEM UNIT2                           05000311 07         23       96     96     -  012       -    00       08         22       96 FACILITY NAME                             DOCKET NUMBER 05000 OPERATING           N     THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check one or more) (11)
: 50. 73(a)(2)(vii)
MODE(9)                     20.2201(b)                     20.2203(a)(2)(v)                       so. 73(*)(2)(i)                       SO. 73(a)(2)(viii)
Part 21 LICENSEE CONTACT FOR THIS LER (12) NAME TELEPHONE NUMBER (Include Are* Code) Dennis v. Hassler, LER Coordinator 609-339-1989 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13) CAUSE SYSTEM """""""' ................  
POWER           000       20.2203(a)(1)                   20.2203(a)(3)(i)                   x   SO. 73(a)(2)(ii)                     SO. 73(a)(2)(x)
...., .... , ...... ...... __ ., ...... ...u... REPORTABLE TONPRDS TONPRDS ilti11111111:
LEVEL (10)                   20.2203(a)(2)(i)               20.2203(a)(3)(ii)                     SO. 73(a)(2)(iii)                     73.71 20.2203(a)(2)(ii)               20.2203(*)(4)                         SO. 73(a)(2)(iv)               x   OTHER 20.2203(a)(2)(iii)             S0.38(c)(1)                           SO. 73(a)(2)(v)                 Spec~ln    Abetract below or In     C Form 366A 20.2203(a)(2)(iv)               S0.38(c)(2)                           50. 73(a)(2)(vii)               Part 21 LICENSEE CONTACT FOR THIS LER (12)
SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED MONTH DAY YEAR IYES . XINO SUBMISSION (If yea, complet111 EXPECTED SUBMISSION DATE). DATE(15) ABSTRACT (Limit to 1400 apace*, i.e ** approximately 1S aingl-paced typewritten lines) (16) On July 23, 1996 a review determined that the keys/keyways on the actuators f.or the Residual Heat Removal (RHR) flow control valves are subject to failure. The valves were made by Fisher Controls International, Model Type 656/7600.
NAME                                                                                               TELEPHONE NUMBER (Include Are* Code)
When using the sirrplified Fisher catalog 14 methodology, the calculated maximum stem torque exceeds the vendor specified allowable torque. Preliminary calculations of the shaft torque and resulting average shear stresses in the valve stem key and keyway were also performed.
Dennis         v. Hassler, LER Coordinator                                                                           609-339-1989 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)
The calculations showed that for normal operating conditions, the average shear stress in the key may exceed the material yield stress. The cal< .llations also showed that although the average shear stress in the valve stem was estimated to be less than the miniml..Il"n material yield stress, the fatigue life of the shaft keyway appeared to be limited to a low number of cycles. Key failures in the past are attributed to an overload during normal operation of these valves. A review of the original design revealed that these valves were installed with little or no design margin and the keys are likely failing due to low cycle fatigue with stress levels exceeding yield strength.
CAUSE         SYSTEM     """""""' ....................,.... ,
Corrective action is to replace the valves and a review of Fisher Model 7600 valves for similar concerns.
TONPRDS
This event is reportable in accordance with 1o*crn 73 (a) (2) (ii); any condition that resulted in the plant being outside the design basis of the plant and in accordance with 10 CFR 21.2 (c). NRC FORll 388 (4-116)
                                                                                        ......       ......     __., .........u...                 REPORTABLE TONPRDS ilti11111111:
*
SUPPLEMENTAL REPORT EXPECTED (14)                                                 EXPECTED           MONTH           DAY       YEAR IYES                             .                                                                   SUBMISSION (If yea, complet111 EXPECTED SUBMISSION DATE).                       XINO                            DATE(15)
* NRC FORM 368A (4-86) U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILrTY NAME (1) DOCKET NUMBER f 2) LER NUMBER (6l 05000272 YEAR I I
ABSTRACT (Limit to 1400 apace*, i.e ** approximately 1S aingl-paced typewritten lines) (16)
SALEM GENERATING STATION, UNIT 1 96 -012 -00 TEXT (If more *p*c* is required, use *dditional copies of NRC Form 366A) (17) PLANT AND SYSTEM IDENTIFICATION Westinghouse  
On July 23, 1996 a review determined that the keys/keyways on the actuators f.or the Residual Heat Removal (RHR) flow control valves are subject to failure. The valves were made by Fisher Controls International, Model Type 656/7600. When using the sirrplified Fisher catalog 14 methodology, the calculated maximum stem torque exceeds the vendor specified allowable torque. Preliminary calculations of the shaft torque and resulting average shear stresses in the valve stem key and keyway were also performed. The calculations showed that for normal operating conditions, the average shear stress in the key may exceed the material yield stress. The cal< .llations also showed that although the average shear stress in the valve stem was estimated to be less than the miniml..Il"n material yield stress, the fatigue life of the shaft keyway appeared to be limited to a low number of cycles. Key failures in the past are attributed to an overload during normal operation of these valves. A review of the original design revealed that these valves were installed with little or no design margin and the keys are likely failing due to low cycle fatigue with stress levels exceeding yield strength. Corrective action is to replace the valves and a review of Fisher Model 7600 valves for similar concerns.
-Pressurized Water Reactor Residual Heat Removal System {BP/-}* PAGE (3) 2 OF 4
This event is reportable in accordance with 1o*crn 73 (a) (2) (ii); any condition that resulted in the plant being outside the design basis of the plant and in accordance with 10 CFR 21.2 (c).
* Energy Industry Identification System (EIIS) codes and component function identifier codes appear as (SS/CCC) CONDITIONS PRIOR TO OCCURRENCE At the time of identification Salem Units 1 and 2 were shutdown and defueled.
NRC FORll 388 (4-116)
DESCRIPTION 0F OCCURRENCE On July 23, 1996 a review determined that the keys/keyways on the actuators for the Residual Heat Removal (RHR) flow control valves {BP/FCV} are subject to failure. There are three RHR control valves in each Salem unit. The valves were made by Fisher Controls International, Model Type 656/7600.
 
In 1993, Fisher issued a 10 CFR 21 Notification that indicated Fisher had supplied some with an inadequate key material and recommended that all keys be replaced to ensure that the proper material was installed.
NRC FORM 368A (4-86)
In 1996, during the current shutdown, three valves were found to have failed keyways. An assessment of changing the valve stem packing configuration of one of the RHR flow control valves revealed a potential problem with the valve stem torque load. When using the simplified Fisher Catalog 14 methodology, the calculated maximum stem torque exceeds the vendor specified allowable torque. Preliminary calculations of the shaft torque and resulting average shear stresses in the valve stem key and keyway were also performed.
* LICENSEE EVENT REPORT (LER)
The calculations showed that for normal operating conditions, the average shear stress in the key may exceed the material yield stress. The calculations also showed that although the average shear stress in the valve stem was estimated to be less than the minimum material yield stress, the fatigue life of the shaft keyway appeared to be limited to a low number of cycles. The RHR flow control valves have a history of valve stem to actuator lever arm key failures, stem keyway failures, valve P.osi tioning problems, and packing gland leaks. Since 1993, five incidents of key failures, keyway failures and/or bent stems have been reported.
* U.S. NUCLEAR REGULATORY COMMISSION TEXT CONTINUATION FACILrTY NAME (1)                           DOCKET NUMBERf 2)     LER NUMBER (6l         PAGE (3) 05000272     YEAR I s~cni:mAL  I':a~  2  OF    4 SALEM GENERATING STATION, UNIT 1                                                       96 -   012     -   00 TEXT (If more *p*c* is required, use *dditional copies of NRC Form 366A) (17)
After the initial three key failures, the keys were replaced in all of the valves. Fisher had supplied some valves with an inadequate key material, and recommended that all keys be replaced to ensure that the proper material was installed.
PLANT AND SYSTEM IDENTIFICATION Westinghouse - Pressurized Water Reactor Residual Heat Removal System {BP/-}*
The work order histories do not indicate whether the failed or replaced keys were of the required material.
* Energy Industry Identification System (EIIS) codes and component function identifier codes appear as (SS/CCC)
In 1996, the Salem Unit 1 valves were found to have failed stem keyways. NRC FORM 366A ('4-95)
CONDITIONS PRIOR TO OCCURRENCE At the time of identification Salem Units 1 and 2 were shutdown and defueled.
*
DESCRIPTION 0F OCCURRENCE On July 23, 1996 a review determined that the keys/keyways on the actuators for the Residual Heat Removal (RHR) flow control valves {BP/FCV} are subject to failure. There are three RHR control valves in each Salem unit. The valves were made by Fisher Controls International, Model Type 656/7600.                                         In 1993, Fisher issued a 10 CFR 21 Notification that indicated Fisher had supplied some valve~ with an inadequate key material and recommended that all keys be replaced to ensure that the proper material was installed.                                 In 1996, during the current shutdown, three valves were found to have failed keyways.
* NRC FORM 366A (4-115) U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER 8) 05000272 SALEM GENERATING STATION, UNIT 1 TEXT (If more apace ia required, uae additional copies of NRC Form 366A) (17) CAUSE OF OCCURRENCE YEAR I SEQUENTIAL NUMBER 96 -012 REVISION NUMBER 00 PAGE (3) 3 OF 4 The key were caused by an overload during normal operation of these valves. A review of the original Westinghouse design revealed that these valves were installed with little or no design margin and the keys are likely failing due to low cycle fatigue with stress levels exceeding yield strength.
An assessment of changing the valve stem packing configuration of one of the RHR flow control valves revealed a potential problem with the valve stem torque load. When using the simplified Fisher Catalog 14 methodology, the calculated maximum stem torque exceeds the vendor specified allowable torque.                                         Preliminary calculations of the shaft torque and resulting average shear stresses in the valve stem key and keyway were also performed. The calculations showed that for normal operating conditions, the average shear stress in the key may exceed the material yield stress. The calculations also showed that although the average shear stress in the valve stem was estimated to be less than the minimum material yield stress, the fatigue life of the shaft keyway appeared to be limited to a low number of cycles.
Additionally, the valves have experienced increased torque demands since installation.
The RHR flow control valves have a history of valve stem to actuator lever arm key failures, stem keyway failures, valve P.osi tioning problems, and packing gland leaks.           Since 1993, five incidents of key failures, keyway failures and/or bent stems have been reported. After the initial three key failures, the keys were replaced in all of the valves.                           Fisher had supplied some valves with an inadequate key material, and recommended that all keys be replaced to ensure that the proper material was installed. The work order histories do not indicate whether the failed or replaced keys were of the required material.                                           In 1996, the Salem Unit 1 valves were found to have failed stem keyways.
The change from asbestos to graphite packing (without an appropriate design review) significantly increased the packing friction loads and it is probable that wear due to the high stress has increased the friction in the valve bearings.
NRC FORM 366A ('4-95)
This provided an opportunity to discover this problem and take corrective actions at that time. A review is in progress to identify other Fisher Model 7600 valves that may be subject to the same conditions of design inadequacies and/or increased torque demands resulting from*packing changes. PRIOR SIMILAR OCCURRENCES In the past two years there were four LERs that attributed their root cause to design deficiencies.
 
However, these four LERs are not considered similar in nature to the RHR flow control valves. SAFETY CONSEQUENCES AND IMPLICATIONS The RHR flow control valves are spring-to-open and air-to-close.
NRC FORM 366A (4-115)
This fail open design ensures that flow through the valve is maintained in the event of loss of air supply to the actuator.
* LICENSEE EVENT REPORT (LER)
Complete failure of the key connection between the valve stem and actuator lever arm would result in loss of the spring function and the valve moving to a near closed position if the hydrodynamic loads are sufficient to overcome the frictional loads at the disk angle position.
* U.S. NUCLEAR REGULATORY COMMISSION TEXT CONTINUATION FACILITY NAME (1)                           DOCKET NUMBER (2)     LER NUMBER 8)            PAGE (3) 05000272     YEAR I  SEQUENTIAL NUMBER REVISION NUMBER  3  OF    4 SALEM GENERATING STATION, UNIT 1                                                       96 -      012      00 TEXT (If more apace ia required, uae additional copies of NRC Form 366A) (17)
The valve would not go fully closed due to the non-symmetric flow profile resulting from the upstream elbow. Partial flow will pass through the valve in the near closed position since the valve is not a leak tight design. If a single failure (in the closed position) of the RHR flow control valves were to occur, the other 100% train would be available to accomplish the safety function.
CAUSE OF OCCURRENCE The key failur~s were caused by an overload during normal operation of these valves. A review of the original Westinghouse design revealed that these valves were installed with little or no design margin and the keys are likely failing due to low cycle fatigue with stress levels exceeding yield strength.
If the bypass valve were to fail closed, both of the loops would be available to throttle the injection flow. In the event both trains were affected by this failure, the operators would enter the abnormal procedure for loss of RHR and take the appropriate actions. NRC FORM 386A (-4-95) c
Additionally, the valves have experienced increased torque demands since installation. The change from asbestos to graphite packing (without an appropriate design review) significantly increased the packing friction loads and it is probable that wear due to the high stress has increased the friction in the valve bearings. This provided an opportunity to discover this problem and take corrective actions at that time.
* NRC FORM 366A (4-95) U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) 05000272 YEAR l SEQUENTIAL I REVISION NUMBER NUMBER SALEM GENERATING STATION, UNIT 1 96 -012 -00 TEXT (If more *pace i* required, ume *dditional copi** of NRC Form 366A) (17) CORRECTIVE ACTIONS PAGE (3) 4 OF 4 1. The RHR flow control valves will be replaced to Mode 6 in Salem Units 1 and 2. 2. A review is in progress to identify other Fisher Model 7600 valves that may be subject to the same conditions of design inadequacies and/or increased torque demands resulting from packing changes. The review will be completed prior to Mode 6. Corrective actions, if any, will be completed to support system operability in accordance with Technical Specifications requirements.
A review is in progress to identify other Fisher Model 7600 valves that may be subject to the same conditions of design inadequacies and/or increased torque demands resulting from*packing changes.
10CFR21 REPORTING 10CFR21 reporting requirements are met by this LER. NRC FORM 36eA (4-95)}}
PRIOR SIMILAR OCCURRENCES In the past two years there were four LERs that attributed their root cause to design deficiencies.                 However, these four LERs are not considered similar in nature to the RHR flow control valves.
SAFETY CONSEQUENCES AND IMPLICATIONS The RHR flow control valves are spring-to-open and air-to-close. This fail open design ensures that flow through the valve is maintained in the event of loss of air supply to the actuator. Complete failure of the key connection between the valve stem and actuator lever arm would result in loss of the spring function and the valve moving to a near closed position if the hydrodynamic loads are sufficient to overcome the frictional loads at the disk angle position. The valve would not go fully closed due to the non-symmetric flow profile resulting from the upstream elbow.                     Partial flow will pass through the valve in the near closed position since the valve is not a leak tight design.                                       If a single failure (in the closed position) of the RHR flow control valves were to occur, the other 100% train would be available to accomplish the safety function.                                       If the bypass valve were to fail closed, both of the loops would be available to throttle the injection flow.             In the event both trains were affected by this failure, the operators would enter the abnormal procedure for loss of RHR and take the appropriate actions.
NRC FORM 386A (-4-95)
 
c NRC FORM 366A (4-95)
* LICENSEE EVENT REPORT (LER)
U.S. NUCLEAR REGULATORY COMMISSION TEXT CONTINUATION FACILITY NAME (1)                           DOCKET NUMBER (2)     LER NUMBER (6)              PAGE (3) 05000272     YEAR l SEQUENTIAL NUMBER IREVISION NUMBER 4  OF    4 SALEM GENERATING STATION, UNIT 1 96 -     012     -     00 TEXT (If more *pace i* required, ume *dditional copi** of NRC Form 366A) (17)
CORRECTIVE ACTIONS
: 1. The RHR flow control valves will be replaced                               pr~or  to Mode 6 in Salem Units 1 and 2.
: 2. A review is in progress to identify other Fisher Model 7600 valves that may be subject to the same conditions of design inadequacies and/or increased torque demands resulting from packing changes. The review will be completed prior to Mode 6. Corrective actions, if any, will be completed to support system operability in accordance with Technical Specifications requirements.
10CFR21 REPORTING 10CFR21 reporting requirements are met by this LER.
NRC FORM 36eA (4-95)}}

Latest revision as of 05:28, 3 February 2020

LER 96-012-00:on 960723,potential Loss of Residual Capability Identified.Caused by Inadequate Valve Design.Rhr Flow Control Valves replaced.W/960822 Ltr
ML18102A335
Person / Time
Site: Salem PSEG icon.png
Issue date: 08/22/1996
From: Garchow D, Hassler D
Public Service Enterprise Group
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-96-012-01, LER-96-12-1, LR-N96253, NUDOCS 9608270273
Download: ML18102A335 (6)
v • d • e


Text

PS~G Public Service Electric and Gas Comj'.?any P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nuclear Business Unit AUG 2 21996 LR-N96253 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:

LER 272/96-012-00 SALEM GENERATING STATION - UNIT 1 FACILITY OPERATING LICENSE NO. DPR-70 DOCKET NO. 50-272 This Licensee Event Report entitled "Potential Loss of Residual Heat Removal Capability Due to Inadequate Valve Design" is being submitted pursuant to the requirements of the Code of Federal Regulations 10CFR50.73(a) (2) (ii) (B).

rr;e;~

David F. Garchow General Manager -

Salem Operations Attachment DVH/tcp

.C Distribution LER File 3.7 1 n'-' ~Y*m?r i~ ~n ,.l t; :- :;. :r.L'..'. .

95-2168 REV. 6/94

Document Control Desk LR-N96253 Attachment A The following represents the commitments that Public Service Electric

& Gas (PSE&G) made to the Nuclear Regulatory Commission (NRC) relative to this LER (272/96-012-00). The commitments are as follows:

1. The RHR flow control valves will be replaced prior to Mode 6 in Salem Units 1 and 2.
2. A review is in progress to identify other Fisher Model 7600 valves that may be subject to the same conditions of design inadequacies and/or increased torque demands resulting from packing changes.

The review will be completed prior to Mode 6. Corrective actions, if any, will be completed to support system operability in accordance with Technical Specifications requirements.

NRCFORM388 (4-95)

U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)

  • APPROVED BY OMB NO. 3150-0104 EXPIRES CM/30198 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS MANDATORY INFORMATION COLLECTION REQUEST: 50.0 HRS.

REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE LICENSING PROCESS AND FED BACK TO INDUSTRY. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE.INFORMATION AND RECORDS MANAGEMENT BRANCH (T.e ~NUCLEAR (See reverse for required number of REGULATORY COMMISSION, WASHINGTON, DC , AND TO THE PAPERWORK REDUCTION PROJECT ~160-0104), OFFICE OF digits/characters for each block) MANAGEMENT AND BUDGET, WASHINGTON, 20l503.

  • FACILITY NAME (1) DOCKET NUMBER (2) PAGE(3)

SALEM GENERATING STATION, UNIT 1 05000272 1 OF4 TITLE (4)

Potential Loss of Residual Heat Removal Capability Due To Inadequate Valve Design EVENT DATE (5) LER NUMBER (8) REPORT DATE , 1) OTHER FACILITIES INVOLVED (8)

IFACILITY NAME DOCKET NUMBER llONTH DAY YEAR SEQUENTIAL 'REVISION MONTH DAY YEAR YEAR I NUllBER NUMBER SALEM UNIT2 05000311 07 23 96 96 - 012 - 00 08 22 96 FACILITY NAME DOCKET NUMBER 05000 OPERATING N THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check one or more) (11)

MODE(9) 20.2201(b) 20.2203(a)(2)(v) so. 73(*)(2)(i) SO. 73(a)(2)(viii)

POWER 000 20.2203(a)(1) 20.2203(a)(3)(i) x SO. 73(a)(2)(ii) SO. 73(a)(2)(x)

LEVEL (10) 20.2203(a)(2)(i) 20.2203(a)(3)(ii) SO. 73(a)(2)(iii) 73.71 20.2203(a)(2)(ii) 20.2203(*)(4) SO. 73(a)(2)(iv) x OTHER 20.2203(a)(2)(iii) S0.38(c)(1) SO. 73(a)(2)(v) Spec~ln Abetract below or In C Form 366A 20.2203(a)(2)(iv) S0.38(c)(2) 50. 73(a)(2)(vii) Part 21 LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER (Include Are* Code)

Dennis v. Hassler, LER Coordinator 609-339-1989 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE SYSTEM """""""' ....................,.... ,

TONPRDS

...... ...... __., .........u... REPORTABLE TONPRDS ilti11111111:

SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED MONTH DAY YEAR IYES . SUBMISSION (If yea, complet111 EXPECTED SUBMISSION DATE). XINO DATE(15)

ABSTRACT (Limit to 1400 apace*, i.e ** approximately 1S aingl-paced typewritten lines) (16)

On July 23, 1996 a review determined that the keys/keyways on the actuators f.or the Residual Heat Removal (RHR) flow control valves are subject to failure. The valves were made by Fisher Controls International, Model Type 656/7600. When using the sirrplified Fisher catalog 14 methodology, the calculated maximum stem torque exceeds the vendor specified allowable torque. Preliminary calculations of the shaft torque and resulting average shear stresses in the valve stem key and keyway were also performed. The calculations showed that for normal operating conditions, the average shear stress in the key may exceed the material yield stress. The cal< .llations also showed that although the average shear stress in the valve stem was estimated to be less than the miniml..Il"n material yield stress, the fatigue life of the shaft keyway appeared to be limited to a low number of cycles. Key failures in the past are attributed to an overload during normal operation of these valves. A review of the original design revealed that these valves were installed with little or no design margin and the keys are likely failing due to low cycle fatigue with stress levels exceeding yield strength. Corrective action is to replace the valves and a review of Fisher Model 7600 valves for similar concerns.

This event is reportable in accordance with 1o*crn 73 (a) (2) (ii); any condition that resulted in the plant being outside the design basis of the plant and in accordance with 10 CFR 21.2 (c).

NRC FORll 388 (4-116)

NRC FORM 368A (4-86)

  • LICENSEE EVENT REPORT (LER)
  • U.S. NUCLEAR REGULATORY COMMISSION TEXT CONTINUATION FACILrTY NAME (1) DOCKET NUMBERf 2) LER NUMBER (6l PAGE (3) 05000272 YEAR I s~cni:mAL I':a~ 2 OF 4 SALEM GENERATING STATION, UNIT 1 96 - 012 - 00 TEXT (If more *p*c* is required, use *dditional copies of NRC Form 366A) (17)

PLANT AND SYSTEM IDENTIFICATION Westinghouse - Pressurized Water Reactor Residual Heat Removal System {BP/-}*

  • Energy Industry Identification System (EIIS) codes and component function identifier codes appear as (SS/CCC)

CONDITIONS PRIOR TO OCCURRENCE At the time of identification Salem Units 1 and 2 were shutdown and defueled.

DESCRIPTION 0F OCCURRENCE On July 23, 1996 a review determined that the keys/keyways on the actuators for the Residual Heat Removal (RHR) flow control valves {BP/FCV} are subject to failure. There are three RHR control valves in each Salem unit. The valves were made by Fisher Controls International, Model Type 656/7600. In 1993, Fisher issued a 10 CFR 21 Notification that indicated Fisher had supplied some valve~ with an inadequate key material and recommended that all keys be replaced to ensure that the proper material was installed. In 1996, during the current shutdown, three valves were found to have failed keyways.

An assessment of changing the valve stem packing configuration of one of the RHR flow control valves revealed a potential problem with the valve stem torque load. When using the simplified Fisher Catalog 14 methodology, the calculated maximum stem torque exceeds the vendor specified allowable torque. Preliminary calculations of the shaft torque and resulting average shear stresses in the valve stem key and keyway were also performed. The calculations showed that for normal operating conditions, the average shear stress in the key may exceed the material yield stress. The calculations also showed that although the average shear stress in the valve stem was estimated to be less than the minimum material yield stress, the fatigue life of the shaft keyway appeared to be limited to a low number of cycles.

The RHR flow control valves have a history of valve stem to actuator lever arm key failures, stem keyway failures, valve P.osi tioning problems, and packing gland leaks. Since 1993, five incidents of key failures, keyway failures and/or bent stems have been reported. After the initial three key failures, the keys were replaced in all of the valves. Fisher had supplied some valves with an inadequate key material, and recommended that all keys be replaced to ensure that the proper material was installed. The work order histories do not indicate whether the failed or replaced keys were of the required material. In 1996, the Salem Unit 1 valves were found to have failed stem keyways.

NRC FORM 366A ('4-95)

NRC FORM 366A (4-115)

  • LICENSEE EVENT REPORT (LER)
  • U.S. NUCLEAR REGULATORY COMMISSION TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER 8) PAGE (3) 05000272 YEAR I SEQUENTIAL NUMBER REVISION NUMBER 3 OF 4 SALEM GENERATING STATION, UNIT 1 96 - 012 00 TEXT (If more apace ia required, uae additional copies of NRC Form 366A) (17)

CAUSE OF OCCURRENCE The key failur~s were caused by an overload during normal operation of these valves. A review of the original Westinghouse design revealed that these valves were installed with little or no design margin and the keys are likely failing due to low cycle fatigue with stress levels exceeding yield strength.

Additionally, the valves have experienced increased torque demands since installation. The change from asbestos to graphite packing (without an appropriate design review) significantly increased the packing friction loads and it is probable that wear due to the high stress has increased the friction in the valve bearings. This provided an opportunity to discover this problem and take corrective actions at that time.

A review is in progress to identify other Fisher Model 7600 valves that may be subject to the same conditions of design inadequacies and/or increased torque demands resulting from*packing changes.

PRIOR SIMILAR OCCURRENCES In the past two years there were four LERs that attributed their root cause to design deficiencies. However, these four LERs are not considered similar in nature to the RHR flow control valves.

SAFETY CONSEQUENCES AND IMPLICATIONS The RHR flow control valves are spring-to-open and air-to-close. This fail open design ensures that flow through the valve is maintained in the event of loss of air supply to the actuator. Complete failure of the key connection between the valve stem and actuator lever arm would result in loss of the spring function and the valve moving to a near closed position if the hydrodynamic loads are sufficient to overcome the frictional loads at the disk angle position. The valve would not go fully closed due to the non-symmetric flow profile resulting from the upstream elbow. Partial flow will pass through the valve in the near closed position since the valve is not a leak tight design. If a single failure (in the closed position) of the RHR flow control valves were to occur, the other 100% train would be available to accomplish the safety function. If the bypass valve were to fail closed, both of the loops would be available to throttle the injection flow. In the event both trains were affected by this failure, the operators would enter the abnormal procedure for loss of RHR and take the appropriate actions.

NRC FORM 386A (-4-95)

c NRC FORM 366A (4-95)

  • LICENSEE EVENT REPORT (LER)

U.S. NUCLEAR REGULATORY COMMISSION TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) 05000272 YEAR l SEQUENTIAL NUMBER IREVISION NUMBER 4 OF 4 SALEM GENERATING STATION, UNIT 1 96 - 012 - 00 TEXT (If more *pace i* required, ume *dditional copi** of NRC Form 366A) (17)

CORRECTIVE ACTIONS

1. The RHR flow control valves will be replaced pr~or to Mode 6 in Salem Units 1 and 2.
2. A review is in progress to identify other Fisher Model 7600 valves that may be subject to the same conditions of design inadequacies and/or increased torque demands resulting from packing changes. The review will be completed prior to Mode 6. Corrective actions, if any, will be completed to support system operability in accordance with Technical Specifications requirements.

10CFR21 REPORTING 10CFR21 reporting requirements are met by this LER.

NRC FORM 36eA (4-95)

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