ML18101A940

LER 91-030-01:on 910920,both Pressurizer PORVs Failed Operability Check.Caused by Inability of Buna-N Diaphragm Matl to Maintain Physical Properties.Installed New Buna-N Fabric Reinforced Diaphragm Matl in Unit 1.W/950828 Ltr

ML18101A940
Person / Time
Site:	Salem
Issue date:	08/25/1995
From:	Lambert C, Warren C Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-91-030, LER-91-30, NUDOCS 9509010010
Download: ML18101A940 (7)

Text

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Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nuclear Business Unit AUG 2 8 1995 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Attn: Document Control Desk SALEM GENERATING STATION LICENSE NO. DPR-70 DOCKET NO. 50-272 UNIT NO. 1 SUPPLEMENTAL L!CENSEE EVENT REPORT 91-030-01 This LER supplement is being submitted pursuant to the requirements of Code of Federal Regulation 10CFR50.73. It is intended to satisfy the intent of the original submittal to update the report apparent cause of occurrence and corrective action sections.

Sincerely,

.f?fL (~ tJruu_

4c. Warren General Manager -

Salem Operations SORC Mtg.95-097 MJPJ C Distribution LER File 3.7.1 i . *! :"\ ,.., F. ....,

• . ..:.. 1.,.' 1..: J J 9509010010 950825 PDR ADOCK 05000272 s .--**-*------.- *-*-*--*--- *- -- -*----'~1>.R~-* -- .. . ..

.The power is in yow- hands.

l NRC FORM 366 U.S. NUC R REGULATORY COMMISSION A OVED BY OMB NO. 3150-0104 (4-95) EXPIRES 04/30/98 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS MANDATORY INFORMATION COLLECTION REQUEST: 60.0 HRS.

REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE LICENSEE EVENT REPORT (LER) LICENSING PROCESS AND FED BACK TO INDUSTRY. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (T-8 F33), U.S. NUCLEAR (See reverse for required number of REGULATORY COMMISSION, WASHINGTON, DC 2055~, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF digits/characters for each block) MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAllE (1) DOCKET NUllBER (2) PAGE (3)

SALEM GENERATING STATION UNIT 1 05000272 1 OF6 TITLE (4)

BOTH PRESSURIZER PRESSURE OPERATED RELIEF VALVES FAILED AN OPERABILITY CHECK EVENT DATE (5) . LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8)

FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR I SEQUENTIAL I REVISION MONTH DAY YEAR NUMBER NUMBER SALEM UNIT2 05000311 9 20 91 9 1

- 030

-0 1 8 2 5 95 FACILITY NAME DOCKET NUMBER 05000 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check one or more) (11)

MODE (9) 4 20.2201(b) 20.2203(a)(2)(v) 50. 73(a)(2)(i) 50. 73(a)(2)(viii)

POWER 20.2203(a)(1) 20.2203(a)(3)(i) 50. 73(a)(2)(ii) 50. 73(a)(2)(x)

LEVEL (10) 000 20.2203(a)(2)(i) 20.2203(a)(3)(ii) 50. 73(a)(2)(iii) 73.71 20.2203(a)(2)(ii) 20.2203(a)(4) 50. 73(a)(2)(iv) OTHER 20.2203(a)(2)(iii) 50.36(c)(1) x 50. 73(a)(2)(v) Spec~ln or In Abstract below C Form 366A 20.2203(a)(2)(iv) 50.36(c)(2) x 50. 73(a)(2)(vii)

LICENSEE CONTACT FOR THIS t.:ER (12)

NAME TELEPHONE NUMBER (Include Area Code)

Craig Lambert 609/339-1848 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED, IN TlilS R~PORT (13)

CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE TONPRDS TONPRDS I

B AB PCV C635 y B AB PCV C635 y SUPPLEMENTAL REPORT EXPECTED (14)' EXPECTED MONTH DAY YEAR IYES (If yes, complete EXPECTED SUBMISSION DATE). xi NO SUBMISSION DATE (15)

ABSTRACT (limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)

On 9/20/91, during a Unit 1 shutdown to support a maintenance outage, both Pressurizer Pressure Operated Relief Valves {PORVs) lPRl and 1PR2, failed to open upon demand.

Visual inspection revealed operation of the valves was prevented as the result of air leakage from the valves' actuator diaphragm enclosures. This was due to loosened fasteners on the enclosures. The cause of this event was inability of the Buna-N diaphragm material to maintain its physical properties and subsequent loss of fastener pre load coupled with loosening of the fasteners due to pipeline induced vibration

{unknown at time of event) . Inunediate corrective action was installation of new Buna-N fabric reinforced diaphragm material in the Unit 1 PORVs and Pressurizer Spray Valves {PSCVs). In 1992, EPIM/fabric reinforced diaphragm material and a back-up 0-ring were installed in the PORVs and PSCVs of both Units until a more reliable diaphragm material could be installed. In 1993, silicone rubber/fabric reinforced diaphragm material and prevailing torque type replacement' fasteners that are frictionally resistant to rotation were installed in these valves on both Units. This event is reportable pursuant to 10CFR50.73{a) (2) {v) {D) and {a) (2) {vii) {D)

• NRC FORM 366 (4-95)

SUPPLEMENTAL LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 1 5000272 91-030-01 2 of 6 PLANT AND SYSTEM IDENTIFICATION:

Westinghouse - Pressurized Water Reactor Energy Industry Identification System and Institute of Electrical and Electronic Engineers, Inc. component function identifier codes are identified in the text as {XX/XX}

IDENTIFICATION OF OCCURRENCE:

Both Pressurizer Pressure Operated Relief Valves (PORVs) Failed An Operability Check Event Date: September 20, 1991 Supplemental Report Date: August 25, 1995 Prior Report Date: October 18, 1991 The original report was initiated by Incident Report No.91-659.

This submittal is intended to update the event apparent- cause and corrective actions.

CONDITIONS PRIOR TO OCCURRENCE:

Mode 4 (Hot Shutdown); Unit preparing to go to Mode 5 (Cold Shutdown)

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DESCRIPTION OF OCCURRENCE:

On September 20, 1991, a plant shutdown to Mode 5 was in progress to support a maintenance outage. In accordance with Technical Specification Surveillance 4.4.9.3.1.1, the PORVs (lPRl and 1PR2)

{AB/PCV} were functionally checked, at 0115 hours0.00133 days <br />0.0319 hours <br />1.901455e-4 weeks <br />4.37575e-5 months <br /> (same day) using procedure II-2.3.4, "Pressurizer Overpressure Protection -

Operability". During this surveillance, both valves failed to open upon demand. lPRl did not lose its closed limit indication and although 1PR2 indicated movement, the Control Room indication did not show the valve reaching full open. At the time of the valves*

functional check, the Control Room overhead alarm for PORV accumulator low air pressure actuated.

Investigation revealed air leakage from the diaphragm enclosure flange. This air pressure loss had prevented both valves from opening. Visual observation determined that loss of air pressure resulted from loosened fastening cap screws/nuts on the valves' air actuator diaphragm enclosures {AB/AHU}. Following this discovery, the fasteners of each diaphragm (12 hole design diaphragm in use at the time of this occurrence) were tightened and both valves were successfully stroked. The Nuclear Regulatory Commission (NRC) was

SUPPLEMENTAL LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 1 5000272 91-030-01 3 of 6 DESCRIPTION OF OCCURRENCE: cont'd notified of this event on September 20, 1991, at 0752 hours0.0087 days <br />0.209 hours <br />0.00124 weeks <br />2.86136e-4 months <br />, in accordance with 10CFR50.72(b) (2) (iii) (D).

APPARENT CAUSE OF OCCURRENCE:

The initiating cause of both PORVs failing to open was loosening of the fasteners in the valves' actuator diaphragm enclosures. The diaphragm material, Buna-N, Copes Vulcan Inc. (CVI) Part No. 182578, incurred "creep" where the diaphragm changed from its original geometry under load and over time. Elevated ambient temperature degraded the Buna-N diaphragm material resulting in the material taking a permanent set, a loss of resilience, and extrusion of the material from the clamped region. Permanent thinning of the material in the clamped region caused loss of fastener preload. This was observed as loosening of the diaphragm fastener cap screws/nuts, which allowed a leakage pathway for the diaphragm control air.

ANALYSIS OF OCCURRENCE:

Based on available industry standards, Buna-N material can be used in a temperature range between -40 degrees Fahrenheit (F) to +250 degrees F. Following this occurrence, CVI was contacted and requested to confirm a Salem Engineering conclusion that the actuator diaphragm material was acceptable for the pressurizer environment of elevated temperatures to approximately 170 degrees F. In a letter, dated October 2, 1991, CVI responded to this request as follows:

"Our review indicates the supplied equipment is suitable for use at the design conditions stated in the original equipment purchase specification. This purchase specification required that the equipment be suitable for use at a maximum temperature of 120°F. Our review indicates that deterioration of the Buna-N diaphragm should not occur due to exposure to 110°F for a period of 18 months. Unfortunately, Copes-Vulcan has no documented field usage under service conditions similar to your (PSE&G) service conditions and no certified testing has been performed which would allow us to guarantee that acceptable performance can be maintained for a period of 18 months at 170°F. 11 CVI was then requested to attempt to certify the diaphragm material as acceptable at elevated temperatures. However, due to subsequent economic considerations of comp le.ting this certification and ongoing material design improvements, the certification of Buna-N by CVI was not completed. The diaphragm material used in these actuators was changed to EPDM in 1992, then to silicone rubber in 1993 (see CORRECTIVE ACTION section).

SUPPLEMENTAL LICENSEE EVENT REPORT (LER} TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 1 5000272 91-030-01 4 of 6 ANALYSIS OF OCCURRENCE: cont'd Failure analysis, review of operational inputs and valve/actuator history, industry experience, and extensive laboratory testing has been conducted to identify contributors to diaphragm leakage. This effort has concluded the observed PORV failures and a subsequent malfunction of a Unit 2 Pressurizer Spray Valve {PSCV} 2PS1 {AB/PCV},

in 1992, were due to the interaction and synergistic effect of the employed diaphragm material's (Buna-N and EPDM} inability to maintain physical properties in the elevated ambient temperature of the Pressurizer enclosure area coupled with diaphragm enclosure fastener loosening. This loosening was influenced by pipeline induced vibration in the Pressurizer enclosure area. The PORVs were compared to other CVI supplied valves not exhibiting similar deficiencies at Salem Generating station and other industrial applications. Salem Units 1 and 2 have approximately 125 CVI supplied valves of a similar design. Of this total, four (4) valves per Unit have exhibited the diaphragm leakage deficiency. These valves (PORVs and PSCVs} are located inside the pressurizer enclosure area. This experience indicates a temperature induced failure mode. In addition, test results have identified deficiencies with EPDM/fabric reinforced diaphragm material. This material was installed in 1992. Silicone rubber/fabric reinforced diaphragm material testing has determined superior performance for these applications. The material lost essentially none of its physical properties during these tests.

SAFETY SIGNIFICANCE:

This event is reportable in accordance with 10CFR 50.73(a} (2) (v) (D) and 50.73(a)(2) (vii} (D). With both PORVs inoperable, the plant was in a condition that alone could have prevented the fulfillment of a safety function needed to mitigate the consequences of an accident.

Use of the PORVs, pressurizer spray, and pressurizer auxiliary spray helps to mitigate the consequences of a steam generator tube rupture accident. Additionally, the PORVs are specifically taken credit for in a loss of all feedwater accident, which is not addressed by the Updated Final Safety Analysis Report as it is considered beyond the plants' design basis. However, loss of all feedwater is addressed by the Westinghouse Emergency Response Guidelines (ERGs}. During a loss of all feedwater accident, reactor pressure control would be maintained by utilizing the PORVs and the safety injection pumps.

This is addressed in the station Emergency Operating Procedures (EOPs) in FRHS-1, "Functional Restoration of Heat Sink".

At the time of the occurrence, the Unit was in Mode 4 with RCS temperature greater than 312 degrees F which is greater than the temperature at which POPS is activated. Prior to reducing temperature below 312 degrees F, the fasteners were tightened and both valves were successfully stroked and returned to service.

SUPPLEMENTAL LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 1 5000272 91-030-01 5 of 6 SAFETY SIGNIFICANCE: cont'd Therefore, the PORVs were not being depended upon for low temperature overpressure protection. Based on the above, the safety significance of this occurrence was considered minimal.

Unit 2 was operating at 100% power. As a result of the Unit 1 PORV failures, the Unit 2 PORVs were stroke tested and found to be operating satisfactorily.

PRIOR SIMILAR OCCURRENCES:

A review of plant records (work order documentation related to work history, completed work orders and valve calibration cards) was conducted which covered a period from the late 1970's through September 1991. This review revealed a trend of frequent diaphragm replacement and required tightening of the diaphragm enclosure fasteners. In addition, this effort identified that the diaphragm replacement procedure did not specify the torque values for the fasteners. The majority of these work orders involved prev'entive and corrective maintenance measures prior to reclassification of the valve actuator to safety related as a result of Generic Letter 90-06.

CORRECTIVE ACTION:

After the initial failure in September 1991, the diaphragm of the Unit 1 PORVs (PRl and PR2) and the PSCVs (PSl and PS3), which comprised all CVI actuators located inside the Unit 1 Pressurizer enclosure, were replaced in-kind. Procedure SC.IC-PM.RC-0001 was used for completing the diaphragm replacement. Salem Engineering assessed the capability of Buna-N material and determined the material was acceptable for use at elevated temperatures.

In 1992, during the Unit 1 refueling/maintenance outage lRlO, the original 12 bolt fastener pattern of the PORV and PSCV actuators were replaced with 24 bolt fastener pattern actuators. This was done to reduce uneven diaphragm loading by increasing the number of bolts used. As part of the Design Change Packages (DCPs), the diaphragm was changed to EPDM/fabric reinforced diaphragm material, CVI Part No. 264331. Also included was implementation of strict controls on the diaphragm assembly procedure. These changes were in accordance with Copes-Vulcan recommendations and based upon the ambient temperature of the actuators environment (approximately 170 degrees F). These design changes were intended to improve the joint.

integrity of the diaphragm enclosure by providing a material capable of operating at higher sustained temperatures and a fastener pattern that reduced the distance between the cap screws/nuts, thereby applying more consistent clamping force over the entire perimeter of the diaphragm enclosure. These DCPs were installed on Unit 2 for the PORVs and PSCVs during refueling outage 2R6 (11/91 - 5/92).

SUPPLEMENTAL LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 1 5000272 91-030-01 6 of 6 CORRECTIVE ACTION: cont'd Three (3) weeks after Unit 2 restart, PSCV 2PS1 failed to stroke on demand. Inspection revealed excessive air leakage at the valve actuator diaphragm enclosure interface seal area and uniform loosening of the fasteners. An assessment led to the conclusion that the material had reached its thermoset temperature in the stressed gasket region. The compressive load caused the EPDM material in the clamped region of the diaphragm to permanently thin, with an attendant loss of fastener load and gasket sealing force.

A backup 0-ring was installed on the actuator diaphragms of the PORVs and the PSCVs of Units 1 and 2 to act as a secondary sealing mechanism. The installed o-ring was 9/16" diameter made of Buna-N material, rated for 250 degrees F and was located on the pressurized side (top) of the diaphragm, centered adjacent to the diaphragm sealing bead. This resulted in a slight interference fit that "captured" the 0-ring between the inside radius surface of the diaphragm cover and base when torquing the diaphragm enclosure fasteners. Additionally, monitoring devices were installed to record diaphragm enclosure and ambient temperatures. This change was intended to serve as an interim measure (temporary 18-month usage) until the next refueling outages (lRll and 2R7) when a permanent resolution would be installed. This interim measure remained in effect for the PORVs of Units 1 and 2 for one fuel cycle, with no identified failures.

In 1993, during refueling outages lRll and 2R7, the diaphragm material and enclosure fasteners were replaced to improve diaphragm to enclosure joint integrity and reliability. The replacement diaphragms were silicone rubber/fabric reinforced material, CVI Part No. 194545. The replacement fasteners were 3/8 11 X 16 UNC cap screws and prevailing torque type nuts, CVI Part Nos. 296986 and 343810.

These replaced CVI's standard commercial fasteners, cap screws and nuts, 3/8" X 16 Unified National Coarse (UNC), CVI Part Nos. 004377 and 041390. The replacement fasteners are in accordance with International Fastener Institute 100/107, and features a lock-type nut that its frictionally resistant to rotation, due to a self-containing prevailing torque feature and do not rely upon the compressive load developed on the cap screw thread surfaces.

Since the installation of the silicon rubber/fabric* reinforced material diaphragm and the new fasteners, there have been no failures of these valves due to loosening fasteners or degraded diaphragm material.

A preventive maintenance task has been established to replace the PORV and PSCV actuator diaphragms during each refueling outage. The replaced diaphragms will be examined for degradation. Based on the examinations, the replacement cycle will be re-evaluated.

SORC Mtg.95-097