ML17263A983

LER 95-001-00:on 950203,pressurizer Safety Valves Lift Settings Found Above TS Tolerance During post-svc Test,Due to Setpoint Shifts That Resulted in Independent Trains Being Considered Inoperable

ML17263A983
Person / Time
Site:	Ginna
Issue date:	03/06/1995
From:	St Martin J ROCHESTER GAS & ELECTRIC CORP.
To:
Shared Package
ML17263A982	List: ML17263A981 ML17263A983
References
LER-95-001, LER-95-1, NUDOCS 9503160241
Download: ML17263A983 (9)

Text

NRC FORH 366 .S. NUCLEAR REGULATORY COHHISSION PPROVED BY OHB NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COHPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.

LICENSEE EVENT REPORT (LER) FORWARD COHMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (HNBB 7714), U.ST NUCLEAR REGULATORY COHHISSION, (See reverse for required number of digits/characters for each block) WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF HANAGEHENT AND BUDGET WASHINGTON DC 20503.

FACILITY NAME (1) R. E. Ginna Nuclear Power Plant DOCKET NUMBER (2) PAGE (3) 05000244 10F9 TITLE (4) Pressurizer Safety Valves Lift Settings Found Above Technical Specifications Tolerance During Post-service Test, Due to Setpoint Shifts, Results in Independent Trains Being Considered Inoperable EVENT DATE (5) LER NUHBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8)

SEQUENTIAL REVISION FACILITY NAHE DOCKET NUMBER HONTH DAY YEAR YEAR MONTH DAY YEAR NUHBER NUHBER 02 03 95 95 --001-- 00 03 06 FACILITY NAME DOCKET NUMBER OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 5: (Check one or more) (11)

HODE (9) N 20.402(b) 20.405(c) 50.73(a)(2)(iv) 73.71(b)

POWER 20.405(a )(1)(i) 50.36(c)(1) 50.73(a)(2)(v) 73.71(c) 098 LEVEL (10) 20.405(a)(1)(ii) 50 '6(c)(2) X 50.73(a)(2)(vii) OTHER 20.405(a)(1)(iii) 50.73(a)(2)(i) 50.73(a)(2)(viii)(A) (Specify in 20.405(a)(1)(iv) 50.73(a)(2)(ii) 50.73(a)(2)(viii)(B) Abstract below and in Text, 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 John T. St. Hartin - Technical Assistant TELEPHONE NUMBER (Include Area Code)

(315) 524-4446 COHPLETE ONE LINE FOR EACH COHPONENT FAILURE DESCRIBED IN THIS REPORT (13)

REPORTABLE REPORTABLE CAUSE SYSTEH COMPONENT HANUFACTURER CAUSE SYSTEH COMPONENT MANUFACTURER TO NPRDS TO NPRDS RV C170 SUPPLEHENTAL REPORT EXPECTED (14) EXPECTED MONTH DAY YEAR YES SUB HISS ION (If yes, complete EXPECTED SUBHISSION DATE).

X NO DATE (15)

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

On February 3, 1995, at approximately 1824 EST, with the reactor at approximately 98'. steady state power, both pressurizer safety valves, which had been previously installed and then removed for testing, were considered inoperable. Recent test results discovered that the "as-found" set pressure for the Technical Specifications.

lift settings had shifted above the tolerance in the Immediate corrective .action was not required, since the valves were not installed.

The underlying cause of the setpoint shift has been attributed to a combination of factors, including long-term operation, removal and shipping to an off-site facility for testing, as well as a restrictive tolerance in the Technical Specifications. This event is NUREG-1022 Cause Code (B) .

Corrective action to preclude repetition is outlined in Section V.B.

9503160241 950306 PDR ADOCK 05000244 8 PDR NRC FORH 366 (5-92)

NRC FORM 366A .S. NUCLEAR REGULATORY COMHISSION PPROVED BY OHB NO. 3150.0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.

FORWARD COHMENTS REGARDING BURDEN ESTIHATE TO LICENSEE EVENT REPORT (LER) THE INFORHATION AHD RECORDS MANAGEMENT BRANCH TEXT CONTINUATION (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF HANAGEHEHT AND BUDGET WASHINGTON DC 20503.

FACILITY NAHE (1) DOCKET NUHBER (2) LER NUHBER (6) PAGE (3)

SEQUENTIAL REVISION YEAR R.E. Ginna Nuclear Power Plant 05000244 UM 2 OF 9 95 001 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

I. PRE-EVENT PLANT CONDITIONS:

The plant was at approximately 98'. steady state reactor power with no major activities in progress. Two new pressurizer (PRZR) code safety valves had been purchased, and were tested in the spring of 1993 at the valve manufacturer's test facilities. The valves were shipped to psig (+/- 1 '.

Rochester Gas and Electric (RGEE) with an "as-left" set pressure of 2485 The original safety valves at Ginna Station were removed during the 1993 outage for annual testing, and these two new safety valves were installed.

These valves (V-434 and V-435) were then considered operable during the 1993/1994 operating cycle (cycle 23) These two valves were then removed for annual lift ~

testing during the 1994 outage, and the original pair of safety valves (which had been tested in 1994) were installed for the 1994/1995 operating cycle (cycle 24). The removed valves were shipped to a test facility in Huntsville, Alabama, for testing, as per RGEE purchase order NQ-14349-C-JW.

The valves were tested to the requirements of RGGE Test Specification MET-049, "Pressurizer Safety Relief Valve Setpoint Testing", with steam as the test medium, on January 10, 1995 (for V-434) and January 11, 1995 (for V-435). RGEE Quality Assurance (QA) witnessed the tests. The test results showed that the "as-found" setpoints were 2525 psig (for V-434) and 2543 psig (for V-435), which exceeded the 1.

Technical Specifications. These results were recognized as nonconforming, lift setting tolerance of and a Nonconformance Report (NCR 95-005) was initiated to document this condition.

On February 3, 1995, during review of NCR 95-005 by System Engineering and Nuclear Engineering Services (NES), it was determined that this represented a potentially reportable condition.

NRC FORM 366A (5-92)

NRC FORM 366A .S. NUCLEAR REGULATORY COMMISSION PPROVED BY OHB NO. 3150 ~ 0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPOHSE TO .COMPLY WITH THIS IHFORHATION COLLECTIOH REQUEST: 50.0 HRS.

FORWARD COHHENTS REGARDIHG BURDEN ESTIHATE TO LICENSEE EVENT REPORT'(LER) THE INFORMATION AND RECORDS MANAGEMENT BRANCH TEXT CONTINUATION (MHBB 7714), UPS. NUCLEAR REGULATORY COHHISSION,

'WASHINGTON, DC 20555-0001 AHD- TO THE PAPERWORK REDUCTION PROJECT (31i0-0104), OFFICE OF HAHAGEMENT AHD BUDGET WASHINGTON DC 20503.

FACILITY NAME (1) DOCKET NUHBER (2) LER NUHBER (6) PAGE (3)

YEAR SEQUENTIAL REVISION R.E. Ginna Nuclear Power Plant 05000244 3 OF 9 95 001 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

DESCRIPTION OF EVENT:

A. DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES:

1. March, 1993: Newly procured PRZR safety valves are satisfactorily tested at manufacturer's test facility.

2. April, 1993: Newly procured PRZR safety valves are installed for the 1993/1994 operating cycle (cycle 23).

3. January 11, 1995: Testing of safety valves completed at off-site testing facility. Test results show that the exceeded the lift setting tolerance. Event date and time.

lift pressure

4. February 3, 1995, 1824 EST: Test results are reviewed with the System Engineer. Discovery date and time.

5. February 3, 1995, 2011 EST: Shift Supervisor notifies NRC per 10 CFR 50.72.

B. EVENT:

On February 3, 1995, at approximately 1824 EST, the reactor was at approximately 98'. steady state reactor power, and no major activities were in progress. NES personnel, from Mechanical Engineering and Nuclear Safety and Licensing (NSEL), were reviewing the status of NCR 95-005 with the System Engineer. Review of the NCR suggested 'an operability question involving these previously, installed safety valves. Since both valves were previously installed during cycle '23, it valves had shifted out of tolerance during cycle 23.

was conservatively assumed that the HRC FORH 366A (5-92)

NRC FORH 366A .S. NUCLEAR REGULATORY COHHISSION PROVED BY OHB NO. 3150 ~ 0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLEC'TION REOUEST: 50.0 HRS.

FORWARD COMHENTS REGARDING BURDEN ESTIHATE TO LICENSEE EVENT REPORT (LER) THE INFORMATION AND RECORDS MANAGEMENT BRANCH TEXT CONTINUATION (MNBB 7714), U.S. NUCLEAR REGULATORY COHHISSION, WASHINGTON, DC 20555-0001, AHD TO THE PAPERWORK REDUCTION PROJECT (3150.0104), OFFICE OF MANAGEHENT AND BUDGET WASHINGTON DC 20503.

FACILITY NAHE (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)

SEOUENTIAL REVISION R.E. Ginna Nuclear Power Plant 05000244 4 OF 9 95 -- 001-- 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

The "as-found" setpoints were 1.6 % (for V-434) and 2.3% (for V-435) above 2485 psig. This is contrary to Ginna Technical Specification 3.1.1.3.c, which states, "Whenever the reactor is at or above an RCS temperature of 350 degrees F, both pressurizer code safety valves shall be operable with a lift setting of 2485 psig +/- 1%." A conservative decision was made to report this event under the criteria of 10 CFR 50.72 (b) (2) (iii) (D), based on input from NS&L that a +/- 1% tolerance for safety valve actuation is an assumption for several design basis events. The NRC was notified at approximately 2011 EST on February 3.

Subsequent evaluations have not been able to conclusively determine the time that the setpoint shift occurred, nor even occurred during cycle 23. Review of design basis events has if the shift confirmed that this condition does not meet the reporting criteria of 10 CFR 50.72.

INOPERABLE STRUCTURES, COMPONENTS, OR SYSTEMS THAT CONTRIBUTED TO THE EVENT:

None D. OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED:

None METHOD OF DISCOVERY:

RG&E QA Surveillance of the test data identified that the test results were unacceptable. This information 'was forwarded to NES, and NCR 95-005 was initiated. During a review of this NCR between NES and System Engineering, this condition was evaluated as potentially reportable.

HRC FORH 366A (5-92)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION PPROVED BY OHB NO. 3150.0104 (5-92> EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REOUEST: 50 ' NRS.

FORWARD COHHENTS REGARDIHG BURDEN ESTIHATE TO LICENSEE EVENT REPORT (LER) THE INFORMATION AND RECORDS MANAGEMENT BRANCH TEXT CONTINUATION (HNBB 7714), U.S. NUCLEAR REGULATORY COMHISSION, WASHINGTON, DC 20555-0001 AND TO THE PAPERWORK REDUCTION PROJECT (3140 0104>, OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) LER NUHBER (6) PAGE (3)

YEAR SEOUENTIAL REVISION R.E. Ginna Nuclear Power Plant 05000244 5 OF 9 95 001 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

F. OPERATOR ACTION:

The System Engineer notified the Shift Supervisor of the test results that affected both PRZR safety valves that were previously installed and considered operable during cycle 23, and that these results did not affect currently installed equipment. A decision was made to notify the NRC per 10 CFR 50.72 (b) (2) (iii) (D). This notification was made at approximately 2011 EST on February 3, 1995.

Since this event did not affect installed plant equipment, no other operator, actions were necessary.

SAFETY SYSTEM RESPONSES:

None III. CAUSE OF EVENT:

IMMEDIATE CAUSE:

The immediate cause for both PRZR safety valves being considered inoperable was that the were above the setpoint "as-found" lift settings for these valves tolerance of Technical Specification 3.1.1.3.c.

B. INTERMED1ATE CAUSE:

The

'etpoint intermediate cause for the "as-found" tolerance of Technical Specifications was a shift in the lift settings above the setpoints from the "as-left" conditions of March, 1993, to the "as-found" conditions of January, 1995.

NRC FORH 366A (5-92>

NRC FORM 366A. .S. NUCLEAR REGULATORY COHHISSION PROVED BY OHB NO. 3150.0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COHPLY WITH THIS INFORHATION COLLECTION REQUEST: 50.0 NRS.

FORWARD COHHENTS REGARDING BURDEH ESTIHATE TO LICENSEE EVENT REPORT (LER) THE INFORHATION AND RECORDS HANAGEHENT BRANCH TEXT CONTINUATION (HNBB 7714), U.S. NUCLEAR REGULATORY COHHISSIONg WASHINGTON, DC 20555-0001, AHD TO THE PAPERWORK REDUCTIOH PROJECT (3150-0104), OFFICE OF HANAGEHENT AND BUDGET WASHINGTON DC 20503.

FACILITY NAHE (1) DOCKET NUHBER (2) LER NUHBER (6) PAGE (3)

YEAR SEQUENTIAL REVISION R.E. Ginna Nuclear Power Plant 05000244 6 OF 9 95 001-- 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

C. ROOT CAUSE:

The underlying cause of the shift in the setpoints is attributed to a combination of factors, including attendant variables affecting the long-term operation of the valves, and the subsequent removal, decontamination, handling, and shipping of the valves to an off-site facility for testing..

The Ginna Technical Specification requirements of +/- 1-'. may not be appropriate with respect to the allowances for normal setpoint shifts during operation, removal, and shipping.

This event is NUREG-1022 Cause Code (B), "Design Manufacturing, Construction '/ Installation."

IV. ANALYSIS OF EVENT:

This event is reportable in accordance with 10 CFR 50.73, Licensee Event Report System, item (a) (2) (vii) (D), which requires a report of, "any event where a single cause or condition caused at least one independent train or channel to become inoperable in multiple systems or two independent trains or channels to become inoperable in a single system designed to ... mitigate the consequences of an accident." Both independent trains of pressure relief for the PRZR were considered inoperable due to the "as-found" Technical Specifications.

lift settings above the tolerance of the NRC FORH 366A (5-92)

NRC FORM 366A .S. NUCLEAR REGULATORY COMMISSION PROVED BY OHB NO. 3150-0104 (5.92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COHPLY WITH THIS INFORHATIOH COLLECTION REOUEST: 50.0 HRS.

FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO LICENSEE EVENT REPORT (LER) THE INFORHATION AHD RECORDS MANAGEMENT BRANCH TEXT CONTINUATION (HNBB 7714), U.S. NUCLEAR REGULATORY COMHISSION, WASHINGTON, DC 20555-0001, AHD TO THE PAPERWORK REDUCTION PROJECT (3150 0104), OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.

FACILITY NAME (1) DOCKET NUHBER (2) LER NUMBER (6) PAGE (3)

SEOUENT I AL REVISION R.E. Ginna Nuclear Power Plant 05000244 7 OF 9 95 -- 001-- 00 TEXT (If more space is required, use additional copies of HRC Form 366A) (17)

The two limiting design basis events that challenge reactor coolant system (RCS) integrity and rely on the PRZR safety valves to mitigate their consequences are the Locked Rotor transient and the Loss of Load transients These transients were reanalyzed .by Westinghouse on behalf of RG&E.

k An assessment was performed considering both the safety consequences and implications of this event with the following results and conclusions:

~ The setpoints for the PRZR safety valves shifted at some unknown time between March, 1993, and January, 1995. This condition did not create a significant safety hazard for the following reasons:

1. While the valves would have been declared inoperable (had the condition been known) based on the Technical Specification tolerance, the reanalyses of the two limiting transients shows that if the valves had lifted at the "as-found" pressure during a design basis event, they would have performed their design function with acceptable results. Thus, the acceptance criteria of all UFSAR Chapter 15 design basis events would still be satisfied.

2. The design basis conditions bound the actual conditions that existed during cycle 23. Factors that would have made the limiting events less severe are:

(a) Steady-state reactor power was approximately 98% during cycle 23, versus the design condition of 100'%b)

The "as-found" setpoints were bounded by the Westinghouse reanalysis assumptions for the setpoint tolerance.

~ No event that would have required PRZR safety valve actuation occurred during cycle 23.

There were no operational or safety consequences or implications attributed to the shift in lift setpoints, because all the required RCS pressure limitations were met, using the "as-found" setpoints.

Based on the above, it can be concluded that the public's health and safety was assured at all times.

HRC FORH 366A (5-92)

NRC FORH 366A .S. NUCLEAR REGULATORY COMMISSION PROVEO BY OHB NO. 3150 ~ 0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORHATION COLLECTION REQUEST: 50.0 HRS.

FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO LICENSEE EVENT REPORT (LER) THE IHFORHATIOH AND RECORDS HANAGEHENT BRANCH TEXT CONTINUATION (MNBB 7714), U.S. NUCLEAR REGULATORY COHMISSION, WASHINGTON, DC 20555-0001, AHD TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.

FACILITY NAME (1) DOCKET NUHBER (2) LER NUMBER (6) PAGE (3)

YEAR SEQUENTIAL REVISION R.E. Ginna Nuclear Power Plant 05000244 M

8 OF 9 95 -- 001-- 00 TEXT ( If more space is required, use additional copies of NRC Form 366A) (17)

V ~ CORRECTIVE ACTION:

A. ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS:

The valve seats were lapped and the valves adjusted as necessary

~

to bring the lift Specifications tolerance.

settings into conformance with the Technical

~ The accident analyses that are affected by a PRZR safety valve with a larger tolerance than required by Technical Specification 3.1.1.3.c have been reanalyzed by Westinghouse. The results are that the functions of the PRZR safety valves were never unacceptable with the "as-found" lift settings.

B. ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:

Based on the results of the accident reanalysis, a revision to the Technical Specifications to provide more realistic and achievable lift setting tolerances and acceptance criteria for operability will be pursued on a priority basis, as part of the RG&EiNRC effort to implement the Improved Technical Specifications (ITS).

To minimize the chance of a shift in the setpoint due to activities associated with on-site removal, handling, decontamination, packaging, shipping, storing, and reinstallation, the administrative controls for removal, shipping, testing, and reinstallation of these valves will be evaluated and enhanced, as appropriate, to ensure that proper controls are in place for key activities that could inadvertently affect the lift settings.

To increase the repeatability of test results, NES will evaluate the adequacy of the test requirements of MET-049, and revise the test specification, as appropriate.

NRC FORH 366A (5-92)

NRC FORM 366A .S. NUCLEAR REGULATORY COHHISSION PROVED BY OHB NO. 3150 ~

0104 (5-92) EXP I RES 5/31/95 ESTIHATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.

FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO LICENSEE EVENT REPORT (LER) THE INFORHATION AND RECORDS MANAGEMENT BRANCH TEXT CONTINUATION (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001 AND TO THE PAPERWORK REDUCTION PROJECT (31/0-0104), OFFICE OF MANAGEHENT AND BUDGET WASHINGTON DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) LER NUHBER (6) PAGE (3)

SEQUENTIAL REVISION YEAR R.E. Ginna Nuclear Power Plant 05000244 9 OF 9 95 001 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

VI. ADDITIONAL INFORMATION:

A. FAILED COMPONENTS:

The failed components are Crosby Valve and Gage Co. safety valves, Model HB-BP-86E, serial numbers N69877-00-0006 and N69877-00-0007.

PREVIOUS LERs ON SIMILAR EVENTS:

A similar LER event historical search was conducted with the following results: No documentation of similar LER events with the same root cause at Ginna Nuclear Power Plant could be identified.

C. SPECIAL COMMENTS:

None NRC FORH 366A (5-92)