05000301/LER-2005-001, Regarding Main Steam Safety Valve 2MS-02008 Lift Set Point Exceeds Acceptance Criteria

Regarding Main Steam Safety Valve 2MS-02008 Lift Set Point Exceeds Acceptance Criteria

ML051820337
Person / Time
Site:	Point Beach  (DPR-027)
Issue date:	06/20/2005
From:	Koehl D Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NRC 2005-0076 LER 05-001-00
Download: ML051820337 (5)

LER-2005-001, Regarding Main Steam Safety Valve 2MS-02008 Lift Set Point Exceeds Acceptance Criteria

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(vi) 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(viii)(B)
3012005001R00 - NRC Website

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Committed to Nuclear Excelln Point Beach Nuclear Plant Operated by Nuclear Management Company, LLC June 20, 2005 NRC 2005-0076 10 CFR 50.73 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington DC 20555 Point Beach Nuclear Plant Unit 2 Docket 50-301 License No. DPR-27 Licensee Event Report 301/2005-001-00 Main Steam Safetv Valve Lift SetDoint Exceeds Accertance Criteria Enclosed is Licensee Event Report (LER) 301/2005-001-00 for the Point Beach Nuclear Plant Unit 2. This LER describes the discovery of the lift set point pressure for one of eight main steam safety valve being in excess of the licensee's acceptance criteria.

This condition is reportable in accordance with 10 CFR 50.73(a)(2)(i)(B).

This submittal contains no new regulatory commitments and no revisions to existing

commitments

Dennis L. Koehl Site Vice-President, Point Beach Nuclear Plant Nuclear Management Company, LLC Enclosure cc:

Administrator, Region IlIl, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC PSCW 6590 Nuclear Road

• Two Rivers, Wisconsin 54241 Telephone: 920.755.2321

I f..

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LICENSEE EVENT REPORT (LER)

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FACILITY NAME (1)

DOCKET NUMBER (2)

PAGE (3)

POINT BEACH NUCLEAR PLANT UNIT 2 05000301 1 of 4 TITLE (4)

MAIN STEAM SAFETY VALVE 2MS-02008 LIFT SET POINT EXCEEDS ACCEPTANCE CRITERIA EVENT DATE (5)

LER NUMBER (6)

REPORT DATE (7)

OTHER FACILITIES INVOLVED (8)

S l

FACILITY NAME DOCKET NUMBER MO DAY YEAR NUMBER NO MO DAY YEAR Not Applicable 04 20 2005 2005

-- 001 00 06 20 2005 FACILITY NAME DOCKET NUMBER OPERATING 6

THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 0: (Check all that apply) (11)

MODE (9) 20.2201(b) 20.2203(a)(3)(ii) 50.73(a)(2)(ii)(B) 50.73(a)(2)(ix)(A)

POWER 0%

20.2201 (d)

_ 20.2203(a)(4)

_ 50.73(a)(2)(iii) 50.73(a)(2)(x)

LEVEL (10)

_ 20.2203(a)(1) 50.36(c)(1)(i)(A) 50.73(a)(2)(iv)(A) 73.71 (a)(4) 20.2203(a)(2)(i) 50.36(c)(1)(ii)(A) 50.73(a)(2)(v)(A) 73.71(a)(5)

.2!,

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_202203(a)(2)(fl) 50.36(c)(2) 50.73(a)(2)(v)(B)

OTHER 20 2203(a)(2)(iii) 50.46(a)(3)(i) 50.73(a)(2)(v)(C)

Specify In Abstract below or In

• 20.2203(a)(2)(iii)

__5.6c()

07()2()B TE

._202203(a)(2iv) 50.73(a)(2)(i)(A)

_50.73(a)(2)(v)(D)

_NRC Form 366A

'20.2203(a)(2)(v)

X 50.73(a)(2)(i)(B)

_ 50.73(a)(2)(vi)_

20.2203(a)(2)(vi) 50.73(a)(2)(i)(C) 50.73(a)(2)(viii)(A) 20.2203(a)(3)(i) 50.73(a)(2)ii)(A) 50.73(a)(2)(viii)(B)

LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER (Include Area Code)

F. Hennessv 1

920-755-6461 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

MANU-REPORTABLE MANU-REPORTABLE

CAUSE

SYSTEM COMPONENT FACTURER TO EPIX

CAUSE

SYSTEM COMPONENT FA CTURER TO EPIX X

SB RV C710 Y

SUPPLEMENTAL REPORT EXPECTED (14)

EXPECTED MONTH DAY YEAR YES (If yes, complete EXPECTED SUBMISSION DATE).

X NO DATE (15)

SUB ABSTRACT This report describes the discovery during routine Technical Specification (TS) surveillance testing that one main steam safety valve (MSSV), 2MS-02008, had a lift set point that exceeded the licensee's acceptance criteria. The lift set point for 2MS-02008 also exceeded the ASME OM Code 1995 Edition, 1996 Addenda acceptance criteria. No conclusive cause for the test failure has yet been determined. This valve was last refurbished on November 3, 2000. Four as-left set-pressure tests performed following valve refurbishment all passed the Licensee's acceptance criterion of 1125 psig +/-1%. These type valves are tested off site at a vendor facility using full pressure boiler testing. PBNP TS Surveillance Requirement (SR) 3.7.1.1 requires MSSV testing in accordance with the IST program. The Inservice Testing (IST) program requires scope expansion if the set point test is failed. This scope expansion resulted in two additional MSSVs being sent off site for testing. The lift set points for all other MSSVs tested were found to be acceptable.U.S. NUCLEAR REGULATORY COMMISSION (1-2001):

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITYNAME (1) s'-.

l DOCKETNUMBER(2)

LER NUMBER (6)

PAGE 3 YERI SEUENTIAL IREVISION POINT BEACH NUCLEAR PLANT UNIT 2 05000301 l

EAR NUMBER I NUMBER l 2 of 4 2005

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TEXT (if more space is required, use additional copies of NRC Fomi 366A) (17)

Event Description

Technical Specification LCO 3.7.1 states that "Four MSSVs per steam generator shall be OPERABLE".

Surveillance Requirement 3.7.1.1 states, "Verify each required MSSV lift setpoint per Table 3.7.1-2 in accordance with the Inservice Testing Program. Following testing, lift setting shall be within +/-1 %." The modes of applicability for this LCO are 1 through 3.

On April 20, 2005, Point Beach Nuclear Plant (PBNP), Unit 2, was in a routine refueling shutdown (U2R27).

During U2R27, three main steam safety valves [SB; RV] (MSSVs) were removed for inservice testing (IST) and replaced with similar Crosby model HA65W valves of identical set pressure and capacity. The removed valves were sent to Wyle Laboratories for full boiler set pressure and seat leakage testing as part of the Inservice Testing (IST) program. The three removed valves were from PBNP equipment numbers 2MS-02006 (S/N XX05950121), 2MS-02007 (S/N XX05950122), and 2MS-02008 (S/N XX05950123).

During testing of valve 2MS-02008 (Serial NumberXX05950123) [RV], the valve exceeded the lift set point acceptance criteria. Although the cause has not been determined, it is possible that this valve was not operable during plant operation. Therefore, this occurrence is considered reportable in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by TS SR 3.7.1.1.

Three successive set pressure tests were conducted on 2MS-02008 on April 20, 2005. The valve lifted on the first test at 1170 psig, which is 4% above the nominal setpoint of 1125 psig. This exceeds the licensee's acceptance criterion of 1.93%, which had been established in licensee calculation 98-0103 based on PBNP accident analysis assumptions of a 3% allowance for setpoint tolerance, an allowance of 6.6 psi for ambient temperature effects on valve setpoint, and test equipment uncertainty of 0.25%. This is also exceeds the ASME OM Code 1995 Edition, 1996 Addenda acceptance criterion.

Two subsequent set pressure tests on 2MS-02008, following in 10-minute intervals, were performed after the first failed test and were within the licensee's acceptance criterion. The valve lifted a second time at a set point of 1123 psig. The valve lifted a third time at a set point of 1118 psig. Prior to the lift tests, this valve had passed initial seat leakage testing at 1012 psig, by showing no steam leakage. Plant criteria establish that the valve shall be leak tight at 1012 psig (90% of set point). After lifting, the valve failed seat leakage testing, at 1012 psig, by showing visual steam leakage.

2MS-02008 is one of four main steam safety valves for the PBNP Unit 2 UB" steam generator [SG]. This valve had been in service since installation on November 20, 2000. Two other MSSVs (2MS-02006 and 2MS-02007) that were in service for the same length of time were tested as part of the original sampling group. Both of these valves passed set point testing within 1 % of set point. This group of three valves was tested because they were the longest in service.

Failure of 2MS-02008 is an ASME Code IST program failure. The IST program requires scope expansion if the set point test is failed. The two additional valves tested for the required scope expansion successfully passed set point testing.

Component and System Description:

The Main Steam Safety Valves protect the steam generators from overpressurization. This isU.S. NUCLEAR REGULATORY COMMISSION (1-2001)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1 DOCKET NUMBER 2 l LER NUMBER (6)

PAGE 3 POINT BEACH NUCLEAR PLANT UNIT 2 05000301 YEAR NUMBERN NMEBVSER E

3 of 4 2005 001 00 TEXT (Itmore space is required, use additionaf copies of NRC Forn 366A) (17) accomplished with four self-actuated valves per steam generator that are located on a 30-inch branch line from the associated main steam header upstream of the Main Steam Isolation Valve [ISV]. The safety valves are designed in accordance with ASME Section 1II. The four valves per steam generator are set to relieve at 1085,1100,1125, and 1125 psig, respectively. The valves are set to lift within 1% of their nominal set point. The staggered settings minimize the effects of simultaneously opening and closing the valve on the protected system. The eight MSSVs per unit have a combined rated capability of 6,664,000 lbs/hr.

The MSSVs are six inch, high capacity nozzle type relief valves manufactured by Crosby Valve and Gage Co., Model HA65W. Spring compression maintains the valves closed against normal system pressure.

The spring compression is adjusted to alter the relief set point. The valves will open when the set point pressure is exceeded with very little simmer or waming. The MSSVs will continue to relieve until the pressure under the valve decreases; at which time the valve will close sharply. The MSSVs and safety relief header are located outside the containment structure in the Unit 2 facade.

In addition to the safety relief valves, each main steam line is provided with one six-inch power operated relief valve (atmospheric steam dump) [RV]. These valves are automatically controlled by pressure or may be manually operated from the Control Room for releasing sensible and core decay heat to the atmosphere. The two atmospheric steam dumps are capable of passing no less than 10% of the maximum calculated steam flow at no-load steam pressure.

Safety Significance

PBNP FSAR Section 14.1.9 presents the accident analysis for "Loss of Extemal Electrical Load". That analysis provides assumptions for MSSV characteristics in Table 14.1.9-2. The lift pressure assumed for the MSSVs with a nominal set pressure of 1125 psig is 1177.7 psig. The lift pressure is listed as the nominal set pressure, plus 3% allowance for set point tolerance, plus 50-psi allowance for frictional pressure drop between the steam generator shell and the MSSVs location. Under these conditions, the conclusion of the analysis is that the capacity of the pressure relieving devices, the MSSVs, are adequate to limit the maximum steam generator shell pressure to within the code requirements of 110% of the design pressure (110% of 1085 psig is 1194 psig). In this event, the as-found initial lift pressure for 2MS-2008 was 1170 psig. Adding an allowance of 6.6 psi for temperature effects, and +0.25% of set point for measuring tolerance (2.9 psi), and applying the 50-psi allowance for the pressure drop between the steam generator shell and the valve; the peak steam generator pressure under the postulated conditions for a loss of load from 100% power could have been as high as 1229.5 psig. This pressure exceeds the test acceptance criteria for the maximum steam generator pressure of 1194 psig. However, considering that the steam generators were subjected to a pre-service secondary side hydrostatic test pressure of 1357 psig, this difference is not significant and the conclusion of the analysis that the accident presents no hazard to the integrity of the main steam system remains applicable.

As a practical matter, it is likely that the maximum steam system pressure would be lower than that assumed in the analysis because no credit is taken in the accident analysis for operation of the steam dump system or steam generator power operated relief valves. Therefore, even in the event of a postulated loss of external electrical load accident, the safety significance of operating with this MSSV (2MS-02008) at the as found condition was minimal.U.S. NUCLEAR REGULATORY COMMISSION (1-2001)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1 DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

IYA SEQUENTLA REVISION POINT BEACH NUCLEAR PLANT UNIT 2 05000301.

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2005 001 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

Both the Loss of Normal Feedwater (FSAR 14.1.10) and Loss of All AC (FSAR 14.1.11) analyses credit MSSV relief and assume the same relief valve parameters. However, the analysis results for main steam peak pressure are bounded by the loss of external electrical load results.

NMC further concludes that the MSSV test failure did not constitute a loss of any safety function; therefore, the failure of 2MS-02008 to lift at its setpoint did not constitute a safety system functional failure.

Cause

No conclusive cause for the test failure has yet been determined. 2MS-02008 is awaiting refurbishment by Crane Nuclear personnel.

Previous maintenance performed on 2MS-02008 in 2000 involved removal of the valve from the system and shipping it to Wyle Laboratories for testing. The valve was reinstalled in the system after refurbishment.

Three as-found set-pressure tests performed prior to refurbishment by Crane Nuclear personnel on November 1, 2000 were successfully completed. After valve refurbishment, Wyle Laboratories performed the as-left tests on November 3, 2000. On that day, four set-pressure tests were performed at 10-minute intervals, all of which passed the Licensee's acceptance criterion of 1125 psig +/-1%.

Based on the previous maintenance history, there was no apparent indicator that would have predicted the April 20, 2005 valve failure.

Failure analysis following valve disassembly by Crane Nuclear personnel remains ongoing.

Corrective Action

Two additional MSSVs were tested per ASME Section Xl requirements. The valves with the longest intervals since testing were 2MS-2012 (S/N XX05950118) and 2MS-2013 (S/N XX0595119). These valves were also sent to Wyle Laboratories for testing. Both valves passed Licensee acceptance criteria. The currently installed MSSVs have all passed required testing and are operable.

Previous Similar Events

A review of recent LERs (past three years) identified one event that involved a failure of a safety valve to lift at required pressure in accordance with the inservice testing program as specified by the Technical Specifications:

LER Number Title 301-2002-002-00/01 Pressurizer Safety Valve Failed to Lift at Test Pressure