05000261/LER-2002-001, From H.B. Robinson, Unit 2 Regarding Four Main Steam Safety Valves Failure to Meet Acceptance Criteria During Lift Pressure Testing

From H.B. Robinson, Unit 2 Regarding Four Main Steam Safety Valves Failure to Meet Acceptance Criteria During Lift Pressure Testing

ML023460172
Person / Time
Site:	Robinson
Issue date:	12/09/2002
From:	Cleary T Carolina Power & Light Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RNP-RA/02-0181 LER 02-001-00
Download: ML023460172 (9)

LER-2002-001, From H.B. Robinson, Unit 2 Regarding Four Main Steam Safety Valves Failure to Meet Acceptance Criteria During Lift Pressure Testing

Event date:
Report date:
Reporting criterion:	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)(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)(B), Prohibited by Technical Specifications
2612002001R00 - NRC Website

text

10 CFR 50.73 a' CP&L U O A Pmgress Energy Company Serial: RNP-RA/02-0181 DEC o 9 2002 United States Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261/LICENSE NO. DPR-23 LICENSEE EVENT REPORT NO. 2002-001-00 FOUR MAIN STEAM SAFETY VALVES FAIL TO MEET ACCEPTANCE CRITERIA DURING LIFT PRESSURE TESTING Ladies and Gentlemen:

The attached Licensee Event Report is submitted in accordance with the requirements of 10 CFR 50.73. Should you have any questions regarding this matter, please contact Mr. C. T. Baucom.

Sincerely, Timothy P. Cleary Plant General Manager CAC/cac Attachment c:

Mr. L. A. Reyes, NRC, Region II Mr. R. Subbaratnam, NRC, NRR NRC Resident Inspector, HBRSEP Robinson Nuclear Plant 3581 West Entrance Road Hartsville, SC 29550

$IEV

NRC FORM 366 U.S. NUCLEAR REGULATORY APPROVED BY OMB NO. 3150-0104 EXPIRES 6-30-2001 (1.2001)

COMMISSION

, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

FACILITY NAME (1)

DOCKET NUMBER (2)

PAGE (3)

H. B. Robinson Steam Electric Plant, Unit No. 2 05000261 l

1 OF 8

TITLE (4)

Four Main Steam Safety Valves Fail to Meet Acceptance Criteria During Lift Pressure Testing EVENT DATE (5)

LER NUMBER (6)

REPORT DATE (7)

OTHER FACILITIES INVOLVED (8)

FACILITY NAME DOCKET NUMBER MO DAY YEAR YEAR SEQUENTiAL REV M

D NOM EMO DAY YEAR ________j 05000 FACILITY NAME DOCKET NUMBER 10 09 2002 2002 - 001 - 00 12 09 2002 1

05000 OPERATING THISREPORTIS SUBMITTEDPURSUANTTO THER EUIREMENTSOF1O C:

(Checkallthatapply) (11)

MODE (9) 1

_ 20 2201(b)

_ 20.2203(a)(3)(ii) 50.73(a)(2)(ii)(B) 50.73(a)(2)(ix)(A)

POWER L

20 2201(d)

_ 20 2203(a)(4)

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

LEVEL (10) 93%

20.2203(a)l)

(

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)(fi)(A)

_ 50 73(a)(2)(v)(A)

_ 73 71 (a)(5) 20.2203(a)(2)(i) 50 36(c)(2) 50.73(a)(2)(v)(B)

OTHER Specify In Abstract below or in 20.2203(a)(2)(iii)

_ 50 46(a)(3)(ii) 50.73(a)(2)(v)(C)

NCFr 366A 20 2203(a)(2)(iv)_

50 73(a)(2)(i)(A)

_ 50 73(a)(2)(v)(D) 20 2203(a)(2)(v)

X 50 73(a)(2)(1)(B) 50 73(a)(2)(vii) 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)(i)(A) 50 73(a)(2)(v)(B)

__in_____________-,_.

LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER (Include Area Code)

C. T. Baucom I

843-857-1253 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE

l SYSTEM COMPONENT FACTURER TOREPIXO l

CAUSE

SYSTEM COMPONENT FANCTU-ER T REPORTABLE X

SB RV Crosby Y

SUPPLEMENTAL REPORT EXPECTED (14)

EXPECTED MONTH DAY_

YEAR SUBMISSION lYES (Iyes,completeEXPECTEDSUBMISSIONDATE).

X NO DATE (15) l l

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

During testing of the'Main Steam Safety Valves (MSSVs) on October 9 and 10, 2002, it was determined that four of the 12 MSSVs had as-found lift pressures that exceeded the Technical Specifications (TS) tolerance of +/- 3%. The four valves' as-found lift pressures were over TS lift pressures by the following amounts:

SV1-1A was 3.1%

over, SV1-2B was 4.1% over, SV-1C was 3.8% over, and SV1-2C was 6.9% over. A condition report (number 73940) was initiated and the corrective action program significant adverse condition investigation has been completed. The root cause of the high lift pressures was attributed to mechanical component failure/degradation based on slight binding of the spindle on the guide bearing. The corrective actions for this event included maintenance on the MSSVs and post-maintenance testing as required for the maintenance performed. This condition was determined to be reportable in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by the plant's TS. based on guidance contained in NUREG-1022. "Event Reporting Guidelines 10 CFR 50.72 and 50.73." Revision 2.

NRC FORM 366 (1-2001)

(If more space is required, use additional copies of (If more space is required, use additional copies of NRC Forn 366A) (17)

The fourth safety valve test failure occurred at 1123 hours0.013 days <br />0.312 hours <br />0.00186 weeks <br />4.273015e-4 months <br /> on October 10, 2002, during testing of valve SV1-1A. which is one of the four main steam system safety valves on the "A" steam generator main steam line. The applicable Technical Specifications LCO required action for inoperability of SV1-1A had been entered at 1122 hours0.013 days <br />0.312 hours <br />0.00186 weeks <br />4.26921e-4 months <br /> to accommodate the valve test. The valve was retested satisfactorily.

The valve was returned to operable status and the LCO action statement was exited at 1145 hours0.0133 days <br />0.318 hours <br />0.00189 weeks <br />4.356725e-4 months <br />.

A condition report (number 73940) was initiated and the corrective action program significant adverse condition investigation has been completed. This reportable event and the associated significant adverse condition investigation was reviewed by the Plant Nuclear Safety Committee on December 4. 2002.

This condition was determined to be reportable based on guidance contained in NUREG-1022, "Event Reporting Guidelines 10 CFR 50.72 and 50.73," Revision 2.

Example 3 in Section 3.2.2 of NUREG-1022, Revision 2, states that multiple test failures are reportable under 10 CFR 50.73(a)(2)(i)(B), "Any operation or condition prohibited by the plant's Technical Specifications," if the existence of similar discrepancies in multiple valves is an indication that the discrepancies may well have arisen over a period of time.

HBRSEP, Unit No. 2, Technical Specifications LCO Section 3.7.1 provides the operability requirements for the Main Steam Safety Valves (MSSVs).

LCO 3.7.1 requires the MSSVs to be operable in MODES 1. 2, and 3. Inoperability of the MSSVs, as determined during the testing, required entry into Condition A of LCO 3.7.1.

The required action for LCO 3.7.1. Condition A, required that thermal power be reduced to < 51% rated thermal power within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

No power reductions were required during the testing, because the valves were restored to operable status prior to the required completion time. Therefore, the reportability assessment of this condition is conservatively based on the guidance of NUREG-1022, Revision 2. which indicates that multiple test failures of this type represent a condition that, if known, would have required the reactor power level to be reduced and, hence, was a condition prohibited by the Technical Specifications.

An evaluation of the safety consequences of the condition was completed. The safety consequences were determined to be minimal, and the overpressure protection of the steam generators, main steam lines, and reactor coolant system [AB] would not be (If more space Is required, use additional copies of (If more space is required, use additional copies of (if more space is required, use additional copies of (If more space is required, use additional copies of NRC Fotn 366A) (17) accordance with the guidance contained in the Progress Energy Nuclear Generation Group procedure, ADM-NGGC-0101, "Maintenance Rule Program," which requires corrective actions, goal setting, and performance monitoring.

V. ADDITIONAL INFORMATION

A. Failed Component Information:

The MSSVs [System Code: SB and Component Code: RV] are style HC-65W, Type C, manufactured by the Crosby Valve and Gage Company.

The SV1-1 and 2 valves are type 6Q8, with a mean seat area of 15.473 square inches. The SV1-3 and 4 valves are type 6R10, with a mean seat area of 22.438 square inches.

B. Previous Similar Events

On April 5, 2001, SV1-1C failed the as-found lift pressure test acceptance criteria of +/- 3% The valve lift pressure was measured as 1211 psig.

The maximum acceptable as-found lift pressure for this valve is 1108 psig.

The valve spindle was replaced and the guide bearing clearance was increased to reduce the possibility of valve binding. The valve performance, although found to be out of tolerance during testing on October 9, 2002. was substantially improved as demonstrated by the as-found lift pressure test data (i.e., in 2001 the as-found lift pressure was 1211 psig and in 2002 the as-found lift pressure was 1126 psig).

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

Figure 1 Main Steam Safety Valve Diagram

=Spindle Nut and Cotter Pin Thrust Bearing

_ Spindle Rod Spring

- Guide Bearing

- Guide Valve Outlet Valve Inlet