05000263/LER-1917-003, Regarding Main Steam Isolation Valve Leakage Exceeds Technical Specification Limits
| ML17166A137 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 06/14/2017 |
| From: | Gardner P Northern States Power Co, Xcel Energy |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-MT-17-046 LER 17-003-00 | |
| Download: ML17166A137 (4) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(x) 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)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) |
| 2631917003R00 - NRC Website | |
text
2807 West County Road 75 Monticello, MN 55362 800.895.4999 xcelenergy.com June 14, 2017 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Monticello Nuclear Generating Plant Docket No. 50-263 (l Xcel Energy*
RES P 0 N S I B L E B V NAT U R E L-MT-17-046 10 CFR 50.90 Renewed Facility Operating License No. DPR-22 LER 2017-003-00 "Main Steam Isolation Valve Leakage Exceeds Technical Specification Requirements" Enclosed is the Monticello Nuclear Generating Plant (MNGP) Licensee Event Report (LER) 2017-003-00, "Main Steam Isolation Valve Leakage Exceeds Technical Specification Limits."
This condition is reportable to the NRC in accordance with 10 CFR 50. 73(a)(2)(i)(B), as an operation or condition which was prohibited by the plant's Technical Specifications.
Summary of Commitments This letter makes no new commitments and no revisions to existing commitments.
~
Peter A Gardner Site Vice President, Monticello Nuclear Generating Plant Northern States Power Company-Minnesota Enclosure cc:
Administrator, Region Ill, USNRC Project Manager, Monticello Nuclear Generating Plant, USNRC Resident Inspector, Monticello Nuclear Generating Plant, USNRC
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150*0104 EXPIRES: 10/31/2018 (06-2016)
............. )\\_
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
rPAGE Monticello Nuclear Generating Plant 05000-263 1 OF 3
't. TITLE Main Steam Isolation Valve Leakage Exceeds Technical Specification Limits
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED YEAR I SEQUENTIAL I REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR NUMBER NO.
MONTH DAY YEAR 05000 FACILITY NAME DOCKET NUMBER 04 20 2017 2017
- - 003
- - 00 06 14 2017 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §:(Check all that apply)
D 20.22o1(b)
D 20.2203(a)(3)(i)
D 50.73(a)(2)(ii)(A)
D 50.73(a)(2)(viii)(A)
D 20.2201 (d)
D 20.2203(a)(3)(ii)
D 50.73(a)(2)(ii)(B)
D 50.73(a)(2)(viii)(B) 5 D 20.2203(a)(1)
D 20.2203(a)(4)
D 50.73(a)(2)(iii)
D 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL D 20.2203(a)(2)(i)
D 50.36(c)(1)(i)(A)
D 50.73(a)(2)(iv)(A)
D 50.73(a)(2)(x)
D 20.2203(a)(2)(ii)
D 50.36(c)(1 )(ii)(A)
D 50.73(a)(2)(v)(A)
D 73.71(a)(4)
D 20.2203(a)(2)(iii)
D 50.36(c)(2)
D 50.73(a)(2)(v)(B)
D 73.71(a)(5) 000 D 20.2203(a)(2)(iv)
D 50.46(a)(3)(ii)
D 50.73(a)(2)(v)(C)
D 73.77(a)(1)
D 20.2203(a)(2)(v)
D 50.73(a)(2)(i)(A)
D 50.73(a)(2)(v)(D)
D 73.77(a)(2)(i)
D 20.2203(a)(2)(vi) 1.81 50.73(a)(2)(i)(B)
D 50.73(a)(2)(vii)
D 73.77(a)(2)(ii)
D 50.73(a)(2)(i)(C) 0 OTHER Specify in Abslracl below or in NRC Form 366A
- 12. LICENSEE CONTACT FOR THIS LER LICENSEE CONTACT rELEPHONE NUMER (lndude Area Code)
Steve Sollom, Licensing Engineer 763-295-1611
- 13. COMPLETE ONE UNE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE FACTURER TOEPIX FACTURER TOEPIX B
SB ISV A391 y
- 14. SUPPLEMENTAL REPORT EXPECTED 15.EXPECTED MONTH DAY YEAR 0 YES (If yes, complete 15. EXPECTED SUBMISSION DATE)
~NO SUBMISSION DATE
~BSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On April20, 2017 during outage 1R28 Local Leak Rate Testing (Appendix J), A0-2-86C, "13 Outboard Main Steam Isolation Valve," had an unacceptable as-found leak rate. The measured leakage rate
!was 187.8 standard cubic feet per hour (seth) which exceeds the Monticello Nuclear Generating (MNGP) Technical Specification (TS) Surveillance Requirement (SR) 3.6.1.3.12 limit of 100 seth.
IA0-2-86C was declared inoperable and the valve was subsequently disassembled to make repairs.
!The valve's stem, discs, upper/lower wedges, disc retainer, and wedge pin were replaced and retested.
The as-left leak rate after completion of the work was 2.64 seth.
This component failure is reportable in accordance with 10 CFR 50. 73(a)(2)(i)(B) as a condition prohibited by TS 3.6.1.3, "Primary Containment Isolation Valves," since A0-2-86C likely had been inoperable for greater than the TS 3.6.1.3, Required Action A.1, Completion Time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to isolate a main steam line, and the Completion Time forTS 3.6.1.3, Required Action F, to be in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Mode 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> when the completion time of A.1 is not met. There were minimal safety consequences associated with the condition since the primary containment isolation function was maintained by the inboard valve.
NRC FORM 366 (06-2016)
EVENT DESCRIPTION
SEQUENTIAL NUMBER 003 REV NO.
- - 00 On April 20, 2017, with the plant at 0% power in Mode 5 (Refueling), during refueling outage 1 R28, Local Leak Rate Testing (Appendix J) of A0-2-86C, "13 Outboard Main Steam [SB] Isolation Valve [ISV]," had an unacceptable as-found leak rate. The measured leak rate was 187.8 standard cubic feet per hour (scfh) which exceeds the Monticello Nuclear Generating (MNGP) Technical Specification (TS) Surveillance Requirement (SR) 3.6.1.3.12 limit of 100 scfh. A0-2-86C was declared inoperable and the valve was subsequently disassembled to make repairs. The valve's stem, discs, upper/lower wedges, disc retainer, and wedge pin were replaced and retested. The as-left leak rate after completion of the work was 2.64 scfh.
This component failure is reportable in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by TS 3.6.1.3, "Primary Containment Isolation Valves," since A0-2-86C likely had been inoperable for greater than the TS 3.6.1.3, Required Action A.1, Completion Time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to isolate a main steam line, and the Completion Time forTS 3.6.1.3, Required Action F, to be in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Mode 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> when the completion time of A.1 is not met.
The basis for the reportable condition is the change in wear rate associated with A0-2-86C valve internals. In 2011, A0-2-86C (Anchor Darling model W9324183 18"-900 venturied double) was disassembled and showed unexpected accelerated wear and excessive damage. The stem, upper and lower wedges, disc retainers and discs were replaced. The last as-found leak rate for A0-2-86C was 64.1 scfh in the 2015 refueling outage (1 R27). After the actuator was replaced in 1 R27 the as-left leakage was 4.4 scfh. Based on these data points it is concluded that the leak rate increased during the cycle and the valve likely had exceeded the TS SR limits during the cycle preceding 1 R28.
EVENT ANALYSIS
The event was determined to be reportable in accordance with 10 CFR 50.73 (a)(2)(i)(B), "Any operation or condition which was prohibited by the plant's Technical Specifications." Specifically, this component failure is reportable in accordance with 10 CFR 50. 73(a)(2)(i)(B) as a condition prohibited by TS 3.6.1.3, "Primary Containment Isolation Valves," since A0-2-86C likely had been inoperable for greater than the TS 3.6.1.3, Required Action A.1, Completion Time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to isolate a main steam line, and the Completion Time forTS 3.6.1.3, Required Action F, to be in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Mode 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> when the completion time of A.1 is not met.
This event is not classified as a safety system functional failure as the inboard valve was fully operational.
Page 2 of 3 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2018 06-2016)
LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. LER NUMBER YEAR Monticello Nuclear Generating Plant 05000-263 2017
SAFETY SIGNIFICANCE
SEQUENTIAL NUMBER 003 REV NO.
- - 00 There were minimal safety consequences associated with the condition. The inboard MSIV on Main Steam line "C" (A0-2-BOC) was tested for both leak rate and closing time over the past cycle and each test was completed satisfactorily. Therefore, the primary containment isolation capability of the main steam lines remained operable which ensured the required isolation safety function was maintained.
CAUSE
The failure was attributed to oscillating of the disc and wear on the trunnion pin. The oscillation caused wear between the downstream disc trunnion and mating upper wedge hole. As the wear increased, the disc dropped, increasing the gap between the disc retainer and disc groove thereby allowing further rotation of the disc. Eventually, the corner at the end of the disc groove started to contact one of the ends of the retainer plate and wear into it. This resulted in interference between the downstream disc groove area and the bottom corner of the disc retainer. This interference prevented or restricted the ability of the upper part of the downstream disc to move axially towards its corresponding body seat thereby resulting in a gap or reduced seating force in portions of the seat. Based on this, the increased leakage of the valve is attributable to wear which led to reduced seating force or a gap (due to interference) in the valve disc as it contacts the valve body seat.
CORRECTIVE ACTION
The entirety of the internal disc pack was replaced. This includes the stem, discs, upper/lower wedges, disc retainer, and wedge pin. A modification was made to hard face the trunnion outer diameter, upper wedge hole inner diameter, and disc grooves with a Stellite 21 overlay to help reduce the amount of wear. Other improvements were made to the disc retainers as well.
PREVIOUS SIMILAR EVENTS
There were no previous similar licensee event reports in the past three years.
ADDITIONAL INFORMATION
The Institute of Electrical and Electronics Engineer codes for equipment are denoted by [XX].