05000333/LER-2010-004, Regarding Main Steam Isolation Valve Leak Rate Exceeds Authorized Limit
| ML103200395 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 11/16/2010 |
| From: | Bronson K Entergy Nuclear Northeast, Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| JAFP-10-0151 LER 10-004-00 | |
| Download: ML103200395 (6) | |
| 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)(A), Seriously Degraded 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(viii)(A) 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)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(i)(8) |
| 3332010004R00 - NRC Website | |
text
~Entergy JAFP-10-0151 November 16, 2010 United States Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555 Entergy Nuclear Northeast Entergy Nuclear Operations, Inc.
James A. FitzPatrick NPP P.O. Box 110 Lycoming, NY 13093 Kevin Bronson Site Vice President - JAF
SUBJECT:
Dear Sir or Madam:
LER: 2010-004-00, Main Steam Isolation Valve Leak Rate Exceeds Authorized Limit James A. FitzPatrick Nuclear Power Plant, Unit No. 1 Docket No.
50-333 License No.
DPA-59 This report is submitted in accordance with 10 CFR 50.73(a)(2)(i)(B), 10 CFR 50.73(a)(2)(ii)(A) and 10 CFR 50.73(a)(2)(v)(C).
There are no new commitments contained in this report.
Questions concerning this report may be addressed to Mr. Joseph Pechacek, Licensing Manager, at (315) 349-6766.
Sinz~
Kevin Bronson Site Vice President KB/JP/ed
Enclosure:
JAF LER: 2010-004-00, Main Steam Isolation Valve Leak Rate Exceeds Authorized Limit cc:
USNRC, Region 1 USNRC, Resident Inspector USNRC, Project Directorate INPO Document Components:
001 Transmittal Letter with Enclosure (LER 2010-004-00)
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10131/2013 (10-2010)
Estimated burden per response 10 comply with lhis mandato!}' collection request 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />.
Repo~ed lessons learned are inCOfpOfated into the licensing process and fed back to indusl!}'.
Send commenls regarding burden estimate to the Records and FOINPrivacy Service Branch (T-5 LICENSEE EVENT REPORT (LER)
F53). u.s. Nuclear Regulatory Commission. Washington. DC 20555-0001. or by intemet e-mail to infocollecls.resource@nrc.gov. and to the Desk Officer. Office 01 Information and ReguJatOfy (See reverse for required number of Affairs, NEOB*l0202, (3150*0104). Office of Management and Budgel, Washington. DC 20503. II ameans used to impose an infOfmalion collection does not cisplay acurrenlly valid OMS control digits/characters for each block) number, the NRC may not conduct or sponsor. and a person is not required to respond 10. the information colleclion.
- 3. PAGE James A. FitzPatrick Nuclear Power Plant 05000333 10F5
- 4. TITLE Main Steam Isolation Valve Leak Rate Exceeds Authorized Limit
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED YEARISEQUENTIALIREV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR NUMBER NO MONTH DAY YEAR N/A 05000 09 17 2010 2010 00 11 16 2010 FACILITY NAME DOCKET NUMBER 004 -
N/A 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) o50.73(a)(2)(i)(C)
~ 20.2201 (b) 20.2203(a)(3)(i) 50.73(a)(2)(vil)
Mode 05
~ 20.2201 (d) 20.2203(a)(3)(i1)
~ 50.73(a)(2)(ii)(A) 50.73(a)(2)(viii)(A)
.... 20.2203(a)(1 )
20.2203(a)(4)
~ 50.73(0)(2)(11)(6)
.... 50.73(a)(2)(viii)(B)
~ 20.2203{a)(2)(i) 50.36(c)(1 )(i)(A) 50.73(a)(2)(iii)
~ 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL
.... 20.2203(a)(2)(ii) 50.36(c)(1 )(ii)(A) 50.73(a)(2)(iv)(A)
~ 50.73(a)(2)(x) 20.2203(a)(2)(iii) 50.36(c)(2) 50.73(a)(2)(v)(A) 73.71 (a)(4) 20.2203(a)(2)(iv) 50.46(a)(3)(ii) 50.73(a)(2)(v)(B) 73.71 (a)(5) 0 20.2203(a)(2)(v)
=
50.73(a)(2)(i)(A)
~ 50.73(a)(2)(v)(C)
OTHER 20.2203(a)(2)(vi)
~ 50.73(a)(2)(i)(B) o 50.73(a)(2)(v)(D)
Specify in Abstract below or in
EVENT DESCRIPTION
The plant entered Refueling Outage 19 on September 13,2010. Type C Local Leak Rate Testing (LLRT) activities on the Primary Containment [NH] penetrations and isolation valves commenced shortly after plant cooldown in accordance with Technical Specification (TS) 5.5.6, Primary Containment Leakage Rate Testing Program. On September 17,2010, the initial LLRT of the Main Steam System [S8] Main Steam Isolation Valves (MSIVs) 29AOV-80C (Inboard) and 29AOV-86C (Outboard) failed to meet acceptance criteria. The initial test was performed between the MSIVs and provided indeterminate results due to testing the inboard valve from below the seat. The test was re-performed following Main Steam Line (MSL) Plug installation, with pressure being applied from above the plug. The test results showed that both 29AOV-80C and 29AOV-86C failed to meet acceptance criteria.
BACKGROUND:
The normal sequence of testing for the MSIVs is:
- 1) Test between the valves at 25 psig per the TS requirement. Check MSL vents to determine which valve is leaking.
- 2) After Main Steam Line Plugs are installed, test the valves individually from the refuel floor through the test connection on the Main Steam Line Plug at 45 psig.
- 3) After leakage rates are measured, determine As-Found (AF) minimum pathway to verify whether or not the TS limit of 46 scfh is met.
- 4) After repairs, if required, leakage rates are measured to determine the As-Left (AL) maximum pathway to verify that the TS limit of 46 scfh is met.
As-Found Leak Rate Test Results:
Minimum Pathway Analysis of "c" Main Steam Line at 45 psig Inboard Outboard Penetration Would not hold pressure Would not hold pressure Would not hold pressure
EVENT ANALYSIS
This report is being submitted in accordance with 10 CFR 50.73(a)(2)(i)(8), "Any operation or condition which was prohibited by the plant's Technical Specifications," 10 CFR 50.73(a)(2)(ii)(A), "Any event or condition that resulted in the nuclear power plant, including its principal safety barriers, being seriously degraded," and in accordance with 10 CFR 50.73(a)(2)(v), "Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to (C) Control the release of radioactive materiaL" The primary Containment System has the capability to limit leakage, during any of the postulated design basis accidents for which it is assumed to be functional, such that offsite doses do not exceed the guideline values set forth in 10 CFR 100. Compliance with 10 CFR 50, Appendix J provides assurance that the Primary Containment including those systems which penetrate the Primary Containment do not exceed the allowable leakaQe rate specified in the TS.(10*2010)
LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET U.S. NUCLEAR REGULATORY COMMISSION
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE YEAR I SEQUENTIAL I NUMBER James A. FitzPatrick Nuclear Power Plant EVENT ANALYSIS (continued):
05000333 2010 -
004 REV NO.
00 30FS Based on the LLRT failures, 29AOV-80C and 29AOV-86C were disassembled and inspected. The inspection of 29AOV-80C did not identify any indications of seat wear or damage. The presence of corrosion products on the seat was noted. Inspection of 29AOV-86C revealed a small wear mark described as erosion between the three o'clock and five o'clock position in the stellite seat and wear in the bore which would cause the valve disc to be off center and slightly off the seat. These material conditions were the result of normal wear on these components.
CAUSE OF EVENT
The LLRT failure of 29AOV-80C was caused by corrosion products fouling the valve seat and the LLRT failure of 29AOV-86C was the result of flow erosion on the valve body and seating surface.
EXTENT OF CONDITION; All MSIVs are leak rate tested, as described above, during each refueling outage. No other MSIVs were determined to exceed the Technical Specification Leak Rate Limits and all Maximum Path As-Left leak rates were within the Technical Specification limits.
As-Left Maximum Pathway Leak Rate Analysis of "C" MSL Inboard 0.215 Outboard 0.215 Penetration 0.215 FAILED COMPONENT IDENTIFICATION:
Component 10:
Noun Name:
Manufacturer:
Model:
Type:
29AOV-80C, 29AOV-86C Main Steam Line Isolation Valves Edward Valves, Inc.
1612 JMMNY 1250 psi, 24 inch, Globe Valve
CORRECTIVE ACTIONS
Completed Actions:
- 2) Cleaned valve internals and repaired identified wear on 29AOV-86C
NRC FORM 36M (10-2010)[1a-201O}
LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET U.S. NUCLEAR REGULATORY COMMISSION
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE YEAR I SEQUENTIAL I NUMBER James A. FitzPatrick Nuclear Power Plant CORRECTIVE ACTIONS (continued):
Open Actions:
05000333 2010 -
004 REV NO.
00 40F5 Develop a Failure Modes Analysis document to be used by Operations to evaluate LLAT test related failure influences and by Maintenance and Central Planning to evaluate valve anomalies that can influence LLRT test related failures.
ASSESSMENT OF SAFETY CONSEQUENCES
An upper bound on the leak rate through 29AOV-80C and 29AOV-86C could not be determined; therefore the potential dose consequences of this event could not be precisely quantified. A Level 2 Probabilistic Aisk Assessment to estimate the incremental large early release frequency (LEAF) increase was conducted to quantify I evaluate the safety significance of this event. The average maintenance model was modified by setting basic events associated with MSIVs 29AOV-80C and 29AOV-86C to TAUE. Specifically, basic events MSV-AOV-OO-80C and MSV-AOV-OO-86C (fails-to-close NO-FO) are changed. These basic events only appear in the primary containment isolation fault tree model as part of the LEAF model. Hence, these events have no impact on the core damage frequency.
In addition to changing basic events MSV-AOV-OO-80C and MSV-AOV-OO-86C failure rate; the primary containment isolation system fault tree was modified to reflect that MSIVs failure to close on demand are applicable to plant events that result in core damage at high Reactor Pressure Vessel pressure.
Using the analysis described above results the estimated frequency of a large early release was 2.85 x 10-7I yr.
This frequency is less than the LERF safety goal of 1 x 10-6/yr. This value represents a 9.2 percent increase from the 'base case' value of 2.61 x 10.7Iyr.
The impact on the LERF as described in Reg. Guide 1.174 defines very small changes in risk as resulting in increases of core damage frequency (CDF) below 10*6/yr and increases in LERF below 10'7Iyr. Since the MSIVs leakage does not impact CDF, the relevant metric is LEAF.
Calculating the increase in LEAF is simply the increase in risk from the above LEAF base case to the MSIV leakage case or 6LEAF = LEAF MSIV CASE - LERF BASE CASE
=2.85 x 10*7/yr - 2.61 x 10.7 /yr
.1LERF =2.40 x 10.8 /yr Since guidance in Reg. Guide 1.174 defines very small changes in LEAF as below 1Q,7/yr, excessive MSIVs leakage from 29AOV-80C and 29AOV-86C is non-risk significant.
SIMILAR EVENTS
There have been other failures of MSIVs due to internal wear. The last failure of both an Inboard and Outboard MSIV in the same penetration occurred in 2000 and was reported under LER 2000-015.(10*2010)
LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET U.S. NUCLEAR REGULATORY COMMISSION YEAR I SEQUENTIAL I NUMBER
- 2. DOCKET
- 1. FACILITY NAME James A. FitzPatrick Nuclear Power Plant
REFERENCES:
CR-JAF-2010-05544, MSIV Leak Rate Test Failure 05000333 2010 -
- 6. LER NUMBER 004 REV NO.
00
- 3. PAGE SOF5 Reg Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis