05000265/LER-1995-001, Informs of Commitments Being Made by Ltr Re Suppl Info Concerning LER 95-001
| ML20092G214 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 03/29/1995 |
| From: | Pearce L COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20092G128 | List: |
| References | |
| LWP-95-034, LWP-95-34, NUDOCS 9509190186 | |
| Download: ML20092G214 (1) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2) |
| 2651995001R00 - NRC Website | |
text
G>mmonwealth Edivin Company ATTACi! MENT 2 Quad Cits C,enerating Statton 22 10 2tXith Asenue North cordos 2. it <>i 2 m> w Td.4tNM+22 il LWP-95-034 s
March 29, 1995 U.S. Nuclear Regulatory Commission Document Control Desk Washington DC 20555
Reference:
Quad Cities Nuclear Power Station Docket Number 50-265. DPR-30. Unit 2 Enclosed is Licensee Event Report (LER)95-001. Revision 00. for Quad Cities Nuclear Power Station.
This report i.<, submitted in accordance with the requirements of the Code of Federal Regulations. Title 10. Part 50.73(a)(2)(ii)(B). The licensee shall report any event or condition that resulted in the condition of the nuclear power plant. including its principal safety barriers, being seriously degraded, or that resulted in the nuclear plant being in a condition that was outside the design basis of the plant.
The following commitments are being made by this letter:
A supplemental report will be issued which will document the valves and penetrations that had excessive leakage, what caused the leakage, and the corrective actions taken to bring all excessive primary containment leakages below the required LLRT limits.
If there are any questions or comments concerning this letter, please refer them to Nick Chrissotimos. Regulatory Assurance Supervisor at 309-654-2241, ext. 3100.
Respectfully.
COMMONWEALTH EDISON QUAD CITIES NUCLEAR POWER STATION b. B. dud f L. W. Pearce Station Manager LWP/TB/lls Enclosure cc:
J. Schrage C. Miller INPO Records Center NRC Region III p!f$co@kojo@$j65 smmmsm S
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1 LICENSEE EVENT REPORT (LER)
Form Rev. 2.0 Feeskry Name (L)
Doctes Number G)
Page '.3)
Quad Cities Unit Two.
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Tatie (4)
Valvc/ Penetration Technical Specincation Leakage Limits Exceeded During Local Leak Rate Testing i
Evens Dene (5)
MR Numper (6)
Report Date (7)
Other Fecilmee krvolved (5) i Monsh Day Year Year Sequeanal Revision Moom Day Year Facihry Doctes Numbens) l Number Number Names i
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IJCENSEE CON TACT FOR THIS LER (IL) l NAME TELEPHON E ;R;;.15ER 2
AREA CODE Nick Chrissotimos. Regulatory Assurance. Ext. 3100 3
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COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN HIS REPORT (13)
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ABSTRACT At approximately 1630 on March 7,1995, Quad Cities Unit-2 was shutdown for the cycle 13 refueling and maintenance outage. After L0 cal Leak Rate Testing (LLRT) the "C"
Main Steam Isolation Valves (MSIV), it was determined that the measured leakage rate of 13.6 standard cubic feet per hour (SCFH) in the outboard MSIV A0-2-203-2C exceeded the individual MSIV Technical Specification (3.7.A.2.a.3) leakage limit of 11.5 SCFH.
The inboard MSIV A0-2-203-1C had a measured leakage rate of 0.8 SCFH.
On March 8,1995, at 1329 hours0.0154 days <br />0.369 hours <br />0.0022 weeks <br />5.056845e-4 months <br /> while performing LLRT on the Unit-2 Inboard Feedwater (FW) check valve 2-220-588, the valve failed with an unquantified leakage.
The unquantified leakage exceeds the 0.6 La (293.75 SCFH) combined leakage limits specified in Technical Specification 3.7. A.2.a.2.
The outboard feedwater 2-220-62B check valve had a measured leakage rate of 0.0 SCFH.
The root cause of the excessive leakages will not be known until the components have been disassembled and inspected.
A supplemental report will be issued when all as found LLRT's are complete and the cause of excessive primary containment leakages are identified.
LER265\\95\\001.WPF
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l LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Form Rec. 2.0 i
FACILrrY NAME (t)
DOCKET NUMBER G)
LER NUMBER (6)
PAGE d)
Year Sequesnel Revmon Number.
Number 0l0l1 0l0 2 lOFl 0 l 4
?uad Cities Unit Two 0l5l0l0l0l2l6l5 9l5 TEXT Energy industry idenu6cauon System (EI15) codes are sdenu6ed in the tant as (XX]
PLANT AND SYSTEM IDENTIFICATION
General Electric - Boiling Water Reactor - 2511 MWt rated core thermal power.
EVENT IDENTIFICATION:
A.
CONDITIONS PRIOR TO EVENT
Unit:
Two Event Date:
March 7, 1995 Event Time:
1630 Reactor Mode: 2 Mode Name:
Refuel Power Level:
0 i
This report was initiated by Licensee Event Report 265\\95-001.
REFUEL (2) - In this position interlocks are established so that one control rod only may be withdrawn when flux amplifiers are set at the proper sensitivity level and the refueling crane is not over the reactor. Also, the trip from the turbine control valves, turbinn stop valves, main steam isolation valves, and condenser vacuum are bypassed.
If the refueling crane is over the reactor, all rods must be fully inserted and none can be withdrawn.
8.
DESCRIPTION OF EVENT
At approximately 1630 on March 7, 1995, Quad Cities Unit 2 was shutdown for the cycle 13 refueling and maintenance outage. After Local Leak Rate Testing (LLRT) the "C" Main Steam Isolation Valves (MSIV) [SB] [ISV), it was determined that the measured leakage rate of 13.6 standard cubic feet per hour (SCFH) in the outboard MSIV A0-2-203-2C exceeded the individual MSIV Technical. Specification (3.7.A.2.a.3) leakage limit of 11.5 SCFH.
The inboard MSIV A0-2-203-IC had a measured leakage rate of 0.8 SCFH.
On March 8,1995, at 1329 hours0.0154 days <br />0.369 hours <br />0.0022 weeks <br />5.056845e-4 months <br /> while performing LLRT on the Unit 2 inboard feedwater (FW) check valve (SJ) (FCV) 2-220-588, the valve failed with an unquantified leakage.
The outboard feedwater 2-220-62B check valve had a measured leakage of 0.0 SCFH.
The unquantified leakage exceeds the 0.6 La (293.75 SCFH) combined leakage limits J
specified in Technical Specification 3.7.A.2.a.2 (combined leakage rate of all penetrations and valves, except the MSIV's, subject to type B and C tests of 10CFR, Appendix J).
The MSIV, the FW check valve, and any other primary containment valves or penetrations that fail LLRT with excessive leakage will be reported under the supplemental report to this LER.
LER265\\95\\001.WPF
LICENSEE EVENT REPORT G.ER) TEXT COPrr!NUATION Form Rev. 2.0 FACILJTY N AME (1)
DOCKET NUMBER G)
LER NUMBER (6)
PAGF, 0)
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Year Sequeouel Revision Number Numbec 0l0 3 lOFl 0 l 4 0l0lt Quad Ciss Unit Two 0l5l0l0l0l2l6l5 9l5 I am Emergy ladustry Idanu6cauon System (Ell 5) codes are idenu6ed in the text se (XXl l
4 C.
APPARENT CAUSE OF EVENT:
This report is being submitted in accordance with the requirements of the 10CFR i
50.73(a)(2)(11)(B). Any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers being seriously degraded, or that resulted in the nuclear plant being in a condition that was outside l
the design basis of the plant.
The root cause of the excessive leakages will not be known until the components have been disassembled and inspected. A supplemental report will be issued when all as found LLRT's are complete and the cause of excessive primary containment leakages are i
identifled.
D. ' SAFETY ANALYSIS OF EVENT:
The safety consequences of this event were minimal since LLRT maximum pathway is a conservative method for quantifying containment leakage.
Containment leakage is t
specified by two separate leak pathways. The minimum pathway and the maximum pathway.
The minimum pathway adds up all the leakages of the best valves of each pathway.
The maximum path adds up all the leakages of the worst valves of each pathway, assuming the best valves of each pathway fail.
The actual leakage under accident conditions would be less than that determined by i
the LLRT maximum pathway. Where only the maximum pathway limit is exceeded as in this case, more than one valve is present in a line, and it is realistic to expect the leakage to be equal to the lesser leakage of the two valves.
E.
CORRECTIVE ACTIONS
A supalemental report will be issued which will document the valves and penetrations that 1ad excessive leakage, what caused the leakage, and the corrective actions taken to bring all excessive primary containment leakages below the required LLRT limits (NTS# 2651809500101).
LER265\\95\\001.WPF
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UCENSEE EVENT REPORT (LER) TEXT CONTINUATION Fonn Rev. 2.0 FACILTTY NAML (t)
DOCKET NUMBER (2)
LER NUMBER (6)
PAGE (3)
Year Sequeanal Revuion Number.
Nuenber ol0 4 lOFl 0 l 4 Quad Cities Unit Two 0l5l0l0l0l2l6l5 9l5 0l0l1 TEXT Energy industry idenu6ceuon System (EII5) codes are idenu6ed in the text as [XX)
F.
PREVIOUS EVENTS:
A three year Nuclear Tracking System (NTS) historical search, for LER's associated with valve LLRT testing exceeding Technical Specifications, was conducted. The results of the search are included below.
LER #
Description
265/92-002 Unit-2 Valves 2-1601-21(22)(55)(56) Failed LLRT Leakage Limit.
254/92-020 Unit-1 Tech Spec Containment LLRT Exceeded Due To Various Component Failures.
254/93-007 Unit-1 B-Loop Main Steam Isolation Valve Local Leak Rate Exceeded 11.5 SCFH.
265/93-025 Unit-2 A-Loop Main Steam Isolation Valve Local Leak Rate Exceeded 11.5 SCFH.
254/94-005 Unit-1 A-Loop Main Steam Isolation Valve Local Leak Rate Exceeded 11.5 SCFH.
G.
COMPONENT FAILURE DATA
4 Component failure data is not available at this time since repairs have not been completed.
Failure data will be included in the Supplemental Report.
4 d
LER265\\95\\001.WPF