05000316/LER-2006-005, Re Failure to Comply with Technical Specification Surveillance Requirement 3.6.1.1
| ML062280061 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 08/02/2006 |
| From: | Weber L Indiana Michigan Power Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| AEP:NRC:2573-33 LER 06-005-00 | |
| Download: ML062280061 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 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)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3162006005R00 - NRC Website | |
text
A unit ofAmerican Electric Power Indiana Michigan Power Cook Nuclear Plant One Cook Place Bridgman, Ml 49106 AEPcom August 2, 2006 AEP:NRC:2573-33 10 CFR 50.73 10 CFR 50.4 Docket No. 50-316 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop O-PI-17 Washington, DC 20555-0001 Donald C. Cook Nuclear Plant Unit 2 LICENSEE EVENT REPORT 316/2006-005-00 FAILURE TO COMPLY WITH TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENT 3.6.1.1 In accordance with the criteria established by 10 CFR 50.73 entitled Licensee Event Report System, the following report is being submitted:
LER 316/2006-005-00:
"Failure to Comply with Technical Specification Surveillance Requirement 3.6.1.1" There are no commitments contained in this submittal.
Should you have any questions, please contact Ms. Susan D. Simpson, Regulatory Affairs Manager, at (269) 466-2428.
Sincerely, Lawrence J. Weber Plant Manager HLE/rdw Attachment
U. S. Nuclear Regulatory Commission AEP:NRC:2573-33 Page 2 c:
J. L. Caldwell, NRC Region III K. D. Curry - AEP Ft. Wayne, w/o attachment J. T. King, MPSC - w/o attachment MDEQ - WHMD/RPMWS - w/o attachment NRC Resident Inspector P. S. Tam, NRC Washington DC
NRC Form 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES 6130/2007 (6-2004)
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. PAGE Donald C. Cook Nuclear Plant Unit 2 05000-316 1 of 3
- 4. TITLE Failure to Comply with Technical Specification Surveillance Requirement 3.6.1.1
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED SEQUENTIAL REVISION FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NUMBER MONTH DAY YEAR FACILITY NAME DOCKET NUMBER 04 10 2006 2006 005 00 08 02 2006
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) o 20.2201(b)
[I 20.2203(a)(3)(i)
[I 50.73(a)(2)(i)(C)
[E 50.73(a)(2)(vii)
Mode 6 0 20.2201(d)
[I 20.2203(a)(3)(ii)
E] 50.73(a)(2)(ii)(A)
El 50.73(a)(2)(viii)(A)
[I 20.2203(a)(1)
[] 20.2203(a)(4)
[I 50.73(a)(2)(ii)(B)
[E 50.73(a)(2)(viii)(B)
[o 20.2203(a)(2)(i) 0l 50.36(c)(1)(i)(A)
[I 50.73(a)(2)(iii)
El 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL
[: 20.2203(a)(2)(ii) 0l 50.36(c)(1)(ii)(A)
[I 50.73(a)(2)(iv)(A)
El 50.73(a)(2)(x) 0%
[1 20.2203(a)(2)(iii)
El 50.36(c)(2) 0l 50.73(a)(2)(v)(A)
[: 73.71(a)(4)
El 20.2203(a)(2)(iv)
[] 50.46(a)(3)(ii)
[I 50.73(a)(2)(v)(B)
I] 73.71(a)(5)
[I 20.2203(a)(2)(v)
El 50.73(a)(2)(i)(A)
[I 50.73(a)(2)(v)(C) tJ OTHER Specify in Abstract below
[: 20.2203(a)(2)(vi) 0 50.73(a)(2)(i)(B) 11 50.73(a)(2)(v)(D) or in (If more space is required, use additional copies of NRC Form (366A)
Conditions Prior to Event
Unit 2 was in refueling outage U2C16, in Mode 6.
Description of Event
On April 10, 2006, 2-SI-1 89, ECCS Safety Valves Discharge Header To Pressurizer Relief Tank Containment Isolation Check Valve [ISV], had a required surveillance inspection performed on it that potentially affected the valve's leak tightness prior to performing the required as-found leak rate test.
This resulted in being unable to meet the Technical Specification Surveillance Requirement (SR) 3.6.1.1 for Containment. SR 3.6.1.1 requires that leakage rate testing be performed in accordance with.
the Containment Leakage Rate Testing Program. Donald C. Cook Nuclear Plant's Containment -
Leakage Rate Testing Program specifies as-found testing prior to performing maintenance, repairs, or inspections that could reduce containment leakage integrity.
Valve 2-SI-1 89, ECCS Safety Valves Discharge Header To Pressurizer Relief Tank Containment Isolation Check Valve, was disassembled per Job Order Activity (JOA) R0267698-04 and an internal visual inspection performed per JOA R0267698-03 prior to performance of the required as-found 'type B&C leak rate test. The scheduling for these activities was correct and the appropriate logic ties were in the schedule.
This event is reportable in accordance with 10 CFR 50.73(a)(2)(i)(B), due to being unable to meet SR 3.6.1.1. This event occurred, and was identified, on April 10, 2006. The non-compliance with the Appendix J program requirements was identified on April 12, 2006. The initial review for applicability to 10 CFR 50.73 reporting requirements determined that the event was not required to be reported.
Subsequent review and discussion between the plant staff and the Nuclear Regulatory Commission determined, after the 60-day time frame had expired, that the event was required to be reported.
Therefore, this Licensee Event Report (LER) is being submitted greater than 60 days after the event.
Cause of Event
The cause of performing the required surveillance internal inspection prior to a required as-found leak rate test was a weakened barrier in that the work activity instructions did not require verification of completion of the as-found leak rate test prior to beginning the internal inspection.
Analysis of Event
The valve disassembly activities performed on 2-SI-189 on April 10, 2006, were for a required surveillance internal inspection. There were no outstanding corrective maintenance activities for 2-SI-189 nor was the valve leakage integrity suspect. Additionally, 2-SI-189 is not on an extended surveillance frequency under the Appendix J program as other more restrictive surveillance requirements outside of the Appendix J program require this valve to be disassembled and inspected on a refueling outage frequency.
(If more space is required, use additional copies of NRC Form (366A)
Prior to the valve disassembly activities performed on 2-SI-1 89 on April 10, 2006, the as-left B&C test performed during the previous Unit 2 refueling outage (October 19, 2004) was satisfactory. No maintenance was performed on 2-SI-1 89 between the satisfactory test on October 19, 2004 and April 10, 2006. Additionally, no flow is expected past this valve during the operating cycle and a review of station Condition Reports between October 2004 and April 2006 did not identify any instances where there was flow past this valve. Thus, it can be concluded that the valve remained in the closed seated position from the satisfactory October 2004 as-left B&C test throughout the last operating cycle.
2-SI-1 89 is a containment isolation check valve that provides a one-way flow path from Residual Heat Removal [BPJ safety valve 2-SV-1 04W [RV] to the Pressurizer Relief Tank [TK]. Failure of this valve does not contribute to core damage frequency, and additional component failures would be required for the failure to contribute to large early release frequency, which would still be well below 1.0 E-8.
Therefore, this event was not risk significant.
Corrective Actions
Corrective Actions taken were performance of an acceptable as-left leak rate test following internal inspection of 2-SI-1 89, and revising the surveillance internal inspection work activity for 2-SI-1 89 to verify completion of the as-found test prior to performing valve disassembly.
Corrective Actions to be taken include revising all recurring work activities for Appendix J components that potentially affect a valve's leak tightness to require a verification of completion of the as-found B&C type leak rate test prior to performing an activity.
Previous Similar Events
A review was conducted of station Condition Reports and LERs for the previous 3 years. No similar events were identified.