05000278/LER-2009-003, Inoperable Containment Isolation Valve Results in Condition Prohibited by Technical Specifications
| ML091350044 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 05/12/2009 |
| From: | Stathes G Exelon Nuclear |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LER 09-003-00 | |
| Download: ML091350044 (6) | |
| 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 |
| 2782009003R00 - NRC Website | |
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Exelon Nuclear www.exeloncorp.com Peach Bottom Atomic Power Station Nucea-1848 Lay Road Delta, PA 17314-9032 10CFR 50.73 May 12, 2009 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Peach Bottom Atomic Power Station (PBAPS) Unit 3 Facility Operating License No. DPR-56 NRC Docket No. 50-278
Subject:
Licensee Event Report (LER) 3-09-03 This LER reports a condition prohibited by Technical Specifications involving the inoperability of a Primary Containment Isolation Valve.
In accordance with NEI 99-04, the regulatory commitment contained in this correspondence is to restore compliance with the regulations. The specific methods that are planned to restore and maintain compliance are discussed in the LER. If you have any questions or require additional information, please do not hesitate to contact us.
Sincerely, Garey L.Stathes Plant Manager Peach Bottom Atomic Power Station GLS/djf/IR 898030 Attachment cc:
S. J. Collins, US NRC, Administrator, Region I F. L. Bower, US NRC, Senior Resident Inspector R. R. Janati, Commonwealth of Pennsylvania S. Grey, State of Maryland P. Steinhauer, PSE&G, Financial Controls and Co-owner Affairs INPO Records Center CCN: 09-39 AJ4
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 08/31/2010 (9-2007)
, the NRC may digits/characters for each block) not conduct or sponsor, and a person is not required to respond to, the infnrmatinn nntlnc.tinn.
- 3. PAGE Peach Bottom Atomic Power Station Unit 3 05000278 1 OF 5
- 4. TITLE Inoperable Containment Isolation Valve Results in Condition Prohibited by Technical Specifications
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
[
- 8. OTHER FACILITIES INVOLVED FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR SEQUENTIAL REV MONTH DAY YEAR NO 05000 NUMBER NO. I
_MONTH
__DAY
__YEAR I rFACILITY NAME DOCKET NUMBER 03 26 2009 09 03 00 05 12 20091 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
El 20.2201(b)
[] 20.2203(a)(3)(i)
El 50.73(a)(2)(i)(C)
[I 50.73(a)(2)(vii) 1 El 20.2201(d)
El 20.2203(a)(3)(ii)
El 50.73(a)(2)(ii)(A)
[I 50.73(a)(2)(viii)(A)
El 20.2203(a)(1)
[I 20.2203(a)(4)
El 50.73(a)(2)(ii)(B)
[I 50.73(a)(2)(viii)(B) 0_ 20.2203(a)(2)(i)
El 50.36(c)(1)(i)(A)
[I 50.73(a)(2)(iii)
El 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL El 20.2203(a)(2)(ii)
El 50.36(c)(1)(ii)(A)
[I 50.73(a)(2)(iv)(A)
El 50.73(a)(2)(x)
El 20.2203(a)(2)(iii)
El 50.36(c)(2)
El 50.73(a)(2)(v)(A)
El 73.71 (a)(4) 100 %
El 20.2203(a)(2)(iv)
El 50.46(a)(3)(ii)
El 50.73(a)(2)(v)(B)
[1 73.71 (a)(5)
El 20.2203(a)(2)(v)
El 50.73(a)(2)(i)(A)
El 50.73(a)(2)(v)(C)
[E OTHER El 20.2203(a)(2)(vi)
[
50.73(a)(2)(i)(B)
[E 50.73(a)(2)(v)(D)
Specify in Abstract below or in
Analysis of the Event
There were no actual safety consequences associated with this event.
The 3D RHR Pump Suppression Pool Suction Isolation Valve (MO-3-10-013D) is the isolation valve that is in the suction flow path for the 3D RHR pump. The 3D RHR pump provides a variety of functions including supplying water for the following modes of RHR: Low Pressure Coolant Injection (LPCI),
Suppression Pool Cooling (SPC), Suppression Pool Spray, Containment Spray and Shutdown Cooling (SDC).
The diagnostic test data obtained on 3/26/09 indicated that the as-found thrust of the valve at the point of torque switch actuation was less than the required thrust value. This condition may have resulted in the valve prematurely stopping motion in the closed direction under design basis demand conditions.
The MO-3-10-013D is a normally open valve and the valve does not have any automatic open or close signals. It can be remotely-closed by use of a key-locked switch in the main control room. The conditions for which the valve must be closed include the following functions:
SDC Mode Alignment Containment Isolation for Primary Containment (Suppression Pool) Penetration N-226 Engineering has determined that for the SDC mode of RHR, there were no concerns with the ability of this valve to close in its degraded condition.
However, for its containment isolation function to isolate containment penetration N-226, the ability of the valve to close during design basis conditions involving an elevated pressure in the Suppression Pool (e.g., Loss-of-Coolant Accident conditions), the valve may not have been able to perform its design function to close.
For other lower pressure conditions other than the design event, the MO-3-10-013D was capable of performing its closure function.
Additionally, for any design basis conditions involving no flow through the penetration, the valve would be capable of performing its design function to close.
Penetration N-226 (i.e., 3D RHR suction from the Suppression Pool) is isolated by the MO 10-01 3D. The water level in the Suppression Pool serves as the second barrier for containment atmosphere fission product releases.
Therefore, there were no concerns with containment atmosphere isolation capability due to the presence of the water level in the Suppression Pool.
In the very unlikely case of a failure of an active mechanical device (e.g., RHR pump failure causing a seal leak) coupled with a design basis event Suppression Pool pressure condition (e.g., Loss-of-Coolant Accident conditions), the MO-3-10-013D may have not been able to be fully closed. For most seal leaks, however, the leak rate would be minor and well within the capability of the reactor building sump system. For larger leaks, the 3D RHR room water level alarm would alert Operations personnel to the condition. Operations personnel would remotely actuate the MO-3-10-013D to terminate the leak. The valve would begin to close, but might I
Analysis of the Event, continued stop travel during the valve stroke as the pressure differential across the valve increases. The valve would stop motion as a result of torque switch actuation.
Based on conservative assumptions, the valve would travel approximately 82% closed, thereby isolating approximately 87% of the flow. Other manual actions could also be pursued to remotely close the valve.
However, as a result of internal flood design requirements of the Reactor Building, the affect of any un-isolated water leakage would preserve operability of redundant emergency core cooling or containment cooling subsystem subsystems.
This event is not considered as risk significant.
Cause of the Event
The cause of the MO-3-10-013D degradation is due to damage to the motor-operator actuator stem nut threads as a result of identified grease hardening. The stem nut is the device in the motor-operator actuator that interfaces with the threaded valve stem.
The motor-operator actuator turns the stem nut resulting in movement of the valve stem. The damage to the stem nut threads included broken threads and the loss of some thread pieces.
There was no significant degradation to the valve stem threads.
A root cause investigation was performed to examine the underlying reasons for the valve condition. This investigation focused on MOV actuator stem lubricant performance and the preventive maintenance program frequencies and actions for MOVs.
The Limitorque Corporation supplied the motor-operator (Model SMB 2-40). The grease was manufactured by Exxon (Nebula EP-1).
Corrective Actions
The MO-3-10-013D valve was repaired and returned to service on 3/27/09.
An extensive extent-of-condition evaluation for other susceptible MOVs on both Units 2 and 3 was performed.
This included inspection, testing, cleaning, and/or re-lubrication as appropriate for more than 20% of the MOV population.
A root cause investigation was completed. Appropriate corrective actions will be performed to upgrade the MOV program as necessary including upgrading the preventive maintenance feedback process, adjusting preventive maintenance and diagnostic testing intervals, and revising appropriate procedures to address procedural weaknesses.
Previous Similar Occurrences There were no previous LERs identified relating to inoperable MOVs resulting from greasing deficiencies.
However, two previous prompt notifications were recently made to the NRC regarding MOVs with similar greasing deficiencies.
On 3/12/09, the Unit 2 High Pressure Coolant Injection (HPCI) Suppression Pool Suction Inboard Isolation Valve (MO-2-23-058) experienced degradation during valve stroking in preparation for routine surveillance testing. On 3/21/09, the Unit 3 High Pressure Coolant Injection (HPCI) Suppression Pool Suction Outboard Isolation Valve (MO-3-23-057) was also identified to have some degradation during surveillance testing.
Both of these valve degradation occurrences were evaluated and it was determined that neither valve was inoperable or had a loss of safety function. Therefore, the prompt notifications were retracted. However, these occurrences were similar in nature to the concerns that existed with the MO-3-10-013D.
The MO-3-10-013D concerns were identified as a result of extent-of-condition testing performed as a result of these previous degradations of the MO-2-23-058 and MO-3-23-057.
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