05000316/LER-2006-006, Re Failure to Comply with Technical Specification 3.5.2, ECCS - Operating
| ML072130022 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 07/18/2007 |
| From: | Jensen J Indiana Michigan Power Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| AEP:NRC:2573-38 LER 06-006-00 | |
| Download: ML072130022 (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 |
| 3162006006R00 - NRC Website | |
text
A unit of American Electric Power July 18, 2007 Indiana Michigan Power Cook Nuclear Plant One Cook Place Bridgman, MI 49106 AEPcom AEP:NRC:2573-38 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'.006-00 FAILURE: 'TO COMPLY WITH TECHNICAL SPECIFICATION 36.52,1 ECCS.
ECCG OPERATION--..,
. In accodance With the. criteria established by ro CFR 50.73. -Li66ehsee.'Even :Report' Syst*m, the followinrg report is being'submitted:
- LER".31i6/2006-006-00:
"Failure to, Comply with-Technical, Specification ;.5.2, ECCS
ýOperation" ECC
,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.
4e-h N. Jensen Site Vice President RAM/rdw Attachment 1tjzP_
U. S. Nuclear Regulatory Commission AEP:NRC:2573-38 Page 2 c:
J. L. Caldwell, NRC Region III K. D. Curry - AEP Ft. Wayne, w/o attachment INPO Records Center 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 4
- 4. TITLE Failure to Comply with Technical Specification 3.5.2, ECCS - Operating
- 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 19 2006 2006 006 00 07 18 2007
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 1 0l 20.2201(b)
El 20.2203(a)(3)(i)
El 50.73(a)(2)(i)(C)
Z 50.73(a)(2)(vii) 0l 20.2201(d)
El 20.2203(a)(3)(ii)
El 50.73(a)(2)(ii)(A)
E] 50.73(a)(2)(viii)(A) 0l 20.2203(a)(1) 0l 20.2203(a)(4)
El 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
El 20.2203(a)(2)(i)
[E 50.36(c)(1)(i)(A)
El 50.73(a)(2)(iii)
El 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL El 20.2203(a)(2)(ii)
[] 50.36(c)(1)(ii)(A)
E] 50.73(a)(2)(iv)(A)
El 50.73(a)(2)(x) 100%
El 20.2203(a)(2)(iii)
/
E] 50.36(c)(2)
[] 50.73(a)(2)(v)(A)
El 73.71(a)(4)
El 20.2203(a)(2)(iv)
El 50.46(a)(3)(ii)
El 50.73(a)(2)(v)(B) 73.71 (a)(5)
E] 20.2203(a)(2)(v)
El 50.73(a)(2)(i)(A)
El 50.73(a)(2)(v)(C)
HOTHER Specify in Abstract below El 20.2203(a)(2)(vi)
[
50.73(a)(2)(i)(g)
El 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 -
100% power
Description of Event
On April 10, 2006, Donald C. Cook Nuclear Plant (CNP) tested the Unit 2 "B" Residual Heat Removal (RHR) system heat exchanger outlet safety valve and on April 19,
The safety valves for both trains of.RHR had an unsatisfactory as-found lift pressure test (high).
Technical Specification (TS) 3.5, Emergency Core Cooling Systems (ECCS); 3.5.2, ECCS - Operating, requires two trains of ECCS to be OPERABLE when in MODES 1, 2,
and 3.
When one or more trains are inoperable, Condition A requires that the inoperable train(s) be restored to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Since a similar cause was determined for the unsatisfactory as-found lift pressures, this condition may have arisen over a period of time, and there is a likelihood that the affected safety valves on both trains of RHR may not have been OPERABLE during plant operation for a time longer than allowed by TS.
Therefore, this occurrence is considered reportable in accordance with 10 CFR 50.73(a) (2) (i) (B) as a Condition prohibited by CNP's TS and 10 CFR 50.73(a) (2) (vii) as a common cause of inoperability.
10 CFR 50.73(a) requires licensees to submit licensee event reports (LER) within 60 days of discovery of the event.
This LER is being submitted greater than 60.days after the event due to CNP's failure to recognize that the multiple test failures constituted a reportable condition..
Cause of Event
The apparent cause of the occurrence is nozzle disc bonding.
Analysis of Event
As described above, both Unit 2 RHR system heat exchanger outlet safety valves failed to initially lift at 1.25 times their design setpoint (design setpoint is 600 psig).
The failure of these safety valves to lift at the setpoint pressure has no direct influence on the behavior of other components, equipment, or conditions.
- Thus, these failures do not increase the probability.of any initiating event in the CNP Probabilistic Risk Assessment (PRA) model and have no impactd6n plant risk from that standpoint.
Failure of these safety. valves, could impact mitigation.
capabilities of the RHR and: safety injection (SI)
[BJ] systems under specific circumstances when these systems would be required to operate under high pressure NRC FORM.366A (1-2001)
(If more space is required, use additional copies of NRC Form (366A)
- - conditions.
Events which could lead to these conditions include a small break loss-of-coolant accident (SBLOCA) and an extended loss of heat sink accident.
Following one of these accidents, the potential exists in the SI-to-RHR injection piping configuration for the SI system to over pressurize the RHR system if the Reactor Coolant System (RCS) remains above 600 psig.
With either an SBLOCA or extendedloss of heat sink, an additional failure would have to occur for the failure of one of these safety valves to affect its associated RHR train.
Specifically, the RHR check valve upstream of the SI-to-RHR'tie-in would also have to fail to prevent backflow/leakage.
Thus, for the SI system to overpressurize the RHR system causing rupture or significant leakage of *the RHR piping, the following would be required:
- an SBLOCA or loss of heat sink,
" failure of the associated RHR train-related check valve.'
Control Room alarms would alert operators to abnormal pressure conditions in the RHR pump discharge lines'.
The annunicator response procedures provide direction for operator action.
In the'worst case, a single SI and RHR train could be disabled due to a large rupture.
If the piping ruptured and spilled water to either the SI *or RHR pump rooms, control room alarms would indicate significant leakage in the.associated room.
Room Sump Level Alarms indicate' leakage in the room.
The annunciator response procedure directs operatorsto determine the source of the leakage and isolate it.
Given the. training that Operations personnel receive on RHR and SI system operation, they would be expected to recognize that RHR/SI system operation was causing this alarm and take action to stop or minimize the impact.
Both RHR safety valves did lift on a second attempt within 5%:of the design setpoint and appeared to reseat satisfactorily.based on subsequent lift tests.
Given this behavior, there is a reasonable probability that the valves would have functioned to protect'the RHR system.
Valve opening would avoid significant RHR train damage and associated leakage.
Neither of the RHR safety valves is explicitly included in the PRA model.
In order for an RHR safety valve to actuate, failure of an RHR check valve upstream of the SSI-to-RHR tie-in must be assumed to cause failure of the associated SI train.
Implicitly, the PRA model does credit the outflow from these valves during Interfacing System LOCA events.
- However, in such events, full RCS pressure would be applied to the valve and would likely open it fully.
.On this basis, there is no quantitative PRA impact of the RHR heat exchanger outlet safety valves lifting'.
above their design setpoint.
Nonetheless, barriers exist to limit the.impact of this condition.
Specifically:
particular events are required (extended loss of auxiliary feedwater or SBLOCA),
- certain RCS conditions (RCS > 600 psig) are required, an additional failure (failure of the East RHR To Reactor Coolant Loops #1 and #4 Check Valve; or Westý RHR To Reactor Coolant Loops #2
& #3 Check Valve).
is required, and (If more space is required, use additional copies of NRC Form (366A)
- redundant alarms exist to alert.personnel to isolate/mitigate the leak.
The likelihood of these conditions occurring simultaneously is extremely small based upon probabilistic insights.
Corrective Actions
Both Unit 2 RHR system heat exchanger outlet safety valves were replaced with new, pre-tested valves and were declared operable.
The Unit 1 and Unit 2 RHR safety valves are necessary for RHR heat exchanger overpressure protection.
Expansion of the test population was performed in accordance with I&M's ISI testing program to identify the extent of condition and included one of the two Unit 1 valves.
This valve passed and did not exhibit indications of nozzle disc bonding.
I&M will continue to work with its vendors and industry peers to ensure it fully
- understands and addresses this condition, with expanded testing and adjustments to.
be per.formed as* appropriate.
Previous Similar Events*
05000316/2006-002-00, MSSV Trevi Testing Failures.
The causal evaluation and corrective actions for this previous similar event have been reviewed.
Based on the differing system operating parameters for..the main steam safety valves and the RHR heat exchanger outlet safety vaive~s, I&M. has determined that the.extent of condition review and corrective actions taken for LER 05000316/2006-002-00 could not have reasonably been expected to prevent the event being reported in this LER.