05000316/LER-2015-001, Regarding Manual Reactor Trip Due to a Secondary Plant Transient
| ML15169B035 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 06/17/2015 |
| From: | Gebbie J Indiana Michigan Power Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| AEP-NRC-2015-45 LER 15-001-00 | |
| Download: ML15169B035 (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 |
| LER closed by | |
| IR 05000315/2017001 (1 May 2017) | |
| 3162015001R00 - NRC Website | |
text
INDIANA Indiana Michigan Power MICHIGAN Cook Nuclear Plant pO1:I R6 One Cook Place Bridgman, MI 49106 A unit ofAmerican Electric Power Indiana MichiganPower.com June 17, 2015 AEP-NRC-2015-45 10 CFR 50.73 Docket No.: 50-316 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk 11555 Rockville Pike, Rockville, MD 20852 Donald C. Cook Nuclear Plant Unit 2 LICENSEE EVENT REPORT 316/2015-001-00 Manual Reactor Trip Due To A Secondary Plant Transient In accordance with 10 CFR 50.73, Licensee Event Report System, Indiana Michigan Power Company, the licensee for Donald C. Cook Nuclear Plant (CNP) Unit 2, is submitting as an enclosure to this letter the following report:
LER 316/2015-001-00: "Manual Reactor Trip Due To A Secondary Plant Transient" There are no commitments contained in this submittal.
Should you have any questions, please contact Mr. Michael K. Scarpello, Regulatory Affairs Manager, at (269) 466-2649.
Sincerely, Joel P. Gebbie Site Vice President JEN/amp Enclosure c:
A. W. Dietrich - NRC Washington, DC J. T. King - MPSC MDEQ - RMD/RPS NRC Resident Inspector C. D. Pederson - NRC Region III A. J. Williamson - AEP Ft. Wayne
Enclosure to AEP-NRC-2015-45 LER 316/2015-001-00 Manual Reactor Trip Due To A Secondary Plant Transient
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 01/3112017 02-2014)
Estimated burden per response to comply with this mandatory collection request: 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />.
Reported lessons learned are incorporated into the licensing process and fed back to industry.
Send comments regarding burden estimate to the FOIA, Privacy and Information Collections SEE EVENT REPORT (LER Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by Linternet e-mail to Infocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and (See Page 2 for required number of Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC digi's/characters for each b
,ock) 20503. If a means used to impose an information collection does not display a currently valid OMB control number, 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 05000316 1 OF 4
- 4. TITLE Manual Reactor Trip Due To A Secondary Plant Transient
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED IACLIYIAM OCKETNUMBE MONTH DAY YEAR YEAR SEQUENTIAL REV MONTH DAY YEAR 05000 NUMBER NO.05000 FACILITY NAME DOCKET NUMBER 04 23 2015 2015-001
- - 00 06 17 2015 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
E] 20.2201(b)
[]
20.2203(a)(3)(i)
[I 50.73(a)(2)(i)(C)
El 50.73(a)(2)(vii)
I] 20.2201(d)
El 20.2203(a)(3)(ii)
E] 50.73(a)(2)(ii)(A)
[]
50.73(a)(2)(viii)(A) 2L 20.2203(a)(1)
D 20.2203(a)(4)
L] 50.73(a)(2)(ii)(B)
[] 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 LI 20.2203(a)(2)(ii)
El 50.36(c)(1)(ii)(A) 50.73(a)(2)(iv)(A)
LI 50.73(a)(2)(x)
LI 20.2203(a)(2)(iii)
LI 50.36(c)(2)
El 50.73(a)(2)(v)(A)
El 73.71(a)(4) 00 20.2203(a)(2)(iv)
El 50.46(a)(3)(ii)
[] 50.73(a)(2)(v)(B)
El 73.71(a)(5) 0E 20.2203(a)(2)(v)
Z 50.73(a)(2)(i)(A)
[J 50.73(a)(2)(v)(C)
El OTHER [I 20.2203(a)(2)(vi)
El 50.73(a)(2)(i)(B)
LI 50.73(a)(2)(v)(D)
Specify in Abstract below or in
[:1 0.7(a)()(V(D)
Secondary Heat Sink was maintained during the transient by feeding all four S/Gs using the Unit 2 MDAFW pumps with steam relief capability available with S/G Power Operated Relief Valves [SB] [PCV] (PORV). Additionally, the Group I Steam Dump valve which was not replaced during the outage was in service before the event and was controlling RCS temperature normally. Although available, neither the TDAFW pump nor the Main Feedwater [SJ]
Pumps were required in order to maintain secondary heat sink. Plant electrical safety busses [BU] were powered from the preferred offsite electrical power source [EB] before and after the manual Reactor trip.
The manual Reactor Protection System (RPS) actuation was reported via Event Notification 51004 in accordance with 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A), and 10 CFR 50.72(b)(2)(i). The valid RPS actuation and the completion of the plant shutdown required by TS are reportable as a Licensee Event Report (LER) in accordance with 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(2)(i)(A) respectively.
EVENT ANALYSIS
Event analysis will be included in the supplement.
COMPONENT 10-inch Copes Vulcan (SPX) Severe Duty class 600 Generation III tandem trim design air operated control valves.
ASSESSMENT OF SAFETY CONSEQUENCES
NUCLEAR SAFETY Actual Impact Failure of the two Steam Dump Valves resulted in minimal nuclear safety impact. All control rods fully inserted as a result of the manual Reactor trip, and were unaffected by the RCS cooldown. The reduction in RCS Tavg did not result in a loss of required shutdown margin or in an uncontrolled return to criticality during the post-trip response.
RCS heat sink was maintained using the S/Gs with make-up from the MDAFW pumps and steam relief to the atmosphere via S/G PORVs.
Potential impact Resulting from a RCS cooldown, the following are potential nuclear safety impacts:
Inability to maintain the Reactor subcritical following shutdown Loss of effectiveness of ex-core nuclear instrument [DET] trip setpoint effectiveness Exceeding Pressurizer [PZR] TS thermal stress limits Inability to maintain the condenser as a secondary heat sink Loss of RCS pressure control due to inability to maintain Pressurizer level INDUSTRIAL SAFETY Actual Impact There was no actual industrial safety hazard resulting from the internal mechanical failure of the Steam Dump Valves. No failure of the Main Steam or Main Condenser pressure boundaries occurred during this event.
Potential Impact The magnitude of the forces involved in the mechanical transient was sufficient to challenge the valve actuator, but was not sufficient to challenge the steam piping pressure boundary. As long as the valve and actuator remained bolted together, an external steam leak or release would not be expected.
RADIOLOGICAL SAFETY Actual Impact There was no actual radiological safety hazard resulting from the internal mechanical failure of the Steam Dump Valves.
Potential Impact The potential failure of the ability to isolate a failed steam dump flow path could result in the inability to control Main Steam pressure. During conditions such as a Steam Generator [SG] Tube Rupture where the Main Condenser remains available, the loss of steam pressure control would necessitate isolation of the Main Steam lines. This would require RCS cooldown using S/G Power Operated Relief Valves. Potential dose resulting from this cooldown would remain bounded by the dose analysis.
PROBABILISTIC RISK ASSESSMENT (PRA)
A PRA risk assessment was performed on the event. The analysis concluded that the Incremental Conditional Core Damage Probability and Incremental Conditional Large Early Release Probability are under the Regulatory Guide 1.174 criteria for significant events. Therefore, this event was considered to be of very low risk.
ROOT CAUSE The Root Cause Evaluation is ongoing; a supplement to this LER will be submitted upon completion.
CONTRIBUTING CAUSES
Contributing causes will be provided in the supplement.
CORRECTIVE ACTIONS
Immediate Corrective Action Taken The two newly installed Steam Dump valves that failed open (of the three installed during the outage) were removed and replaced with the originally installed valves.
A Temporary Modification has been implemented to allow, operation throughout the remainder of the operating cycle with the remaining Copes Vulcan Steam Dump valve isolated.
Corrective Action to Preclude Repetition (CATPR)
The CATPR will be included in the supplement.
Additional Corrective Actions
Planned Planned corrective actions will be included in the supplement, as applicable.
PREVIOUS SIMILAR EVENTS
Pertinent similar events will be included in the supplement, as applicable.