05000316/LER-2016-001, Regarding Manual Reactor Trip Due to Moisture Separator Heater Expansion Joint Failure
| ML16246A005 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 08/31/2016 |
| From: | Lies Q AEP Indiana Michigan Power Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| AEP-NRC-2016-68 LER 16-001-00 | |
| Download: ML16246A005 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| LER closed by | |
| IR 05000315/2017001 (1 May 2017) | |
| 3162016001R00 - NRC Website | |
text
IN DIANA MICHIGAN POWER A unit of American Electric Power August 31, 2016 Docket No.: 50-316 U.S. Nuclear Regulatory Commission A TIN: Document Control Desk 11555 Rockville Pike Rockville, MD 20852 Donald C. Cook Nuclear Plant Unit 2 LICENSEE EVENT REPORT 316/2016-001-00 Indiana Michigan Power Cook Nuclear Plant One Cook Place Bridgman, Ml 49106 lndianaMichiganPower.com AEP-NRC-2016-68 10 CFR 50.73 Manual Reactor Trip Due to Moisture Separator Heater Expansion Joint Failure In accordance with 10 CFR 50.73, Licensee Event Report (LER) System, Indiana Michigan Power Company, the licensee for Donald C. Cook Nuclear Plant Unit 2, is submitting as an enclosure to this letter the following report:
LER 316/2016-001-00: Manual Reactor Trip Due to Moisture Separator Heater Expansion Joint Failure 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, Q.l~J.~
Site Vice President MPH/mil
Enclosure:
LER 316/2016-001-00: Manual Reactor Trip Due to Moisture Separator Heater Expansion Joint Failure
U. S. Nuclear Regulatory Commission Page 2 c:
R. J. Ancona - MPSC A. W. Dietrich - NRG Washington, DC MDEQ - RMD/RPS NRG Resident Inspector C. D. Pederson - NRG Region Ill A. J. Williamson - AEP Ft. Wayne AEP-NRC-2016-68
Enclosure to AEP-NRC-2016-68 LER 316/2016-001-00 Manual Reactor Trip Due to Moisture Separator Heater Expansion Joint Failure
NRCFORM366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 1013112018 (11-2015)
, the NRG may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 2. DOCKET YEAR 05000316 2016
- 3. LER NUMBER SEQUENTIAL NUMBER 001 REV NO.
00 On July 6, 2016, Donald C. Cook Nuclear Plant Unit 2 Reactor [RCT] was operating at 100 percent power. At 0038, the Unit 2 Reactor was manually tripped due to the sudden failure of the balance bellows on the Unit 2 B Right Moisture Steam Reheater (MSR) [RHTR] crossover expansion joint [EXJ], which also resulted in damage to the west wall of the turbine building. As a result of damage to the MSR, main condenser [COND] vacuum slowly degraded and heat removal was accomplished through the Steam Generator Power Operated Relief Valves (PORVs) [SB][RV].
EVENT DESCRIPTION
On July 6, 2016, at 0038, Donald C. Cook Nuclear Plant Unit 2 Reactor was manually tripped from 100 percent power, after operators heard a loud noise and observed a large steam leak in the vicinity of the B Right MSR crossover line. This information resulted in the operators' decision to manually trip the Unit 2 Reactor. The Auxiliary Feedwater System [BA] automatically initiated to maintain Steam Generator levels within acceptable control limitations, and the steam release immediately dissipated upon trip of the main turbine [TRB]. Due to the B Right MSR crossover expansion joint damage, main condenser vacuum slowly degraded and heat removal was accomplished through the steam generator PORVs. However, main feedwater flow could have been established, if needed, and therefore this trip is considered to be "uncomplicated". Plant response was within acceptable control limitations and all system parameters stabilized at normal post-trip levels. The Emergency Plan was entered at 0050 due to declaration of an Unusual Event based on EAL N-7, Explosion (Report by plant personnel of an unanticipated EXPLOSION within the protected area boundary resulting in visible damage to permanent structures or equipment). The Unusual Event was terminated at 0207. A Root Cause Evaluation (RCE) of the bellows failure is in progress. A supplement to this Licensee Event Report (LER) with additional information will be submitted upon completion of the RCE.
COMPONENT 2-XJ-113-6: B Right MSR Reheat Steam to Low Pressure Turbine B Outlet Elbow Expansion Joint Bellows.
ASSESSMENT OF SAFETY CONSEQUENCES
NUCLEAR SAFETY This event was not significant with respect to the health and safety of the general public since all safe.ty systems functioned per design during this event.
INDUSTRIAL SAFETY No personnel injuries occurred as a result of this event. However, this event was significant with respect to the safety of plant personnel due to the potential for personnel injury resulting from the bellows failure.
RADIOLOGICAL SAFETY There was no actual or potential radiological safety hazard resulting from the bellows failure.
PROBABILISTIC RISK ASSESSMENT (PRA)
, the NRG may not conduct or sponsor, and a person is not reouired to resoond to, the information collection.
- 2. DOCKET YEAR 05000316 2016
- 3. LER NUMBER SEQUENTIAL NUMBER 001 REV NO.
00 This event was low risk significant from a PRA perspective because the MSR is not a risk significant component in the PRA. Reactor trips with a loss of main feedwater or condenser heat sink have a Conditional Core Damage Probability of approximately 7E-7 and a Conditional Large Early Release Probability of 9.5E-8, which is considered low in comparison to other risk significant events.
ROOT CAUSE The Root Cause Evaluation is currently in progress. 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 Unit 2 Reactor was manually tripped and the steam release immediately dissipated upon trip of the main turbine. Repairs to the expansion joint bellows were completed on July 11, 2016.
Corrective Action Planned Corrective Actions will be included in the supplement.
PREVIOUS SIMILAR EVENTS
LERs for both units for the past three years were reviewed for similar events related to 1 O CFR 50.73(a)(2)(iv)(A) reporting criteria for system actuation. The following event was identified:
LER-316-2013-001-00 "Unit 2 Manual Reactor Trip due to Lowering Steam Generator Level" LER-315-2014-001-00 "Manual Reactor Trip Due to Lake Debris Intrusion Causing Degraded Forebay Conditions" LER-316-2015-001-00 "Manual Rector Trip Due to a Secondary Plant Transient" Although these three events were similar with respect to Manual Reactor Trips, the Corrective Actions resulting from these LERs are not expected to have prevented this occurrence.