05000263/LER-2012-003-01, Regarding Automatic Reactor Scram During Maintenance on 4160V 12-Bus Ammeter
| ML13022A480 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 01/18/2013 |
| From: | Schimmel M Northern States Power Co, Xcel Energy |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| L-MT-13-018 LER 12-003-01 | |
| Download: ML13022A480 (4) | |
| 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 10 CFR 50.73(a)(2)(iv)(B), System Actuation |
| 2632012003R01 - NRC Website | |
text
Xcel Energy January 18, 2013 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Monticello Nuclear Generating Plant Docket 50-263 Renewed Facility Operating License No. DPR-22 Monticello Nuclear Generating Plant 2807 W County Road 75 Monticello, MN 55362 L-MT-13-018 10 CFR 50.73 LER 2012-003-01 "Automatic Reactor Scram during Maintenance on 4160V 12-Bus Ammeter" A supplement to Licensee Event Report (LER) for this occurrence is attached.
Summary of Commitments This letter contains no new commitments and no revisions to existing commitments.
Mark A. Schimmel Site Vice-President, Monticello Nuclear Generating Plant Northern States Power Company-Minnesota Enclosure cc:
Regional Administrator, Region III, USNRC Project Manager, Monticello Nuclear Generating Plant, USNRC Resident Inspector, Monticello Nuclear Generating Plant, USNRC
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150*0104 EXPIRES 10/31/2013 (10-2010)
, the NRC may not conduct or sponsor, and a person is no reauired to respond to, the information collection.
- 3. PAGE Monticello Nuclear GeneratinQ Plant 05000 - 263 1 OF 3
- 4. Til LE Automatic Reactor Scram DurinQ Maintenance on 4160V 12-Bus Ammeter
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED YEAR I SEQUENTIAL I REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR NUMBER NO MONTH DAY YEAR 05000 FACILITY NAME DOCKET NUMBER 09 25 2012 2012 - 003 - 01 01 18 2013 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) o 20.2201(b) o 20.2203(a)(3)(i) o 50.73(a)(2)(i)(C) o 50.73(a)(2)(vii) 1 o 20.2201(d) o 20.2203(a)(3)(ii) o 50.73(a)(2)(ii)(A) o 50.73(a)(2)(viii)(A) o 20.2203(a)(1) o 20.2203(a)(4) o 50.73(a)(2)(ii)(B) o 50.73(a)(2)(viii)(B) o 20.2203(a)(2)(i) o 50.36(c)(1)(i)(A) o 50.73(a)(2)(iii) o 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL o 20.2203(a)(2)(ii) o 50.36(c)(1 )(ii)(A)
~ 50.73(a)(2)(iv)(A) o 50.73(a)(2)(x) o 20.2203(a)(2)(iii) o 50.36(c)(2) o 50.73(a)(2)(v)(A) o 73.71(a)(4) 100%
o 20.2203(a)(2)(iv) o 50.46(a)(3)(ii) o 50.73(a)(2)(v)(B) o 73.71 (a)(5) o 20.2203(a)(2)(v) o 50.73(a)(2)(i)(A) o 50.73(a)(2)(v)(C) o OTHER Specify in Abstract below or in o 20.2203(a)(2)(vi) o 50.73(a)(2)(i)(B) o 50.73(a)(2)(v)(D)
EVENT DESCRIPTION
Monticello Nuclear Generating Plant (MNGP) was in Mode 1 at 100% power prior to the event.
- 3. PAGE 2 OF 3 On September 25, 2012, work was being performed by relay technicians to test the 2R Transformer [XFMR]
to 12-Bus [BU] local and remote ammeter switches [IS]. The 5-pole Current Transformer isolation switch was opened to isolate the protective relaying and ammeter circuits from the 2R source to 12-Bus breaker. A Doble set was connected downstream of the open isolation switch with one lead connected to each phase and one lead to the neutral of the relaying / ammeter circuits for the purpose of providing a three-phase AC input to the ammeters and permit testing of the ammeter switches.
At approximately 1042 hours0.0121 days <br />0.289 hours <br />0.00172 weeks <br />3.96481e-4 months <br />, the phase outputs of the Doble sets were turned on one-by-one to provide 2.5 Amps per phase. After turning on the first phase the 2R to 12-Bus Feeder, the time neutral over current relay [RL Y] actuated, causing a lockout of 12-Bus. The lockout of 12-Bus resulted in 12-Reactor Feedwater Pump [SK] and 12-Reactor Recirculation Pump [AD] tripping. The operating crew took actions in accordance with plant procedures for a loss of 12-Bus, 12-Feedwater trip, 12-Recirculation Pump trip, and neutron flux oscillations. Water level initially lowered to approximately +23 inches and then began to rise with the Feedwater Regulating Valves [V] in automatic. It was expected that the Digital Feedwater Control System (DFCS) in automatic would stabilize reactor level without operator actions. Prior to water level reaching +40 inch alarm setpoint, reactor operators determined that if level reached +46 inches they would manually scram the reactor. At 1044 hours0.0121 days <br />0.29 hours <br />0.00173 weeks <br />3.97242e-4 months <br />, while completing the 3-way communication for manually scramming the reactor, the Reactor [RCT] water level reached the Hi Hi setpoint (+48 inches) resulting in a Turbine Generator [TB]
load reject, initiating a trip of the Turbine [TA], and subsequent Reactor scram. Reactor water level lowered resulting in a Primary Containment Isolation signal at a water level of +9 inches.
There were no inoperable systems, structures, or components prior to the event that contributed to the event.
EVENT ANALYSIS
This event is being reported in accordance with 10 CFR 50.73(a)(2)(iv)(A) as an event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph 10 CFR 50.73(a)(2)(iv)(B). Specifically, the Reactor Protection System (RPS) and the Primary Containment Isolation System (PCIS) actuations.
SAFETY SIGNIFICANCE
The safety objective of both RPS and PCIS are to provide timely protection at the onset of conditions that could challenge the integrity of the fuel barrier and nuclear system process barriers. The RPS prevents the release of radioactive material from the fuel and nuclear system process barriers by terminating excessive temperature and pressure increases through the initiation of an automatic plant shutdown. PCIS prevents release of radioactive materials by isolating the reactor vessel and closing containment where required. For this event, the RPS, PCIS, and plant safety systems functioned as designed and fuel and nuclear system process barriers remained intact. Consequently, the event did not have an adverse impact on the health and safety of the public and was not considered a safety system functional failure.
CAUSE
- 3. PAGE 3 OF 3 The root cause for the 12-Bus lockout was determined to be that fleet work management guidance does not require the appropriate level of detail in work plans needed to expose the potential plant impact when injecting energy into plant structures, systems, and components.
Additionally, a second root cause focused on the level transient determined that the DFCS was not designed to control reactor pressure vessel level below +48 inches on a 12-Bus lockout from 100% power. The organization had a mind-set based on procedure bases and training that DFCS in automatic would maintain reactor level below +48 inches following a 12-Bus lockout.
CORRECTIVE ACTION
The immediate corrective action was to stop work, remove the Doble test equipment, and reset the 12-Bus lockout. Long-term corrective actions include revising Work Management guidance to require the appropriate level of detail in work plans needed to expose the potential plant impact when injecting energy into systems, structures, or components.
Classroom and simulator training for all licensed operators will be performed on this event and will include changes to the abnormal operating procedures to address DFCS.
PREVIOUS SIMILAR EVENTS
There have been no similar licensee event reports in the past three years.
ADDITIONAL INFORMATION
Energy industry identification system (EllS) codes are identified in the text within brackets [xx].