05000336/LER-2014-006, Regarding Millstone Power Station Dual Unit Reactor Trip on Loss of Offsite Power
| ML14211A526 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 07/24/2014 |
| From: | Adams W Dominion Nuclear Connecticut |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 14-349 LER 14-006-00 | |
| Download: ML14211A526 (10) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 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)(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)(1) |
| 3362014006R00 - NRC Website | |
text
- Dominion Dominion Nuclear Connecticut, Inc.
Rope Ferry Rd., Waterford, CT 06385 Mailing Address: P.O. Box 128 Warerford, CT 06385 dor.tom JUL.,2 4 2014 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Serial No.
MPS Lic/GJC Docket Nos.
License Nos.14-349 RO 50-336 50-423 DPR-65 NPF-49 DOMINION NUCLEAR CONNECTICUT, INC.
MILLSTONE POWER STATION UNITS 2 AND 3 LICENSEE EVENT REPORT 2014-006-00 MILLSTONE POWER STATION DUAL UNIT REACTOR TRIP ON LOSS OF OFFSITE POWER This letter forwards Licensee Event Report (LER) 2014-006-00 documenting an event at Millstone Power Station Units 2 and 3 on May 25, 2014. This LER is being submitted pursuant to 10 CFR 50.73(a)(2)(iv)(A).
If you have any questions or require additional information, please contact Mr. William D. Bartron at (860) 444-4301.
Sincerely, W. Matthew Adams Plant Manager - Millstone Attachments: 1 Commitments made in this letter: None
Serial No.14-349 Docket No. 50-336 50-423 Licensee Event Report 2014-006-00 Page 2 of 2 cc:
U.S. Nuclear Regulatory Commission Region I 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 M. C. Thadani NRC Senior Project Manager Millstone Units 2 and 3 U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 B-1 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station
Serial No.14-349 Docket No. 50-336 50-423 Licensee Event Report 2014-006-00 ATTACHMENT LICENSEE EVENT REPORT 2014-006-00 MILLSTONE POWER STATION DUAL UNIT REACTOR TRIP ON LOSS OF OFFSITE POWER MILLSTONE POWER STATION UNITS 2 AND 3 DOMINION NUCLEAR CONNECTICUT, INC.
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 01/31/2017 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 LICENSEE E(LER)
Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by EVENT REPORT internet 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 20503. If a means used to impose an information collection does not display a currently valid OMB digits/characters for each block) control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. PAGE Millstone Power Station Unit 2 05000336 1 OF 7
- 4. TITLE Millstone Power Station Dual Unit Reactor Trip on Loss of Offsite Power
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MO YEAR YEA SEQUENTIAL REV M
FACILITY NAME DOCKET NUMBER NUMBER NO MONTH DAY YEAR Millstone Power Station Unit 3 05000423 FACILITY NAME DOCKET NUMBER 05 25 2014 2014 "006 "00 I07 24 2014 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
L 20.2201(b)
L 20.2203(a)(3)(i)
D1 50.73(a)(2)(i)(C)
El 50.73(a)(2)(vii)
D 20.2201(d)
E 20.22(03(a)(3)(ii)
[I 50.73(a)(2)(ii)(A)
D 50.73(a)(2)(viii)(A)
El 20.2203(a)(1)
[
20.2203(a)(4)
F1 50.73(a)(2)(ii)(B)
E 50.73(a)(2)(viii)(B) 20.2203(a)(2)(i)
El 50.36(c)(1)(i)(A)
L1 50.73(a)(2)(iii)
E] 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL E] 20.2203(a)(2)(ii)
F1 50.36(c)(1)(ii)(A) 0 50.73(a)(2)(iv)(A)
[i 50.73(a)(2)(x)
ED 20.2203(a)(2)(iii)
El 50.36(c)(2) 50.73(a)(2)(v)(A)
[
73.71(a)(4) 10 20.2203(a)(2)(iv)
D1 50.46(a)(3)(ii)
El 50.73(a)(2)(v)(B)
El 73.71(a)(5) 100 E] 20.2203(a)(2)(v)
El 50.73(a)(2)(i)(A)
El 50.73(a)(2)(v)(C)
El OTHER nt 20.2203(a)(2)(vi)
[1 50.73(a)(2)(i)(B)
El 50.73(a)(2)(v)(D)
Specity in Abstract below or in These circuits connect the station to the 345 kV system transmission grid and offsite AC power source.
constitute the MPS STATE OF THE SYSTEM PRIOR TO THE EVENT 5-25-14 AT 07:00:00 DýEnergiz"Out of ServAkm MPS2 PLANT RESPONSE:
On the loss of all offsite AC power, the reactor protection system tripped the reactor in response to the turbine trip as designed. Both safety related emergency diesel generators (EDGs) started automatically based on the under voltage condition, as expected. Both motor driven auxiliary feedwater (AFW) pumps automatically started as expected. All systems responded as expected with the exception of an anomaly in generator voltage regulator performance during the trip. MPS2 remained within the generator capability curve throughout the transient. All safety systems functioned normally. Reactor coolant system (RCS) flow decreased gradually due to the gradual degradation of the electrical supply to the reactor coolant pumps (RCPs). There were no personnel injuries or offsite radiological consequences resulting from this event. A water hammer in the condensate polishing facility was observed during the event. This was entered into MPS corrective action program.
MPS3 PLANT RESPONSE:
On the loss of all offsite AC power, the reactor protection system tripped the reactor in response to the turbine trip as designed. Both safety related EDGs started automatically based on the under voltage condition, as expected. All safety systems responded as expected. The motor driven AFW (MDAFW) pumps automatically started as expected on the reactor trip, and re-sequenced on to the EDG following
the loss of power (LOP). The turbine driven auxiliary feedwater (TDAFW) pump started as expected on the reactor trip.
The MPS3 trip was affected by a loss of air when the 'B' instrument air compressor failed to start. The loss of instrument air affected head vent letdown alignment. After unsuccessful attempts to align the head vent to the volume control tank (VCT), letdown was returned to the pressurizer relief tank (PRT) and the PRT rupture disk burst approximately 20 minutes later.
Normal charging and letdown was subsequently restored which resulted in lifting a relief valve in the letdown line. This relief valve discharged to the primary drains transfer tank (PDTT). As a result, a PDTT high level alarm was received and normal charging and letdown was isolated nine minutes later.
Several areas inside the auxiliary building were contaminated as the result of overfilling the PDTT. As a result, tritium and noble gas were released into the auxiliary building and subsequently released through monitored ventilation pathways. A review of the auxiliary building and secondary leak collection and release system radiation monitors for the time period showed no increase. The contribution to the overall dose to the public was very small, less than 0.001 mrem.
The restoration of plant systems was affected by the loss of offsite AC power and instrument air resulting in delaying the transition from emergency operating procedure E-0, Reactor Trip or Safety Injection, to ES-0.1, Reactor Trip Response, to ES-0.2, Natural Circulation Cooldown. As a result the power operated relief valves (PORVs) operated automatically six times and there were five manual operations of a single PORV.
There were no personnel injuries.
This event is being reported per 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in a manual or automatic actuation of systems listed in 10 CFR 50.73(a)(2)(iv)(B)(1), (6), and (8). Actuations of the reactor protection system, the AFW system, and the EDGs are reportable under this paragraph.
2. CAUSE
With both MPS2 and MPS3 on line at full power and the TO 371 line out of service for scheduled maintenance, a phase to ground fault on one of the phases on the TO 383 line occurred. The fault was caused by an insulator failure on one of the phases of the TO 383 line motor operated disconnect (MOD) switch at a TO substation offsite.
The TO incorrectly set the set-point for the TO 310 line ground instantaneous over-current (IOC) element. The 310 line relay ground IOC element settings were derived without full consideration given to the effects of mutual coupling with adjacent lines. This resulted in a setting that was prone to over-trip for a fault on the TO 383 line. The TO 348 line carried both MPS2 and MPS3 output for about a second then tripped as expected. This resulted in a loss of offsite AC power to both MPS2 and 3.
3. ASSESSMENT OF SAFETY CONSEQUENCES
CONCLUSION:
The response of both MPS2 and MPS3 to the May 2 5th loss of AC power event and dual unit reactor trip in terms of fuel and RCS integrity remain bounded by the events presented in the FSAR Chapter 14 and 15, respectively, accident analysis.
The less severe response of the plant was due to the action of non-safety grade reactor trips, which cannot be credited in the accident analysis, and due to the apparent gradual degradation of the electrical supply to the RCP rather than a sudden loss of motive force at the event onset.
While the recovery phase of the MPS3 transient was complicated with a loss of instrument air and related control issues, none of the observed responses invalidated the bounding nature of the analyses presents in FSAR Chapter 15.
EVALUATION OF MPS2 PLANT RESPONSE:
The MPS2 FSAR does not contain an evaluation of a Loss of Non-emergency AC Power. The Loss of External Load event is presented in Section 14.2.1 of the MPS2 FSAR. The Loss of Forced Reactor Coolant Flow is presented in Section 14.3.1. The May 25, 2014 event combined features of both of these events.
The major difference between those FSAR Chapter 14 accident analysis events and the event of May 25, 2014 is that the accident analysis allows the event to proceed until the generation of the first safety grade reactor protection system trip occurs. In the case of the Section 14.2.1 event, a reactor trip signal is not credited until the generation of a high pressurizer pressure trip at 4.9 seconds with CEA insertion beginning at 6.3 seconds. In the case of the Section 14.3.1 event, a reactor trip is not credited until the generation of a low flow trip at 1.29 seconds with CEA motion beginning at 2.44 seconds. In the May 25, 2014 event, the non-safety grade reactor trip on turbine trip was first out.
The RCS flow exhibited an unusual behavior for a freewheeling coast-down of the RCPs. A typical flow coast-down usually begins with a very rapid rate for flow reduction, the rate of reduction gradually slowing as the pumps reach lower speed. The flow rate for MPS2 experienced a nearly linear rate of reduction for about a minute before taking on a more typical loss of flow behavior. This linear behavior made the transient approach to a DNBR safety limit much less severe than those analyzed in the Chapter 14 safety analysis for MPS2.
After the RCP coast-down, the plant established natural circulation as indicated by the approximately 25 degrees F hot to cold leg differential temperature.
The pressurizer pressure remained below its 'pre-event' value for the duration of the event. Therefore there was no need for the power operated relief valves or pressurizer safety valves to act in response to the event.
Pressurizer level reached a minimum of 27.6% with the initial RCS shrinkage upon reactor trip, before recovering due to both charging and the establishment of the differential temperatures for natural circulation, The maximum steam generator pressure reached during the event was 931 psia. This pressure is well below the set-points for the lowest setting (1000 psia +/-3 %) main steam safety valves (MSSVs).
A review of plant process computer points confirmed that no MSSVs lifted during the event.
Steam generator reached a minimum of 17.2% narrow range before recovering with AFW delivered from 2 motor driven auxiliary feedwater pumps.
EVALUATION OF MPS3 PLANT RESPONSE:
The MPS3 FSAR contains an evaluation of a Loss of Nonemergency AC Power in Section 15.2.6 of the accident analysis, although it does defer to the Loss of Flow event, Section 15.3.2, for limiting departure from nucleate boiling ratio (DNBR) and to the Turbine Trip, Section 15.2.3 for limiting primary and secondary pressures. Note that the Chapter 15 accident analyses of these events are performed primarily to quantify the peak over-pressurization of the 15.2.3 event or the DNBR crisis of the 15.3.2 event. For these two events, this quantification is less than 30 seconds in length. Beyond this time, the plant is in recovery following emergency operating procedures (EOPs). While there may be malfunctions that complicate the recovery period, unless they result in additional events (i.e., PORVs
which fail to reseat or releases of contaminated steam) they typically do not factor into a judgment as to whether a plant transient remains bounded by the Chapter 15 accident analysis.
The major difference between those Chapter 15 accident analysis events and the May 25, 2014 event is that the accident analysis allows the event to proceed until the generation of the first safety grade reactor protection system trip occurs. In FSAR Table 15.2-1, reactor trip is not credited until the generation of a high pressurizer pressure trip at 6.2 seconds with CEA insertion beginning at 8.2 seconds. In the case of the loss of flow event, Table 15.3-1 shows a reactor trip is not credited until the generation of a low flow trip at 0.9 seconds with CEA motion beginning at 1.5 seconds. In the May 25, 2014 event, the non-safety grade reactor trip on turbine trip was first out.
The RCS flow exhibited an unusual behavior for a freewheeling coast-down of the RCPs. A typical flow coast-down usually beings with a very rapid rate for flow reduction, the rate of reduction gradually slowing as the pumps reach lower speed. The flow rate for MPS3 experienced a nearly linear rate of reduction for about a minute before taking on a more typical loss of flow behavior. This linear behavior made the transient approach to a DNBR safety limit much less severe than those analyzed in the Chapter 15 safety analysis for MPS3.
As expected, after the RCP coast-down, natural circulation was established as indicated by the approximately 35 degrees F hot to cold leg differential temperature.
Both EDGs auto started as expected and energized 4160V busses 34C and 34D.
The pressurizer pressure remained below its 'pre-event' value for the period of the initial potential over-pressurization at the time of the Loss of AC. Therefore there was no need for the PORVs or the pressurizer safety valves (PSVs) to act in response to the initial over-pressurization event. Later, during event recovery without instrument air, pressurizer pressure did increase to the point of automatically cycling the PORVs six times.
The maximum steam generator pressure reached during the event was 1133 psig. This pressure is below the set-point for the lowest setting (1185 psig +/-3 %) MSSVs.
Steam generator level reached a minimum of 44.2% wide range before recovering with AFW delivered from two MDAFW and the TDAFW pumps.
4. CORRECTIVE ACTION
The faulted TO 383 line MOD switch at the TO Card Substation has been replaced by the TO. The ground IOC elements for all remote line terminals associated with MPS have been temporarily disabled by the TO through relay setting changes. Additional corrective actions are being taken in accordance with the station's corrective action program.
5. PREVIOUS OCCURRENCES
None
- 6. Energy Industry Identification System (EllS) codes Emergency AC Power System EK AC Power Distribution System EA, EB, EC, ED, EF, Auxiliary Feedwater System BA Reactor Coolant System AC Main Steam System SB Instrument Air Supply System LD Plant (reactor) Protection System JC Motor Operated Disconnect Switch SWGR Turbine TRB Pump P
Air Compressor CMP Main Generator TB Voltage Regulator RG Relay RLY Over-current Element SWGR Safety Valve RV Pressurizer PZR