05000458/LER-2015-005, Regarding Automatic Reactor Scram Due to Low Reactor Water Level Following a Loss of Instrument Power
| ML15231A180 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 07/29/2015 |
| From: | Olson E Entergy Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RBG-47598 LER 15-005-00 | |
| Download: ML15231A180 (5) | |
| 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) 10 CFR 50.73(a)(2)(ix) 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)(A), System Actuation |
| 4582015005R00 - NRC Website | |
text
Entergy Operations, Inc.
~River Bend Station 5485 U. S. Highway 61 N Enter TeI~~~~St.
Francisville, LA 70775Fx2553184734 eolson@entergy.com
- Eric W, Olson Site Vice President.
RBG-47598 July 29, 2015 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
Subject:
Licensee Event Report 50-458 / 2015-005-00 River Bend Station - Unit 1 Docket No. 50-458 License No. NPF-47 RBF1 -1 5-0121
Dear Sir or Madam:
In accordance with 10.CFR 50.73, enclosed is the subject Licensee Event Report.
This document contains no commitments. lfyou have any questions, please contact Mr. Joseph Clark at 225-381-4177.
Sincerely, EWO/d hw Enclosure cc:
U. S. Nuclear Regulatory Commission Region IV 1600 East Lamar Blvd.
Arlington, TX 76011-4511 NRC Sr. Resident Inspector P. O. Box 1050 St. Francisville, LA 70775
Licensee Event Report 50-458 I 201 5-005-00 July 29, 2015 RBG-47598 Page 2 of 2 INPO (via ICES reporting)
Central Records Clerk Public Utility Commission of Texas 1701 N. Congress Ave.
Austin, TX 78711-3326 Department of Environmental.Quality Office of Environmental Compliance Radiological Emergency Planning and Response Section Ji Young Wiley P.O. Box 4312 Baton Rouge, LA 70821-4312
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 0113112017 (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 />.
,Z,,,Reported lessons learned ame incorporated into the licensing process and fed back to industry.
o20 tections
.. *Brandchmet
- rT5F3, Negard Regbuldnesiator CommissiOnA, Wasingto n, DIn2555-001,or byetin internet e-mail to Infocolleots.Resourceo~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. Ifsa means sused to impose an information collection does nut display a currently valid 0MB digits/characters for each block) contol number, the NRC may not conduct or sponsor, sod a person is not required to respond to, the information collection.
- 1. FACILITY NAME 2
OKTNME
.PG River Bend Station - Unit 1 000481O
- 4. TITLE Automatic Reactor Scram Due to Low Reactor Water Level Following a Loss of Instrument Power
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MOT A
ER YA EUNIL RVFACILITY NAME DOCKET NUMBER MOYANYA EUMENILR EVO MONTH IDAY YEAR 05000 21 05 0
07 2
205 FACILITY NAME DOCKET NUMBER 2015 2015
____-_005 00 07 2
2015_
05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
[ 2.20(b)
[]
20.2203(a)(3)(i)
El 50.73(a)(2)(i)(C)
El 50.73(a)(2)(vii) 1 E 20.2201(d)
El 20.2203(a)(3)(ii)
El 50.73(a)(2)(ii)(A)
[]
50.73(a)(2)(viii)(A)
El 20.2203(a)(1)
[] 20.2203(a)(4)
[] 50.73(a)(2)(ii)(B)
[j] 50.73(a)(2)(viii)(B) 1.PWRLVL
[] 20.2203(a)(2)(ii)
[]
50.36(c)(1)(ii)(A)
[]
50.73(a)(2)(iii()
[]
50.73(a)(2)(ix)(A El 20.2203(a)(2)(iii)
[]
50.36(c)(2)
[]
50.73(a)(2)(v)(A)
[]
73.71 (a)(4) 90 El 20.2203(a)(2)(iv)
[]
50.46(a)(3)(ii)
[]
50.73(a)(2)(v)(B)
El 73.71 (a)(5)
El 20.2203(a)(2)(v)
El 50.73(a)(2)(i)(A)
El 50.73(a)(2)(v)(C)
El OTHER E
20.2203(a)(2)(vi)
El 50.73(a)(2)(i)(B)
El 50.73(a)(2)(v)(D)
Specify in Abstract below or in '
- 12. LICESEE CONACT FORTHIS6LE LICENSEEEECONACCONTACT LE Joseph A. Clark, Manager - Regulatory Assurance (2)3147CAS YTM ICMOET MANU-REPORTABLE I*
M AS YTANU-REPORTABLE
- 14.
S SYTE/aMONN FTUE TOE IX CAS SYST NEM OPNNT 1
FACTURER TO EPIX 4,SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR
[] YES (If yes, complete 15. EXPECTED SUBMISSION DATE)
[]
NO DATESIO ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On June 1, 2015, at 9:09 p.m. CDT, with the plant operating at 90 percent power, an unplanned automatic reactor scram occurred due to low reactor water level. This event resulted from the loss of a non-safety related instrument power panel, apparently caused by an internal electrical transient in a 125-volt AC / DC inverter. This event is being reported in accordance with 10 CFR 50.73(a)(2)(iv)(A) as the automatic actuation of the reactor protection system. All reactor control rods inserted normally, and control of reactor parameters was promptly established using the main turbine bypass valves and the main feedwater system. An expected general containment isolation signal oecurred when reactor water level decreased to Level 3. The "A" reactor recirculation pump shifted to slow speed as designed, while the "B" pump tripped off. The runback feature of the reactor recirculation flow control valves failed to operate due the loss of instrument power. No plant parameters requiring the actuation of the emergency diesel generators, the main steam safety-relief valves, or the emergency core cooling systems were exceeded. This event was, thus, of minimal safety significance to the health and safety of the public.
NRC FORM 366 (02-201[4)
REPORTED CONDITION On June 1, 2015, at 9:09 p.m. CDT, with the plant operating at 90 percent power, an unplanned automatic reactor scram occurred due to low reactor water level. All reactor control rods inserted normally, and control of reactor parameters was promptly established using the main turbine bypass valves and the main feedwater system. No reactor main steam relief valves actuated, and no emergency core cooling systems were required to initiate. Operators entered the emergency operating procedures for reactor pressure vessel control (for the low water level condition), primary containment control (for high containment atmospheric pressure), and secondary containment control (due to abnormally high drain sump water levels). An expected general containment isolation signal occurred when reactor water level decreased to Level 3. The "A" reactor recirculation pump shifted to slow speed as designed, while the "B" pump tripped off. The runback feature of the reactor recirculation flow control valves failed to operate.
Troubleshooting by the operators determined that this event was initiated by the loss of 24-volt DC instrumentation power (EE). This event is being reported in accordance with 10 CFR 50.73(a)(2)(iv)(A) as the automatic actuation of the reactor protection system.
INVESTIGATION AND IMMEDIATE CORRECTIVE ACTIONS The initial troubleshooting during the scram recovery determined that a power failure had occurred on a balance-of-plant (non-safety related) instrumentation panel. The panel is supplied with 120-volt AC through a disconnect switch with a 30-amp fuse. The 120-volt power feeds four parallel 24-volt DC power supplies. On the output of each power supply is a power indication status light and a power supply failure alarm relay. The loss of power caused the following malfunctions in plant systems:
- 1. The main reactor feedwater pump minimum flow valves and the heater drain pump recirculation valves all failed open. This had the effect of diverting a significant portion of feedwater system flow directly back to the main condenser, causing reactor water level to decrease. As the feedwater flow control valves opened in response to the low reactor water level, feedwater system pressure decreased to the point where the low suction pressure switches for the feedwater pumps tripped. The trip of the "A" and "C" feedwater pumps was a normal response to this condition. The "B" pump should have tripped, but a relay failure in the circuitry caused it to continue to operate.
- 2. The turbine building chillers, the normal source of cooling for the primary containment, shut down due to the loss of instrumentation power.
- 3. The reactor water cleanup system shutdown due to the high area temperatures following the loss of cooling fr'om the turbine building chillers.
- 4. The runback feature of the reactor recirculation flow control valves failed to function due to the loss of power to main feedwater flow instruments.
Electricians performed detailed troubleshooting, and it was found that the 120-volt disconnect switch was closed and that power was available downstream of the 30-amp fuse. The four input fuses to the 24-volt power supplies were all found to be blown, while no other fuses in the panel were affected. Prior to the event, the 120-volt panel had been aligned to the normal uninterruptible power supply (UPS). No work activities were being performed in or on the affected control room panel, the UPS, or the 120-volt panel prior to the event. No abnormal indications were present on the UPS panel prior to or following the event.
NRC FORM 3e6A (02-2014)
The investigation team concluded that the most probable cause of this event was a power transient, created by the failure of a capacitor in the output circuitry of the UPS. The loss of instrument power resulted from the failure of the input fuses on all four 24-volt power supplies. The cause of the power supply input fuse failures was not conclusively identified. However,.failure*-analysis~determined the fuses failed due to one or a combination of the following conditions:
- A failed capacitor in the output of the UPS caused a transient that exposed the power supply to a large inrush current, which exceeded the rating of the fuses.
- One or more blown input fuses on the power supplies, coupled with low margin in the power supply fuse design and load imbalance.
Testing was not able to create a cascading failure by removing individual power supplies from service at normal loading conditions.
This is a probable cause but could not be proven or disproven.
CORRECTIVE ACTIONS to PREVENT RECURRENCE Based on the UPS vendor recommendation, the 6-amp fast-blow fuses on the input side of the 24-volt power supplies were replaced with 10-amP slow-blow fuses. Following an upcoming UPS maintenance outage in September, all the replaced capacitors will be tested for obvious signs of failure or degradation. Any suspect capacitors will be sent offsite for failure analysis.
PRIOR OCCURRENCE EVALUATION No similar events have been reported by River Bend Station in the previous three years.
SAFETY SIGNIFICANCE
Aside from the specific abnormalities described above, the overall response of the plant to this actuation of the reactor protection system was as expected. No plant parameters requiring the actuation of the emergency diesel generators or the emergency core cooling systems were exceeded. This event was, thus, of minimal safety significance to the health and safety of the public.