05000416/LER-2008-004, Automatic Reactor Scram on Decreasing Coolant Level Due to Inadvertent Reactor Feed Pump Steam Supply Valve Closure
| ML083510838 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 12/15/2008 |
| From: | Larson M Entergy Corp, Entergy Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| GNRO-2008/00076 LER 08-004-00 | |
| Download: ML083510838 (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)(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(b)(5) |
| 4162008004R00 - NRC Website | |
text
-===-Entergy GNRO-2008/00076 December 15, 2008 Entergy Operations, Inc.
PO Box 756 Port Gibson, MS 39150 Tel 601 4372800 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001
Subject:
LER 2008-004 Automatic Reactor SCRAM On Decreasing Coolant Level Due to Inadvertent Reactor Feed Pump Steam Supply Valve Closure Grand Gulf Nuclear Station, Unit 1 Docket No. 50-416 License No. NPF-29
'Dear Sir or Madam:
Attached is Licensee Event Report (LER) 2008-004-00 which is a final report.
This letter does not contain any commitments.
Yours truly, Michael J. Larson Acting-Manager, Licensing MJUJEO:jeo
attachment:
cc:
LER 2008-004-00 (See Next Page)
GNRO-2008/00076 Page 2 cc:
NRC Senior Resident Inspector Grand Gulf Nuclear Station Port Gibson, MS 39150 U. S. Nuclear Regulatory Commission ATTN: Mr. Elmo E. Collins (w/a)
Regional Administrator, Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-4005 U. S. Nuclear Regulatory Commission ATTN: Mr. Carl F Lyon, NRRlADRO/DORL (w/2)
ATTN: ADDRESSEE ONLY ATTN: U. S. Postal Delivery Address Only Mail Stop OWFN/8 B1 Washington, DC 20555-0001
GNRO-2008/00076 NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 08/31/2010 9-2007)
, the NRC may digits/characters for each block) not conduct or sponsor, and a person is not required to respond to, the information collection.
r. PAGE Grand Gulf Nuclear Station, Unit 1 05000416 1 OF 3
- 14. TITLE Automatic Reactor SCRAM On Decreasing Coolant LeVel Due To Inadvertent Reactor Feed Pump Steam Suoolv Valve Closure
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED SEQUENTIAL REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NO.
MONTH DAY YEAR N/A N/A FACILITY NAME DOCKET NUMBER 10 23 2008 2008
- - 004 - 00 12 15 2008 N/A N/A
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check a/l 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) 010 o 20.2203(a)(2)(iv) o 50A6(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 o 20.2203(a)(2)(vi) o 50.73(a)(2)(i)(B) o 50.73(a)(2)(v)(D)
Specify in Abstract below or in NRC Form 366A
- 12. LICENSEE CONTACT FOR THIS LER FACILITY NAME I~ElEPHONE NUMBER (Include Area Code)
Grand Gulf Nuclear Station - Michael J. Larson. Actino Licensino Manaoer 601-437-6685CAUSE SYSTEM COMPONENT MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE FACTURER TO EPIX FACTURER TO EPIX
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR SUBMISSION o YES (If yes, complete 15. EXPECTED SUBMISSION DATE)
~NO DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On October 23,2008 at 0721 Grand Gulf Nuclear Station was operating in MODE 1 at approximately 10 percent power when an actuation of the Reactor Protection System (RPS) occurred due to decreasing reactor coolant level. The decreasing reactor coolant level was the result of closure of the "A" Reactor Feed Pump Turbine (RFPT) steam inlet valves. During restoration of the "8" RFPT Overspeed Trip Test a non-licensed operator (NLO) closed the steam supply drain valves on the "A" RFPT train instead of the "8" train. Closure of the drain valves did not cause loss of feedwater to the reactor vessel. Upon discovery of the mistake the NLO tried to correct the error and closed the "A" RFPT steam inlet valves.
This led to the loss of feedwater flow to the reactor vessel.
A reactor SCRAM was received per design at Level 3 (+11.4 inches) and all control rods fully inserted.
Operators manually initiated the Reactor Core Isolation Cooling system to restore and maintain reactor vessel level. No emergency core cooling system initiation setpoint was reached and no safety relief valves lifted. The normal heat sink (main condenser) remained available during this event.
The root cause of this event was determined to be a lack of commitment to Human Performance Program implementation.
NRC FORM 366 (9-2007)U.S. NUCLEAR REGULATORY COMMISSION (9-2007)
LICENSEE EVENT REPORT (LER)
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE I SEQUENTIAL I REV YEAR NUMBER NO.
Grand Gulf Nuclear Station Unit 1 05000416 2008 004 --
00 2
of 3
A. REPORTABLE OCCURRENCE
.-On October 23,20.08 at 0721 Grand Gulf Nuclear Station experienced and'automatic reactor SCRAM due to decreasing reactor coolant level. Upon decreasing to Level 3 (+11.4 inches) a Reactor Protection System (RPS) [JC] SCRAM signal was auto initiated. The initiation of RPS resulting in an automatic reactor SCRAM is reportable pursuant to 10 CFR 50,73(a)(2)(iv)(A),
B.
INITIAL CONDITIONS At the time of the event, the reactor was in OPERATIONAL MODE 1 with reactor power at approximately 10 percent. There were no additional inoperable structures, systems, or components at the start of the event that contributed to the event.
C.
DESCRIPTION OF OCCURRENCE On October 23,2008 at 0721 Grand Gulf Nuclear Station experienced an automatic reactor SCRAM due to decreasing reactor coolant level caused by loss of "A" RFPT train. At the initiation of the SCRAM the unit was operating in Mode 1 at approximately 10 percent power. Prior to the SCRAM Operations had performed a Reactor Feed Pump Turbine (RFPT) [SJ] "B" Overspeed Trip Test with the RFPT uncoupled from the pump, The system was ready for restoration therefore the control room reactor operator (RO) dispatched a non licensed operator (NLO) into the plant to restore the system. The NLO proceeded to manipulate controls on what was thought to be the "B" RFPT train however the NLO manipulated the steam supply drains on the operating "A" RFPT train. These steps did not initiate the loss of the "A" RFPT. As the NLO proceeded to the next step it was noticed that there was no light indication for the next valve to be manipulated. Having performed the valve line up on the "A" RFPT train the day before the NLO now realized it was the "A" train rather than the "B" train due to the fact that the NLO remembered the lack of light indication. At the same time the NLO realized the mistake the main control room called the NLO on the plant public address system.
The NLO concluded that the page concerned the mistake and attempted to correct the error. In an attempt to undo the error the NLO closed the High Pressure Steam Supply Valves to the "A" RFPT. As a result of the isolation of the steam inlet valves the "A" RFPT lost steam pressure. Shortly thereafter the Reactor Feed Pump discharge pressure dropped below reactor pressure, which resulted in a loss of feedwater flow to the reactor vessel. This resulted in a reactor SCRAM upon reaching Level 3 (+11.4 inches).
The reactor SCRAM came in as designed at Level 3 (+11.4 inches). Operators manually initiated Reactor Core Isolation Cooling (RCIC) to restore and maintain reactor vessel level. The lowest indicated reactor vessel level was -39 inches Wide Range. All withdrawn control rods fully inserted and no emergency core cooling system (ECCS) initiations were received. No Safety Relief Valves lifted as a result of this event and all other equipment operated as expected.U.S. NUCLEAR REGULATORY COMMISSION (9-2007)
LICENSEE EVENT REPORT (LER)
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE I SEQUENTIAL I REV YEAR NUMBER NO.
Grand Gulf Nuclear Station Unit 1 05000416 2008 004 --
00 3
of 3
D.
CAUSE~OCCURRENCE The ca~Jse of the event was that the NLO did not use the self checking standard of TOUCH-READ-READ which involves touching the component intended to be manipulated, reading the tag, and reading the procedural step to verify the correct component was about to be manipulated. The NLO also did not use the required circle and slash method of place keeping.
The root cause of this event was determined to be a lack of commitment to Human Performance Program implementation.
E.
CORRECTIVE ACTIONS
Immediate Corrective Actions - Reactor water level was restored and the plant placed in a stable condition.
The individual involved was removed from shift and a human performance error review was conducted.
Additionally, temporary covers were placed over the switches on the 1H22-P175 Panel for the RFPT "A" and "B" controls.
Long Term Corrective Actions - Condition Report CR-GGN-2008-06195 was written and will address any additional actions.
F.
SAFETY ASSESSMENT
Immediate actions performed by the Operations staff were adequate and appropriate in placing and maintaining the reactor in a safe shutdown condition. The lowest reactor level indicated was -39 inches Wide Range. This is above the initiation setpoint for ECCS systems. RCIC was manually initiated to restore and maintain reactor vessel level.
The Group 2 Residual Heat Removal (RHR) to Radwaste and Group 3 RHR Shutdown Cooling automatic isolations were received however no valves isolated because they were in their normally closed position prior to the event. No Safety Relief Valves lifted as a result of this event and all other systems performed as required.
This event did not prevent the fulfillment of a safety function therefore there were no safety system functional failures. The health and safety of the public was not compromised by this event.
G.
ADDITIONAL INFORMATION
Previous Similar Events - Pursuant to 10CFR50.73(b)(5) this issue is considered an infrequent event. There has not been any occurrence of the same underlying concern in the past two years at Grand Gulf Nuclear Station.