05000461/LER-2011-008, Reactor Protection System Actuation and Loss of Shutdown Cooling
| ML12046A645 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 02/06/2012 |
| From: | Noll W Exelon Nuclear |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| U-604053 LER 11-008-00 | |
| Download: ML12046A645 (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 |
| 4612011008R00 - NRC Website | |
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Nuclear Clinton Power Station 8401 Power Road Clinton, IL 61727 U-604053 February 6, 2012 10 CFR 50.73 SRRS 5A.108 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461
Subject:
Licensee Event Report 2011-008-00 Enclosed is Licensee Event Report (LER) No. 2011-008-00: Reactor Protection System Actuation And Loss Of Shutdown Cooling. This report is being submitted in accordance with the requirements of 10 CFR 50.73.
There are no regulatory commitments contained in this report.
Should you have any questions concerning this report, please contact J. E. Cunningham, at (217) 937-2200.
Rest William G. Noll Site Vice President Clinton Power Station RSF/blf
Enclosure:
Licensee Event Report 2011-008-00 cc:
Regional Administrator - NRC Region III NRC Senior Resident Inspector - Clinton Power Station Office of Nuclear Facility Safety - IEMA Division of Nuclear Safety kAUZ4(l
4RC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2013
, the NRC may sfor each block) not conduct or sponsor, and a person is not required to respond to, the digits/characters finformation collection.
- 3. PAGE Clinton Power Station, Unit 1 05000461 1 OF 4
- 4. TITLE Reactor Protection System Actuation And Loss Of Shutdown Cooling
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE 1
- 8. OTHER FACILITIES INVOLVED SEQUENTIAL REV I
FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NO.
MONTH DAY YEAR N/A N/A NUBR N.
_I I
I_
j I
IFACILITY NAME DOCKET NUMBER 12 18 2011 2011
- - 008 00 02 06 20121 N/A N/A
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMIT7ED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
El 20.2201(b)
El 20.2203(a)(3)(i)
El 50.73(a)(2)(i)(C)
El 50.73(a)(2)(vii) 4l 20.2201(d)
El 20.2203(a)(3)(ii)
El 50.73(a)(2)(ii)(A)
El 50.73(a)(2)(viii)(A)
[I 20.2203(a)(1)
[I 20.2203(a)(4)
El 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
E_ 20.2203(a)(2)(i)
El 50.36(c)(1)(i)(A)
[I 50.73(a)(2)(iii)
[I 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL 0l 20.2203(a)(2)(ii)
[I 50.36(c)(1)(ii)(A) 0 50.73(a)(2)(iv)(A)
El 50.73(a)(2)(x)
E-20.2203(a)(2)(iii)
El 50.36(c)(2)
El 50.73(a)(2)(v)(A)
El 73.71(a)(4) 000 El 20.2203(a)(2)(iv)
El 50.46(a)(3)(ii) 0 50.73(a)(2)(v)(B)
El 73.71(a)(5)
El 20.2203(a)(2)(v)
El 50.73(a)(2)(i)(A)
[E 50.73(a)(2)(v)(C)
[E OTHER El 20.2203(a)(2)(vi)
El 50.73(a)(2)(i)(B)
[E 50.73(a)(2)(v)(D)
Specify in Abstract below or in commenced injecting approximately 200 gallons of water per minute into the RPV. Operators reset the RHR system isolation signal, allowing the operators to reopen the Shutdown Cooling containment isolation valves.
Approximately five minutes after the initial actuation, Operations directed an extra Reactor Operator in the control room to obtain readings from the Analog Trip Modules for the transmitters which are used in the Level 3 actuation and containment isolation logic. At that time, the readings for all four divisions of RPV water level indication were nominally +24 inches and trending up. After the Level 3 actuation, Operations ordered a rising trend on Shutdown Range and shortly thereafter, provided a level band of +130 to +150 inches Shutdown Range. At about 0956, operators restored of RHR SDC Train A. During the period of time shutdown cooling was not in service, reactor coolant temperature increased approximately three degrees Fahrenheit.
RR Pump A remained in operation in low speed throughout this event so forced cooling of the reactor core was not lost. RWCU System Train A also remained in operation throughout the event.
Following recovery of shutdown cooling, the Operations crew recognized that the vessel indicated level was higher than vessel actual level due to an incompletely filled reference leg for the shutdown and upset level instruments reference leg pipe. Maintenance was contacted to re-perform the fill of the shutdown and upset level instruments reference leg pipe. After the reference leg pipe fill was completed at 1253 hours0.0145 days <br />0.348 hours <br />0.00207 weeks <br />4.767665e-4 months <br />, RPV shutdown level indication lowered 109 inches from +195 inches to +86 inches as indicated on the shutdown instrument. This was determined to be the amount of level error due to the initial incompletely filled shutdown and reference instruments reference leg pipe.
There were no structures, systems, or components that were inoperable at the start of the event that contributed to this event.
This event was an 8-hour reportable event under 10 CFR 50.72 (b)(3)(iv)(A) as a valid actuation of the Reactor Protection System [JC] and 10 CFR 50.72(b)(3)(v)(B) as an event that at the time of discovery could have prevented the fulfillment of a safety function needed to remove residual heat. (Completed Event Notification 47533)
This event is also reportable under the provisions of 10 CFR 50.73 (a)(2)(iv)(A) due to a valid actuation of the Reactor Protection System [JC] and 10 CFR 50.73(a)(2)(v)(B) as an event that could have prevented the fulfillment of the safety function of structures or systems that are needed to remove residual heat. Issue Report 1304323 was initiated to investigate this event.
C. CAUSE OF EVENT
The cause of this event was the lack of rigorous process controls while removing and installing the permanent shutdown and upset level instruments reference leg pipe which resulted in level indication indicating high on the shutdown level instrument being used to monitor water level. Specifically, instructions on how to fill the shutdown and upset level instruments reference leg pipe were inadequate and there was insufficient guidance on how to perform a check of the restored instrument.
D. SAFETY CONSEQUENCES
The risk associated with losing SDC in Mode 4 is the loss of a key system used to remove decay heat from the RPV while shutdown. While there are other methods of decay heat removal, the primary method is the SDC mode of RHR, which is a significant contributor to risk reduction. With the loss of SDC, the temperature of the
RPV coolant will begin to rise and there is a potential for an unplanned entry into Mode 3 (Hot Shutdown) condition.
All control rods were fully inserted prior to the event thus the generation of the Low RPV Water Level 3 signal had no impact on plant status. The SDC system was restored immediately after the event as reactor water level was restored above the Level 3 setpoint using the CRD system. The operators reset the logic within minutes of the scram signal, and cooling was fully restored in 26 minutes. Reactor coolant temperature increased approximately three degrees F.
Since RPV level was immediately restored above the Low RPV Water Level 3 scram setpoint following the scram; the RHR isolation signal was available to be reset; thus, both RHR SDC loops were always available for Shutdown Cooling. No change to plant risk was incurred by this isolation. The RHR system was available for SDC throughout the event because operators always had the means to restore level, reset the isolation, and restore SDC prior to any negative consequences occurring, such as a mode change.
E. CORRECTIVE ACTIONS
Procedure CPS 3007.01, Preparation and Recovery from Refueling Operation, is being revised to control the entire evolution of shutdown and upset level instruments reference leg pipe reassembly and recovery of vessel level indication, to include overall evolution monitoring from the MCR, the instrument maintenance steps how to fill and vent the shutdown and upset level instruments reference leg pipe, including the required volume needed to completely fill the shutdown and upset level instruments reference leg pipe, preferred method for injecting water and precautions associated with use of the equalizing valve.
An alternate method will be developed for determining RPV level during shutdown conditions or maintaining shutdown instrumentation available during RPV disassembly/ re-assembly.
F. PREVIOUS OCCURRENCES
The station issued LER 2004-004, Reactor Scram While Placing Residual Heat Removal B into Shutdown Cooling, involving a Level 3 isolation and loss of SDC in July 2004. The 2004 event and this December 2011 event both had issues with communications, the 2004 issues led to inadequate resolution of conflicting indication prior to proceeding, but in the December 2011 event, Operations did not proceed until they were given assurances that the RPV water level indication was technically validated and accurate after the second fill and vent of the reference leg.
G. COMPONENT FAILURE DATA
None