05000255/LER-2011-008, Regarding Reactor Protection System and Auxiliary Feedwater System Actuation
| ML120440664 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 02/13/2012 |
| From: | Vitale A Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| PNP 2011-012 LER 11-008-00 | |
| Download: ML120440664 (5) | |
| 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) 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 |
| 2552011008R00 - NRC Website | |
text
m Entergy Nuclear Operations, Inc.
Palisades Nuclear Plant
~Entergy
~ Highway Tel 269 764 2000 Anthony J. Vitale Site Vice President PNP 2011-012 February 13, 2012 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001
SUBJECT:
Reactor Protection System and Auxiliary Feedwater System Actuation Palisades Nuclear Plant Docket 50-255 License No. DPR-20
Dear Sir or Madam:
Licensee Event Report (LER) 2011-008 is enclosed. The LER describes a manual actuation of the reactor protection system and an automatic actuation of the auxiliary feedwater system. The occurrence is reportable in accordance with 10 CFR 50.73(a)(2)(iv)(A).
This letter contains no new commitments and no revisions to existing commitments.
Sincerely, ajv/tad
Attachment:
LER 2011-008, Reactor Protection System and Auxiliary Feedwater System Actuation CC Administrator, Region Ill, USNRC Project Manager, Palisades, USNRC Resident Inspector, Palisades, USN RC
ATTACHMENT LER 2011-008 REACTOR PROTECTION SYSTEM AND AUXILIARY FEEDWATER SYSTEM ACTUATION 3 Pages Follow
NRC FORM 366 APPROVED BY 0MB NO. 3150-0104 EXPIRES 10/31/2013 U.S. NUCLEAR REGULATORY COMMISSION 010)
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
(See reverse for required number of digits/characters tor each block)
- 3. PAGE PALISADES NUCLEAR PLANT 05000255 1
OF 3
- 4. TITLE Reactor Protection System and Auxiliary Feedwater System Actuation
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED SEQUENTIAL REV FACILITY NAME MONTH DAY YEAR YEAR NUMBER NO MONTH DAY YEAR FACILITY NAME 12 14 2011 2011 008
- - 00 02 13 2012
- 9. OPERATING ODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply
~
20.2201(b) 0 20.2203(a)(3)(i) 0 50.73(a)(2)(i)(C) 0 50.73(a)(2)(vii) 1 0
20.2201(d) 0 20.2203(a)(3)(ii) 0 50.73(a)(2)(ii)(A) 0 50.73(a)(2)(Viii)(A)
LJ 20.2203(a)(1) 0 20.2203(a)(4) 0 50.73(a)(2)(ii)(B) 0 50.73(a)(2)(viii)(B) 0 20.2203(a)(2)(i) 0 50.36(c)(1)(i)(A) 0 50.73(a)(2)(iii) 0 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL Q
20.2203(a)(2)(ii)
Q 50.36(c)(1)(ii)(A)
~
5073(a)(2)(iv)(A)
~ 50.73(a)(2)(x) 0 20.2203(a)(2)(iii) 0 50.36(c)(2) 0 50.73(a)(2)(v)(A)
Q 73.71 (a)(4)
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0 20.2203(a)(2)(iv) 0 5046(a)(3)(ii) 0 50.73(a)(2)(v)(B) 0 73.71(a)(5)
I VU 0
20.2203(a)(2)(v) 0 50.73(a)(2)(i)(A) 0 50.73(a)(2)(v)(C) 0 OTHER Specify in Abstract below or fl
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20.2203(a)(2)(vi) 0 50.73(a)(2)(i)(B) 0 5073(a)(2)(v)(D)12. LICENSEE CONTACT FOR THIS LER FA ITY NAME TELEPHONE NUMBER (Include Area Code Terry Davis (269) 764-2117AUSE SYSTEM COMPON NT FACTURER REPORTABLE
CAUSE
SYSTEM OMPONENT FACTURER REPORTABLE D
SJ FUB B569 Y
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPE~.,TED MONTH DAY YEAR SUBMISSION Q
YES (If yes, complete 15. EXPECTED SUBMISSION DATE)
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NO DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines On December 14, 2011, main feedwater (MEW) pumps, P-i A and P-i B, tripped due the loss of suction pressure. Operators entered the loss of MEW procedure and appropriately carried out the procedurally directed action to manually trip the reactor. As expected, the auxiliary feedwater system started automatically to recover level in the steam generators.
The immediate cause for the loss of suction pressure to the MEW pumps was the spurious opening of MEW pump, P-iA, recirculation control valve, CV-071 1. The spurious opening of CV-071 1 resulted from a loss of electrical control power to the valve. The loss of electrical control power to CV-071 1 was due to insufficient contact between the fuse holder clip and a fuse ferrule within a fuse block assembly of the control power circuitry. The possible causes for the lack of sufficient contact between the fuse holder clip and the fuse ferrule were the stretching of the fuse holder clip during the installation/removal of a fuse tag-out device, and/or fatigue of the fuse holder clip, induced by the repetitive removal/installation of the fuse.
The control power circuitry, for CV-071 1, was rerouted through an acceptable spare fuse block assembly and reenergized. The event is reportable in accordance with 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in a manual actuation of the reactor protection system and the automatic actuation auxiliary feedwater system.
NRC FORM 366 (10-2010)U.S. NUCLEAR REGULATORY COMMISSION (1O~2O1O)
LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE SEQUENTIAL REVISION YEAR NUMBER NUMBER PALISADES NUCLEAR PLANT 05000255 2011 008 00 2
OF 3
EVENT DESCRIPTION
On December 14, 2011, with the plant in Mode 1 at 100% power, main feedwater (MEW) pumps, P-iA and P-lB [P;SJ], tripped due the loss of suction pressure. Operators entered the loss of MEW procedure and appropriately carried out the procedurally directed action for a loss of both MEW pumps, which is to manually trip the reactor [RCT;AB]. As expected, the auxiliary feedwater system [BA] started automatically to recover level in the steam generators [SG;SB].
The event is reportable in accordance with 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in a manual actuation of the reactor protection system [JC] and the automatic actuation auxiliary feedwater system.
CAUSE OF THE EVENT
For pump protection, both MEW pumps have minimum flow requirements. This minimum flow is maintained automatically by providing MEW flow through recirculation control valves, CV-071 0 (P-i B) and CV-07i i (P-iA), which control recirculation flow back to the main condenser. Both CV-07i 0 and CV-071 1 are designed such that a loss of instrument control air or electrical control power will cause the recirculation control valves to fully open. During normal operation, both recirculation control valves are in the closed position. MEW flow from the MEW pumps discharge to the steam generators through MEW regulating valves. The speed of the MEW pumps, and the modulation of the MEW regulating valves to control steam generator levels, is automated from the MEW level control system.
The immediate cause for the loss of suction pressure to the MEW pumps was the spurious opening of MEW pump, P-iA, recirculation control valve, CV-07i1. Due to the discharge piping of the MEW pumps containing an open cross connection, the spurious opening of CV-07i 1 affected both trains of MEW. The MEW level control system responded to the excess flow through CV-071 1, as designed, by raising the speed of the MEW pumps and modulating the MEW regulating valves further open to maintain steam generator levels within the normal operating band.
Operators entered the off-normal procedure for a loss of MEW and took manual actions to stabilize the MEW system. Due to the combined effects of the excess discharge flow through CV-071 1 to the main condenser, and additional discharge flow from the MEW pumps required to maintain steam generator levels within the normal operating range, suction pressure to the MEW pumps continued to lower to their trip setpoint.
The spurious opening of recirculation control valve, CV-07i 1, resulted from a loss of electrical control power to the valve. The loss of electrical control power to CV-071 1 was due to insufficient contact between the fuse holder clip and a fuse ferrule within a fuse block assembly of the control power circuitry. The possible causes for the lack of sufficient contact betweenU.S. NUCLEAR REGULATORY COMMISSION (10-2010)
LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE SEQUENTIAL REVISION YEAR NUMBER NUMBER PALISADES NUCLEAR PLANT 05000255 2011 008
- - 00 3
OF 3 the fuse holder clip and the fuse ferrule were the stretching of the fuse holder clip during the installation/removal of a fuse tag-out device, and/or fatigue of the fuse holder clip, induced by the repetitive removal/installation of the fuse.
CORRECTIVE ACTIONS TAKEN The control power circuitry, for CV-071 1, was rerouted through an acceptable spare fuse block assembly and reenergized. The fuses in the control power circuitry for CV-071 1, installed at the time of the event, were analyzed and found acceptable. The fuse holder clip and fuse ferrules associated with the MEW pump, P-1B, recirculation control valve, CV-0710, were examined for a possible common mode failure mechanism and found acceptable.
CORRECTIVE ACTIONS TO BE TAKEN The procedures and training associated with the installation/removal of fuses and tag-out devices will be evaluated and revised, as necessary, to mitigate similar future occurrences.
ASSESSMENT OF SAFETY CONSEQUENCES
The event is considered to be of very low safety significance. All safety systems functioned as expected.
PREVIOUS SIMILAR EVENTS
None