05000219/LER-2004-007, Regarding Automatic Containment Isolation Bypassed During Reactor Startup Due to an Inadequate Procedure
| ML050270512 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 01/21/2005 |
| From: | Swenson C AmerGen Energy Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 2130-05-20018 LER 04-007-00 | |
| Download: ML050270512 (3) | |
| 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)(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 |
| 2192004007R00 - NRC Website | |
text
AmerGen SM AmerGen Energy Company, LLC Oyster Creek US Route 9 South, RO. Box 388 Forked River, NJ o8731-0388 www.exeloncorp.com An Exelon Company 10 CFR 50.73 January 21, 2005 2130-05-20018 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 - 0001 Oyster Creek Generating Station Facility Operating License No. DPR-1 6 NRC Docket No. 50-219
Subject:
Licensee Event Report 2004-007-00, Automatic Containment Isolation Bypassed During Reactor Start-up Due to an Inadequate Procedure Enclosed is Licensee Event Report 2004-007, Revision 0. This event did not affect the health and safety of the public or plant personnel. This event resulted in a safety system functional failure.
If any further information or assistance is needed, please contact David Fawcett at 609-971-4284.
Since1 H
d C. N. Swenson Vice President, Oyster Creek Generating Station CNS/DIF Enclosure: NRC Form 366, LER 2004-007-00 cc:
S. J. Collins, Administrator, USNRC Region I P. S. Tam, USNRC Senior Project Manager, Oyster Creek R. J. Summers, USNRC Senior Resident Inspector, Oyster Creek File No. 05027
O)J-
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 06/302007 (6-2004)
, the NRC may not conduct or digits/characters for each block) sponsor, and a person is not required respond to, the information collection.
- 3. PAGE Oyster Creek, Unit 1 05000 219 1
OF 2
- 4. TITLE Automatic Containment Isolation Bypassed During Reactor Startup Due to an Inadequate Procedure
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE S. OTHER FACILITIES INVOLVED SEQUENTAL I REV FACILITY NAME DOCKET NUMBER MONTH DAY EAR Y
EAR NuMBER NO MONTH lDAY YEAR 06000 11 22 2004 2004 007 - 00 01 21 2005 FACILITY NAME DOCKET NUMBER 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANTTO THE REQUIREMENTS OF 10 CFR l: (Check all that apply)
N 20.2201 (b) 20.2203(a)(3)(i) 50.73(a)(2)(i)(C) 50.73(a)(2)(vii) 20.2201 (d) 20.2203(a)(3)(ii) 50.73(a)(2)(ii)(A) 50.73(a)(2)(viii)(A) 20.2203(a)(1) 20.2203(a)(4) 50.73(a)(2)(ii)(B) 50.73(a)(2)(viii)(B) 20.2203(a)(2)(i) 50.36(c)(1)(j)(A) 50.73(a)(2)lQii) 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL 20.2203(a)(2)ii) 50.36(c)(1)(ii)(A) 50.73(a)(2)(iv)(A) 50.73(a)(2)(x) 20.2203(a)(2)(iii) 50.36(c)(2) 50.73(a)(2)(v)(A) 73.71 (a)(4) 20 20.2203(a)(2)(v) 50.46(a)(3)(ii) 50.73(a)(2)(v)(B) 73.71 (a)(5) 20.2203(a)(2)(v) 50.73(a)(2)(i)(A)
/
50.73(a)(2)(v)(C)
OTHER 20.2203(a)(2)(vi) 50.73(a)(2)(i)(B) 50.73(a)(2)(v)(D)
Specify In Absrc below I___
I__
_I_
L__
_I__
or in NRC Form 368A
- 12. LICENSEE CONTACT FOR THIS LER _
FACILITY NAME TELEPHONE NUMBER ndude Area Code)
Herbert Tritt, Operations Support Manager (609) 971-4204CAUSE SYSTEM COMPONENT MANU-REPORTABLE
CAUSE
S COMPONENT MANU-REPORTABLE FACTURER TO EPIX l
FACTURER TO EPIX
- 14. SUPPLEMENTAL REPORT EXPECTED 15.EXPECTED MONTH DAY YEAR SUBMISSION I
YES (ifyes, complete EXPECTED SUBMISSION DATE)
F71 NO DATE ABSTRACT al to 1400 spacs %L.. apozitraet 15s Aspacd U ewrsa ten kres)
On November 22, 2004 at 10:41 a.m. during startup operations following 1 R20 Refueling Outage, a reactor operator (RO) bypassed the automatic Containment Ventilation and Purge Isolation (EIIS-JM) function due to an incorrect step in the plant start-up procedure. The incorrect step had the wrong nomenclature and directed the RO to a key-locked bypass control switch (CFI-HS) only intended to be used in implementing Emergency Operating Procedure (EOP) actions. The procedure guidance was reviewed and determined to be incorrect and the bypass switch realigned to normal after 1 Hour 41 Minutes. During the time that the switch was placed in bypass, the automatic isolation function of both V-27-1 and V-27-2 was bypassed and consequently the safety system function of these Primary Containment Isolation Valves to close and isolate in response to a LOCA signal was prevented. This condition is, by definition, a Safety System Functional Failure (SSFF) in accordance with NEI 99-02.
The apparent cause of this event was an incorrectly worded step in the plant startup procedure.
All other safety systems were fully operable and there were no circumstances that would have required automatic containment isolation while the function was bypassed.
There were no previous similar events involving inadvertent bypassing of the containment isolation function due to inadequate procedures.
NRC F-ORM U: (1b-ZJ4)
(If more space is required, use additional copies of NRC Form 366A)
Description of Event
On November 22, 2004 at 10:41 a.m. during startup operations following 1 R20 Refueling Outage, a reactor operator (RO) bypassed the automatic Containment Ventilation and Purge Isolation (EIIS-JM) function due to an incorrect step in the plant start-up procedure. The incorrect step had the wrong nomenclature and directed the RO to a key-locked bypass control switch (CFI-HS) only intended to be used in implementing Emergency Operating Procedure (EOP) actions. The bypass condition was addressed after reviewing the Nitrogen System and Containment Atmosphere Control (CAC) Procedure 312.11 and placing the switch in the correct position, thus clearing the alarm and exiting the condition.
Total elapsed time for the event is as follows:
10:41 a.m. Containment Ventilation and Purge Isolation Bypassed in Alarm.
12:22 a.m. Containment Ventilation and Purge Isolation Switch placed in Normal
Analysis of Event
The procedure error was the insertion of the wrong bypass switch nomenclature (Containment Ventilation and Purge Isolation Bypass vice the correct nomenclature, Drywell Vent and Purge Valves Interlock Bypass). The error was introduced into the procedure in July 2001. Startups since July 2001 had inerting in progress in accordance with procedure 312.11, prior to the Reactor Mode Switch placed in RUN, and as such, the procedure 201 step in error was not applicable. No plant conditions were present (placing the Reactor Mode Switch in RUN prior to containment being inerted) that required the bypassing of the Drywell Vent and Purge Valve Interlock until November 22, 2004.
This event is reportable in accordance with 1 OCFR50.73 (a)(2)(v), many event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to: (C) Control the release of radioactive material". This condition was also, by definition, a Safety System Functional Failure (SSFF) in accordance with NEI 99-02 for PI Data reporting purposes.
Cause of Event
The apparent cause of this event was an incorrectly worded step in the plant startup procedure 201.
Corrective Actions
Procedure 201, Plant Startup, will be revised to correct the bypass switch nomenclature.