05000390/LER-2008-004, Automatic Reactor Trip in Response to Opening of Exciter Field Breaker
| ML083250295 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  (NPF-090) |
| Issue date: | 11/19/2008 |
| From: | Skaggs M Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LER 08-004-00 | |
| Download: ML083250295 (6) | |
| 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)(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)(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 |
| 3902008004R00 - NRC Website | |
text
10 CFR 50.73 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Gentlemen:
In the Matter of
)
Docket No. 50-390 Tennessee Valley Authority
)
WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 - LICENSEE EVENT REPORT (LER) 390/2008-004, REVISION 0 - AUTOMATIC REACTOR TRIP IN RESPONSE TO OPENING OF EXCITER FIELD BREAKER This submittal provides LER 390/2008-004. This LER documents an event where the reactor was automatically tripped by a Nuclear Assistant Unit Operator (NAUO) making a human performance error while answering a question about breaker operation. The report regarding this condition is provided in accordance with 10 CFR 50.73(a)(2)(iv)(A).
There are no regulatory commitments associated with this submittal. If you have any questions concerning this matter, please contact Chris Riedl at (423) 365-1742.
Sincerely, Mike Skaggs Site Vice President Watts Bar Nuclear Plant Enclosure cc: See Page 2 November 19, 2008 Original signed by
U.S. Nuclear Regulatory Commission Page 2 Enclosure cc (Enclosure):
NRC Resident Inspector Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381 ATTN: John G. Lamb, Project Manager U.S. Nuclear Regulatory Commission Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation MS O-8 H4 Washington, DC 20555-0001 U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303 Institute of Nuclear Power Operations 700 Galleria Parkway, NW Atlanta, Georgia 30339-5957 November 19, 2008
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION (9-2007)
LICENSEE EVENT REPORT (LER)
(See reverse for required number of digits/characters for each block)
APPROVED BY OMB: NO. 3150-0104 EXPIRES: 08/31/2010
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 1. FACILITY NAME Watts Bar Nuclear Plant
- 2. DOCKET NUMBER 05000 390
- 3. PAGE 1 OF 4
- 4. TITLE Automatic Reactor Trip in Response to Opening of Exciter Field Breaker
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL NUMBER REV NO.
MONTH DAY YEAR FACILITY NAME N/A DOCKET NUMBER N/A 09 20 2008 2008 - 004 -
0 11 19 2008 FACILITY NAME N/A DOCKET NUMBER N/A
- 9. OPERATING MODE 1
- 10. POWER LEVEL 100%
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply) 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)(i)(A) 50.73(a)(2)(iii) 50.73(a)(2)(ix)(A) 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) 20.2203(a)(2)(iv) 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 Abstract below or in II.
DESCRIPTION OF EVENT (continued):
G.
Safety System Responses All systems responded as designed and as previously stated, the operations staff took manual control of AFW to regulate steam generator level.
III.
CAUSE OF EVENT
The NAUO had lapses in judgment and violated sound operator fundamentals and procedures. The NAUO failed to meet the expectation to notify Shift Manager or use error reduction tools when opening the exciter field breaker panel for inspection. No time or schedule pressures existed that contributed to this event.
IV.
ANALYSIS OF THE EVENT
Plant safety systems performed their intended safety functions in response to the automatic reactor trip. All control rods fully inserted into the core, and plant decay heat removal functioned properly. During the trip, the C feedwater heater string (EIIS Code SJ HX) isolated. This was expected, per a previous Engineering Design Change (EDC) 52270. See Section V, Assessment of Safety Consequences, below for further discussion.
V.
ASSESSMENT OF SAFETY CONSEQUENCES
The automatic reactor trip on 09/20/2008 can be compared to the Final Safety Analysis Report (FSAR) "LOSS OF ELECTRICAL LOAD AND/OR TURBINE TRIP", Updated FSAR (UFSAR) section 15.2.7. The reactor (EIIS Code RCT) was automatically tripped at approximately 09:03 due to a loss of electrical load on the main generator (EIIS Code TB) when the exciter breaker tripped open. Subsequently, a hi-hi level in heater C2 resulted in the C-heater string isolation. Therefore, the FSAR "LOSS OF NORMAL FEEDWATER" UFSAR section 15.2.8 is also applicable to this event. Main feedwater flow was isolated due to the low Tavg with Rx trip signal. The hi-hi C-2 heater isolation was caused by the bypass to condenser level control valves (LCVs) being closed during high turbine load operation in accordance with plant design to minimize water hammer.
The plant was stabilized using Auxiliary Feedwater and the Main Steam dump valves (EIIS Code SB). The secondary side steam generator atmospheric relief valves (EIIS Code RV) and safety valves did not operate during the transient. The Reactor Coolant System responded to the initial transient as expected with no pressurizer PORV relief, no safety injection initiation, and no steam generator PORV relief.
Therefore, the 09/20/2008 trip is bounded by the FSAR safety analysis assumptions.
VI.
CORRECTIVE ACTIONS-The corrective actions for this condition are being managed within TVAs Corrective Action Program (PERs 152951, 152954, and 152955) and therefore are not considered to be regulatory commitments. An overview of the corrective action plan is provided below:
A.
Immediate Corrective Actions
- 1.
The NAUO was restricted from duty.
- 2.
A crew stand-down was held to discuss expectations for use of Human Performance Tools.
- 3.
A training brief was issued by the Operations Superintendent.
VI.
B.
Corrective Actions to Prevent Recurrence
- 1.
A stand-down brief was held during the Initial Turnover Meeting upon each crews first shift back to direct the rules for inspecting normal plant equipment for training purpose and ensure no hands-on manipulations occur without the proper procedures in hand, notifications of training being in progress, and Main Control Room concurrence.
- 2.
A training brief on TI-12.10, Control of Sensitive Equipment, was completed on each crews first shift back.
- 3.
The TI-12.10 procedure will be included in the Cycle 8 Non-Licensed Operator training.
- 4.
Reactor trip warning labels will be revised to reference the TI.
- 5.
The area around the sensitive equipment listed in TI-12.10 will be marked with a noticeable caution.
VII.
ADDITIONAL INFORMATION
A.
Failed Components None.
B.
Previous LERs on Similar Events No similar LERs have occurred at Watts Bar.
C.
Additional Information
None.
D.
Safety System Functional Failure This event did not involve a safety system functional failure as defined in NEI 99-02, Revision 5.
E.
Loss of Normal Heat Removal Consideration There was no loss of normal heat removal due to this condition.
VIII.
COMMITMENTS
None.