05000482/LER-2008-003, Re Manual Reactor Trip Due to Loss of Steam Generator Level

Re Manual Reactor Trip Due to Loss of Steam Generator Level

ML081420013
Person / Time
Site:	Wolf Creek
Issue date:	05/13/2008
From:	Matthew Sunseri Wolf Creek
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
WO 08-0013 LER 08-003-00
Download: ML081420013 (4)

LER-2008-003, Re Manual Reactor Trip Due to Loss of Steam Generator Level

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 10 CFR 50.73(a)(2)(iv), System Actuation
4822008003R00 - NRC Website

text

CREEK OPERATING CORPORATION Matthew W. Sunseri Vice President Operations and Plant Manager May 13, 2008 WO 08-0013 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555

Subject:

Docket No. 50-482: Licensee Event Report 2008-003-00, Manual Reactor Trip due to loss of Steam Generator Level Gentlemen, The enclosed Licensee Event Report (LER) 2008-003-00 is being submitted pursuant to 10 CFR 50.73(a)(2)(iv)(A) regarding a manual reactor trip at Wolf Creek Generating Station.

This letter contains no commitments. If you have any questions concerning this matter, please contact me at (620) 364-4008, or Mr. Richard D. Flannigan, Manager Regulatory Affairs at (620) 364-4117.

Sincerely, Matthew W. Sunseri MWS/rlt Enclosure cc:

E. E. Collins (NRC), w/e V. G. Gaddy (NRC), w/e B. K. Singal (NRC), w/e Senior Resident Inspector (NRC), w/e P.O. Box 411 / Burlington, KS 66839 / Phone: (620) 364-8831 An Equal Opportunity Employer M/F/HCNET Jf2-g~

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 06/30/2007 (6-2004)

, the NRC may (See reverse for required number of not conduct or sponsor, and a person is not required to respond to, the digits/characters for each block) information collection.

/3. PAGE WOLF CREEK GENERATING STATION 05000 482 1 OF 3

4. TITLE Manual Reactor Trip due to loss of Steam Generator Level

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL REV MONTH DAFYAR IIT NAEDCETNME NUMBER NO.

I05000 IFACILITY NAME DOCKET NUMBER 03 17 2008 2008 003 00 05 1

20805000

9. OPERATING MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply) 1 20.2201(b)

[]

20.2203(a)(3)(i)

E] 50.73(a)(2)(i)(C)

[]

50.73(a)(2)(vii) 5 20.2201(d)

E] 20.2203(a)(3)(ii) 5 50.73(a)(2)(ii)(A) 5] 50.73(a)(2)(viii)(A) 5 20.2203(a)(1) 5 20.2203(a)(4) 5 50.73(a)(2)(ii)(B) 5 50.73(a)(2)(viii)(B) 5 20.2203(a)(2)(i) 5 50.36(c)(1)(i)(A) 5 50.73(a)(2)(iii) 5 50.73(a)(2)(ix)(A)

10. POWER LEVEL 5

20.2203(a)(2)(ii) 5 50.36(c)(1)(ii)(A)

[]

50.73(a)(2)(iv)(A) 5 50.73(a)(2)(x) 10 20.2203(a)(2)(iii) 5] 50.36(c)(2) 5 50.73(a)(2)(v)(A) 5 73.71(a)(4) 100 5] 20.2203(a)(2)(iv) 5 50.46(a)(3)(ii)

E] 50.73(a)(2)(v)(B) 5 73.71(a)(5)

[]

20.2203(a)(2)(v) 5 50.73(a)(2)(i)(A) 5 50.73(a)(2)(v)(C) 5 OTHER E] 20.2203(a)(2)(vi)

El 50.73(a)(2)(i)(B)

[J 50.73(a)(2)(v)(D)

Specify in Abstract below or in (If more space is required, use additional copies of (If more space is required, use additional copies of NRC Form 366A)

In addition to being installed in the wrong configuration, the cross-tie arrangement of the busses was in place for approximately eleven hours. This condition resulted in a higher than normal current load being applied to the connection for an extended period of time. Together with the loose connection, the higher current caused overheating of the individual conductors and insulation resulting in a fault to ground.

BASIS FOR REPORTABILITY:

The reactor trip and subsequent actuation of Engineered Safety Features (ESF) described in this event is reportable per 10 CFR 50.73(a)(2)(iv)(A), which requires reporting of "Any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B) of this section." Paragraph (B)(1) of 10 CFR 50.73(a)(2)(iv) includes "Reactor protection system (RPS) including: reactor scram or reactor trip" and "PWR auxiliary or emergency feedwater system."

ROOT CAUSE:

The cause of the XPB03 incident was human error. The multi-cable conductor connector was not properly installed after DOBLE testing. The connector saddle size was too large for the conductor causing a loose connection when torque was applied. With higher than normal connection resistance and current load due to busses PB03 and P804 being cross-tied, the conductor insulation became overheated and resulted in a fault to ground.

CORRECTIVE ACTIONS

Work Orders were issued to inspect the other transformers that had similar connectors. Thermography was performed on the other transformers and no other indications of overheating were found.

DOBLE testing was re-performed on transformer XPB03 and associated bushings. There was no damage to the transformer other than the bushings and cable that were previously identified.

Personnel error issues were handled within station procedures.

SAFETY SIGNIFICANCE

The loss of transformer XPB03 indirectly caused the Wolf Creek Generating Station to be manually tripped from 100% power. The safety significance of this event is low. This event is bounded by the current licensing basis analysis as reported in WCGS Updated Safety Analysis Report (USAR) section 15.2.7, "Loss of Normal Feedwater Flow." There were no adverse effects on the reactor core, the reactor coolant system, or the main steam system, due to the auxiliary feedwater system's capacity to supply the necessary heat sink. All safety related equipment performed as designed and there were no adverse effects on the health and safety of the public.

OPERATING EXPERIENCE/PREVIOUS EVENTS:

None