05000237/LER-2009-008, Regarding Core Spray Break Detection Instrument Line Not Seismically Supported
| ML100540127 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 02/05/2010 |
| From: | Hanley T Exelon Generation Co, Exelon Nuclear |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| SVPLTR # 10-0005 LER 09-008-00 | |
| Download: ML100540127 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 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 |
| 2372009008R00 - NRC Website | |
text
Exelkn Exelon Generation www.exeloncorp.com Dresden Generating Station 65oo North Dresden Road Morris, IL 60450-9765 10 CFR 50.73 SVPLTR # 10-0005 February 5, 2010 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Dresden Nuclear Power Station, Unit 2 Renewed Facility Operating License No. DPR-19 NRC Docket No. 50-237
Subject:
Licensee Event Report 237/2009-008-00, Unit 2 Core Spray Break Detection Instrument Line not Seismically Supported Enclosed is Licensee Event Report 237/2009-008-00, Unit 2 Core Spray Break Detection Instrument Line not Seismically Supported, for Dresden Nuclear Power Station, Unit 2. This event is being reported in accordance with 10 CFR 50.73(a)(2)(ii)(A), Any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.
There are no regulatory commitments contained in this submittal.
Should you have any questions concerning this letter, please contact Ms. Marri Marchionda at (815) 416-2800.
Respectfully, Tim Hanley Site Vice President Dresden Nuclear Power Station Enclosure cc:
Regional Administrator - NRC Region III NRC Senior Resident Inspector - Dresden Nuclear Power Station I/
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 08/31/2010
ý9-2007)
Estimated burden per response to comply with this mandatory collection request: 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />. Reported lessons learned are incorporated into the licensing process and fed back to industry. Send comments regarding burden estimate to the Records and FOIA/Privacy Service Branch (T-5 F52), U.S. Nuclear Regulatory Commission, LICENSEE EVENT REPORT (LER)
Washington, DC 20555-0001, or by internet e-mail to infocollects@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not display a currently valid OMB control (See reverse for required number of number, the NRC may not conduct or sponsor, and a person is not required to respond digits/characters for each block) to, the information collection.
- 3. PAGE Dresden Nuclear Power Station, Unit 2 05000237 1 OF 4
- 4. TITLE Unit 2 Core Spray Break Detection Instrument Line not Seismically Supported
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED SEQUEATI R
FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NO.
MONTH DAY YEAR N/A N/A FACILITY NAME DOCKET NUMBER 11 03 2009 2009 008 00 02 05 2010 N/A N/A
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
El 20.2201(b)
[I 20.2203(a)(3)(i)
El 50.73(a)(2)(i)(C)
El 50.73(a)(2)(vii)
El 20.2201(d)
El 20.2203(a)(3)(ii)
Z 50.73(a)(2)(ii)(A)
El 50.73(a)(2)(viii)(A)
[I 20.2203(a)(1)
El 20.2203(a)(4)
El 50.73(a)(2)(ii)(B)
E] 50.73(a)(2)(viii)(B)
El 20.2203(a)(2)(i)
El 50.36(c)(1)(i)(A)
El 50.73(a)(2)(iii)
E] 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL El 20.2203(a)(2)(ii)
[I 50.36(c)(I)(ii)(A)
[] 50.73(a)(2)(iv)(A) 0l 50.73(a)(2)(x)
El 20.2203(a)(2)(iii)
El 50.36(c)(2)
El 50.73(a)(2)(v)(A) 0l 73.71(a)(4) 000 [3 20.2203(a)(2)(iv) 0l 50.46(a)(3)(ii) 0l 50.73(a)(2)(v)(B) 0l 73.71(a)(5)
[0] 20.2203(a)(2)(v)
El 50.73(a)(2)(i)(A)
El 50.73(a)(2)(v)(C)
El OTHER El 20.2203(a)(2)(vi)
El 50.73(a)(2)(i)(B)
El 50.73(a)(2)(v)(D)
Specify in Abstract below or in disconnected, the instrument line could not be qualified to meet the plant's seismic design qualifications.
Consequently, in the event that the plant experienced ground acceleration equivalent to a safe shutdown earthquake (0.2 g horizontal; 0.133 g vertical), the line could have failed resulting in an unisolable leak from the reactor vessel.
The condition represents degradation of a principal safety barrier and is being reported in accordance with 10 CFR 50.73(a)(2)(ii)(A).
C.
Cause of Event
An investigation was initiated to review the condition and determine the potential causes. Interviews with plant personnel were conducted; however, no relevant information was obtained to aid in cause determination. A review of work history and corrective action documents did not reveal any indication of when the supports may have been broken or disconnected. The cause of this condition is indeterminate. The most likely cause of the instrument line being disconnected from the existing supports and the support being broken from the wall is attributed to personnel stepping on or attaching equipment to the piping and/or the supports.
Previous inspections of the drywell had not identified this condition; therefore no corrective action documents had been generated to rectify the issues.
The type of supports originally designed for this instrument line utilized an open bottom Unistrut design, which allows the clips to slide out the bottom of the support if they become loose and a downward force is applied.
D.
Safety Analysis
In the event that the plant experienced a safe shutdown earthquake and the instrument line failed, a small break loss of coolant would have occurred. The first indication of this condition would have been the associated line-break annunciator in the main control room. The three-quarter inch break-size is well within the break spectrum assumed in the accident analyses. A break of this size would not have resulted in rapid depressurization. The High Pressure Coolant Injection [B1J] or the Automatic Depressurization System was available to mitigate the consequences of an unisolable small break in the primary coolant boundary.
Therefore, reactor makeup capability was not affected and was available to provide adequate flow for level control in the event of a leak. Emergency Core Cooling Systems were operable and capable of performing their intended safety functions during the time that the instrument line was not properly supported. Therefore, the safety significance of this event is minimal due to the health and safety of the public not being compromised.
E.
Corrective Actions
The supports for the instrument line were restored to their original design configuration prior to startup from the outage.
A request for training was initiated to include a discussion regarding the investigation of the as-found condition of the instrument line for contractor in-processing and in-house training.
A preventative maintenance activity will be created to perform pipe support walk downs in specified areas of the drywell prior to closeout.
An action was created to perform walk down of the Unit 3 drywell and verify that piping supports are connected and tight.
F.
Previous Occurrences
A review of DNPS Licensee Event Reports (LERs) was performed and the following events were identified.
249/95-023 (12/19/1995) Unsupported Cable In Panel 903-33 Could Have Rendered Safety Related Relays Inoperable During A Seismic Event Due To Inadequate Modification Package 249/94-022 (12/14/1994) Control Rod Drive Insert/Withdraw Lines Outside Design Basis Due To Inadequate Seismic Supports G.
Component Failure Data
N/A IPRINTED ON RECYCLED PAPERPRINTED ON RECYCLED PAPER