05000219/LER-2004-002, Regarding Change in Methodology Used by General Electric and Global Nuclear Fuels to Demonstrate Compliance with Emergency Core Cooling System Performance Requirements
| ML042010033 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 07/09/2004 |
| From: | Swenson C AmerGen Energy Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 2130-04-20142 LER 04-002-00 | |
| Download: ML042010033 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 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)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(B) |
| 2192004002R00 - NRC Website | |
text
AmerGen.
AmerGen Energy Company, LLC www.exeloncorp.coM An Exelon Company Oyster Creek US Route 9 South, P.O. Box 388 Forked River, NJ o8731-0388 10 CFR 50.73 July,9, 2004 2130-04-20142 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 - 0001 Oyster Creek Generating Station Facility Operating License No. DPR-16 NRC Docket No. 50-219
Subject:
Licensee Event Report 2004-002-00: Change in Methodology Used by General Electric and Global Nuclear Fuels to Demonstrate Compliance with Emergency Core Cooling System Performance Requirements Enclosed is Licensee Event Report 2004-002, Revision 0. This event did not affect the health and safety of the public or plant personnel and is being submitted as a Voluntary Report. A supplemental LER is expected based on further analyses that have been initiated. There are no new regulatory commitments contained in this report.
If any further information or assistance is needed, please contact David Fawcett at 609-971-4284.
Sincerely,
<~4Y~46-/.6t A1i WtAI$3 AJ C. N. S'venson Vice President, Oyster Creek Generating Station CNS/DIF Enclosure cc:
H. J. Miller, Administrator, USNRC Region I P. S. Tam, USNRC Senior Project Manager, Oyster Creek R. J. Summers, USNRC Senior Resident Inspector, Oyster Creek File No. 04105
WC FORM 366 U.S. NUCLEAR REGULATORY APPROVED BY OMB NO. 3150-0104 EXPIRES 7-31-2004 (7-2001)
COMMISSION digits/characters for each block) the NRC may not conduct or sponsor, and a person Is not required to respond to, the information collection.
- 1. FACILUTY NAME
- 2. DOCKET NUMBER
- 3. PAGE Oyster Creek, Unit 1 05000219 1
OF 5
- 4. TITLE Change in the Methodology Used by General Electric and Global Nuclear Fuel to Demonstrate Compliance with Emergency Core Cooling System Performance Requirements
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MO DAY YEAR YEAR I SEQUENTIAL REV MO DAY YEAR FACILITY NAME DOCKET NUMBER I NUMBER NO I I
_ L L_
0 5 0 0 0 05 1 4 2004 2004 -
002 -
00 07 09 2004 FACILITY NAME DOCKETNUMBER i n...
I 05000
- 9. OPERATING N
- 11. THIS REPORTIS SUBMITED PURSUANTTO THE REQUIREMENTS OF 10 CFR 5: (Check all that apply)
MODE
- 10. POWER 100
- - 20.2201(b) 20.2203(a)(3)(ii) l 50.73(a)(2)Qii)(B) l 50.73(a)(2)(ix)(A)
LEVEL
_ 20.2201 (d) 20.2203(a)(4) 50.73(a)(2)(iii) 50.73(a)(2)(x) 20.2203(a)(1) 50.36(c)(1)(i)(A) 50.73(a)(2)(iv)(A) 73.71 (a)(4) 20.2203(a)(2)(i) 50.36(c)(1)(ii)(A) 50.73(a)(2)(v)(A) 73.71 (a)(5)
_20.2203(a)(2)(ii)
_50.36(c)(2) 50.73(a)(2)(v)(B)
X OTHER 20.2203(a)(2)(iii)
X 50.46(a)(3)(ii)
_50.73(a)(2)(v)(C)
Voluntary Report 20.2203(a)(2)(iv) 50.73(a)(2)(i)(A) 50.73(a)(2)(v)(D) 20.2203(a)(2)(v) 50.73(a)(2)(i)(B) 50.73(a)(2)(vii) 20.2203(a)(2)(vi) 50.73(a)(2)(i)(C) 50.73(a)(2)(viii)(A) i__ _
20.2203(a)(3)(i) 50.73(a)(2)(ii)(A) 50.73(a)(2)(viii)(B)
- 12. LICENSEE CONTACT FOR THIS LER NAME TELEPHONE NUMBER (Include Area Code)
Fred Buckley, Manager - Reactor Engineering (609) 971-4941X MES (if yes, complete EXPECTED SUE
ABSTRACT
On May 14, 2004, at approximately 0813 hours0.00941 days <br />0.226 hours <br />0.00134 weeks <br />3.093465e-4 months <br /> with the plant operating in the RUN Mode at 100% power, General Electric (GE) informed Oyster Creek Generating Station (OCGS) of a change in the calculation of Peak Cladding Temperature (PCT) and maximum local cladding oxidation. A new heat source has been postulated during the Loss of Coolant Accident (LOCA) event that involves the recombination of hydrogen and oxygen within the fuel bundles during core heatup. Based on 10 CFR 50, Appendix K, inputs and assumptions, the additional heat generated resulted In an estimated 25 degrees F increase in PCT and a 1.73% increase in maximum local cladding oxidation. Consequently, the previous LOCA analysis was non-conservative relative to PCT and maximum local cladding oxidation. This event was initially reported to the NRC on May 14, 2004 in a voluntary notification as a result of the 10 CFR 50.46(a)(3)(ii) requirement to report this issue in accordance with 10 CFR50.72 and 10 CFR50.73.
The cause of this event is that the potential oxygen source and subsequent heating effects of the hydrogen-oxygen recombination phenomenon were not properly considered during the original development of the LOCA evaluation methodology.
Interim Corrective actions include a Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) reduction and associated Core Operating Limits Report (COLR) revision and an 8% Peak Linear Heat Generation Rate (PLHGR) reduction when the containment is allowed to be deinerted above 25% power. Long term corrective actions include initiation of additional analysis to further evaluate the hydrogen-oxygen recombination phenomenon using more appropriate best-estimate/upper bound LOCA evaluation methodology. Based on the results of this analysis and a probabilistic risk analysis based evaluation relative to operation while deinerted, the above interim corrective actions may be revised and/or revisions to OCGS Technical Specifications may be implemented. OCGS is expecting to provide a supplement to this LER following completion of these further evaluations.
No previous similar events were identified.
NRC FORM 366 (7-2001)
(If more space Is required, use additional copies of (If more space is required, use additional copies of (If more space is required, use additional copies of (If more space is required, use additional cc ies of NRC Form 366A)
Additional Information
A.
Failed Components:
None
B. Previous similar events
No previous similar events were identified.
C.
Identification of components referred to in this Licensee Event Report:
Components IEEE 805 System ID Reactor Core AC ECCS BM Reactor Vessel AD Primary Containment NH IEEE 803A Function RCT P. MO RPV VSL