05000458/LER-2003-005, Operation Greater than Maximum Licensed Power Due to Erroneous Feedwater Flow Measurement
| ML031640175 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 06/09/2003 |
| From: | King R Entergy Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| G9.25.1.3, G9.5, RBF1-03-0103, RBG-46132 LER 03-005-00 | |
| Download: ML031640175 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 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) 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)(1) 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) |
| 4582003005R00 - NRC Website | |
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1.
Entey Entergy Operations, Inc.
River Bend Station 5485 U.S. Highway 61 R 0. Box 220 St. Francisville, LA 70775 Tel 225 336 6225 Fax 225 635 5068 Rick J. King Director Nuclear Safety Assurance June 9, 2003 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
Subject:
River Bend Station Docket No. 50-458 License No. NPF-47 Licensee Event Report 50-458 03-005-00 File Nos.
G9.5, G9.25.1.3 RBG-46132 RBF1 0103 Ladies and Gentlemen:
In accordance with 10CFR50.73, enclosed is the subject Licensee Event Report.
There are no commitments in this document.
Sincerely, RJKldhw enclosure
Licensee Event Report 50-458 03-005-00 June 9, 2003 RBG-46132 Page 2 of 2 cc:
U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011 NRC Sr. Resident Inspector P. 0. Box 1050 St. Francisville, LA 70775 INPO Records Center E-Mail Mr. Jim Calloway Public Utility Commission of Texas 1701 N. Congress Ave.
Austin, TX 78711-3326 Mr. Prosanta Chowdhury Program Manager - Surveillance Division Louisiana DEQ Office of Radiological Emergency Planning and Response P. 0. Box 82215 Baton Rouge, LA 70884-2215
NRC FORM 366 U.S. NUCLEAR REGULATORY APPROVED BY OMB NO. 3150-0104 EXPIRES 6-30-2001 (1-2001)
COMMISSION
, the NRC may not conduct or sponsor, and a Demnon Is not required to respond o. the information collection.
FACIUTY NAME (1)
DOCKET NUMBER (2)
PAGE (3)
River Bend Station 050- 458 1 OF 5
TITLE (4)
Operation Greater Than Maximum Licensed Power Due to Erroneous Feedwater Flow Measurement EVENT DATE (5)
LER NUMBER (6)
REPORT DATE (7)
OTHER FACILMES INVOLVED (8)
FACILITY NAME DOCKET NUMBER MO DAY YEAR YEAR NUMBER NO MO DAY YEAR 05000 FACILITY NAME DOCKET NUMBER 05 10 2003 2003 005 00 06 09 2003 05000 OPERATING
(
THIS REPORT IS UEMITTED PURSUANT TO TFE REQUIREMENTS OF t0 C R 4 (Check all that aviv) (11 IIIIODE (9)
I1 20.2201(b) 20.2203(a)(3)(ii) 50.73(a)(2)(ii)(B) 50.73(a)(2)(ix)(A)
POWER 20.2201 (d) 20.2203(a)(4) 50.73(a)(2)(iii) 50.73(a)(2)(x)
LEVEL (10) 98%
20.2203(a)(1) 50.36(c)(1)(i)(A) 50.73(a)(2)(iv)(A) 73.71 (a)(4) 20.2203(a)(2)(1) 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)
_ 50.46(a)(3)(ii)
= 50.73(a)(2)(v)(C)
_ NRC Fon 366A 20.2203(a)(2)(iv) 50.73(a)(2)(l)(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)(1)(C) 50.73(a)(2)(viii)(A) 20.2203(a)(3)(1) 50.73(a)(2)(ii)(A) 50.73(a)(2)(viii)(B) a LICENSEE CONTACT FOR THIS LER (112)
NAME TELEPHONE NUMBER (Include Area Code)
J.W. Leavines, Manager-Licensing_
225-381-4642 COMPLETE ONE UNE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)
MSNST IO ERTRTABLE I I MANU-REPORTABLE
CAUSE
SYS1EM COMPONENI FACTURER TO EPIX
CAUSE
l SYSEM lCOMNENT FACTURER TO EPIX B
SJ PE Caldon YES XIl_l_l SUPPLEMENTAL REPORT EXPECTED (14)
EXPECTED MONTH DAY YEAR l_ _
l l
I SUBM ISSION X YES (If yes, complete EXPECTED SUBMISSION DATE).
NO DATE (15) 08 l
01 2003 ABSTRACT (Limit to 1400 soaces. i.e.. anoroximately 15 sinale-soaced tvpewritten lines) (16)
On May 10, 2003, with the plant operating at 98 percent power, information was received from the manufacturer of the station's reactor feedwater flow measurement device indicating the station had been operated in excess of maximum licensed reactor thermal power In the past. This condition is being reported in accordance with the Facility Operating License (NPF-47) as a violation of License Condition 2.C(1), "Maximum Licensed Power." The investigation of this condition is in progress. A completed root cause analysis will be provided in a supplement to this report. An evaluation has determined that adequate design margin existed to accommodate the overpower condition.
NRC FORM 366 (1-2001)
(if more space is required, use additional copies of (ff more space is requied, use additionl copies of (If more space is required, use additional copies of NRC Fonn 366A) (17)
SAFETY SIGNIFICANCE
An analysis of this condition was performed to assess the postulated effects on fuel cladding integrity, reactor vessel Integrity, primary containment integrity, and post-accident activity releases.
Fuel cladding Integrity is protected by operating to a number of parameters collectively referred to as thermal limits. For River Bend, these thermal limits are average planar heat generation rate, minimum critical power ratio, and linear heat generation rate. The two percent power measurement uncertainty per 10CFR50.46 and Regulatory Guide 1.49 are already imbedded within these thermal limits so that the incremental increase over this limit of 0.7 percent of licensed power level needs to be addressed. Each thermal limit is linearly dependent upon core thermal power. As such, with the power measurement bias in place, the measure of the margin to a particular thermal limit would have been 0.7 percent less than calculated by the core monitoring system.
Operating logs were reviewed for the period of operation of most concern (i.e., from the implementation of the five percent power uprate until RF11) in order to assess the impact on fuel cladding integrity. Based upon this review, it was concluded that margin to the operating limits was sufficient to accommodate the 0.7 percent overpower.
Vessel integrity s evaluated every cycle by analyzing the MSIV closure event with failure to scram on MSIV position. This event was analyzed at 102 percent thermal power, thus, the impact of an additional 0.7 percent in initial thermal power was evaluated.
Operating at the slightly higher power limit would not significantly change the peak pressure attained as operation of the main steam safety / relief valves following the reactor scram is the primary means of limiting the pressure rise in the vessel. Further, the Cycle 10 and 11 reload analyses were reviewed to identify the margin to the acceptance criteria for the postulated overpressure event. There is margin in these analyses sufficient to absorb the 0.7 percent overpower condition. It is concluded that reactor vessel Integrity was not challenged during the period in which the external LEFM 8300 correction factors were in-service.
The current containment analysis was performed for the 5 percent power uprate project.
The 5 percent power uprate project resulted in an increase In licensed reactor power from 2894 to 3039 MWth and an Increase In reactor pressure of 30 PSI. As required, the analyses were performed assuming an initial reactor power of 102 percent, thus the impact of an additional 0.7 percent power was assessed. As part of the power uprate containment analysis, evaluations at 102 percent of the pre-and post-power uprate initial conditions were performed. These evaluations indicate a small sensitivity in peak containment parameters relative to the initial conditions assumed. The effect of the 0.7 percent overpower would much smaller than the effect of the 5 percent power uprate as (if more space is required, use additional copies of NRC Form 3664) (17) the latter included an increase in reactor dome pressure. Therefore, based upon the small sensitivity, and the existing margin to containment design limits, it is concluded that containment integrity would not have been challenged during the period in question.
The post-accident radiological dose consequence evaluation is performed to assess dose consequences to individuals at the exclusion area boundary, the low population zone, and the main control room. River Bend has recently completed upgrading these evaluations to include the Alternative Source Term per 10CFR50.67 and Regulatory Guide 1.183. The associated analyses were based upon a reactor operating at 3100 megawatts thermal, which is 102 percent of the previous licensed thermal power of 3039 megawatts thermal. Therefore, the impact of an additional 0.7 percent power was evaluated. In reviewing the subject calculations, it is concluded that the source term, and therefore the dose consequences, are linear with respect to reactor power. As such, the dose consequences would be expected to increase 0.7 percent. A review of the calculations indicates that there is more than enough margin in the 10CFR50.67 acceptance criteria to accommodate the slight overpower.
(NOTE: Energy Industry Component Identification codes are annotated as (**XX**).)