ML20206G867
| ML20206G867 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 04/03/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20206G856 | List: |
| References | |
| TAC-63967, NUDOCS 8704150182 | |
| Download: ML20206G867 (4) | |
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NUCLEAR REGULATORY COMMISSION o
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION, UNIT NO. 1 DOCKET N0. 50-285 FORT CALHOUN RELOAD CORE METHODOLOGY CHANGES
1.0 INTRODUCTION
0 By letter dated November 17, 1986 (LIC-86-537) as supplemented by letters dated January 26, 1987, March 9 and 18, 1987, Omaha Public Power District j
(0 PPD), the licensee, submitted modifications to their reload methodology topical reports (Refs. 1, 2 and 3). These modifications are necessitated j
because of changes in the Fort Calhoun Station operation during Cycle 11.
l Beginning with Cycle 11, Combustion Engineering (CE) will again be supplying reload fuel. Advanced Nuclear Fuels Corporation (ANF) had t
provided Fort Calhoun Station with new fuel for Cycles 6 through'10.
Except for the control element assembly (CEA) ejection accident and the-loss-of-coolant accident (LOCA), the staff has previously found it acceptable for OPPD to perform the transient and accident analyses for Fort Calhoun reload cores using documented CE computer codes and methodology (Ref. 4). Therefore, for Cycle 11, CE is reanalyzing the CEA ejection and LOCA events for OPPD.
2.0 EVALUATION The major revision to OPPD-NA-8301, Revision 1, which provides an over-view of the OPPD reload core analysis methodology, refers to CE as the current nuclear fuel vendor-and adds the CE fuel mechanical design and design methods as a reference. Other changes are either editorial or typographical error corrections. The staff finds the proposed changes to OPPD-NA-8301-P, Revision 1, acceptable.
Minor revisions to OPPD-NA-8302-P, which describes the steady-state nuclear analysis methods used by OPPD for reload' analysis for Fort Calhoun, are also-i, primarily due to the change in reload fuel vendor'from'ANF to CE.
In-addition to typographical error corrections, the generation of neutron cross-sections for CE shimmed fuel assemblies has been added. Also, the approved ROCS. computer code has replaced the BETAF computer code for the calculation of neutron kinetics parameters.
In response to the staff's recommendations, additional benchmarking of OPPD calculations with Cycles 9 and 10 measurements will be included in the OPPD-NA-8302 update.
The staff finds the proposed changes to 0 PPD-NA-8302-P acceptable.
Revisions are also made to OPPD-NA-8303-P, which describes the OPPD-reload core transient and accident methods for application to F. ort Calhoun. Some of these revisions are, again, editorial in nature. An evaluation'of the major revisions follows.
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' For Cycle 11 operation of Fort Calhoun, the four part-length CEA's will be replaced by full-length CEA's. Since the Technical Specifications for Fort Calhoun prohibit the use of the part-length CEA's (Group P) during operations, the use of full-length CEA's in the place of the part-length CEA's will have no impact on operation. Technical Specifications will still require that Group P be withdrawn above 114 inches. Although no credit will be taken for Group P in the event of a reactor trip (Ref. 5), credit for additional shutdown margin may be taken for these rods when they are driven in by the operators following a reactor shutdown. The drop of a Group P CEA is considered in the full-length CEA drop analysis. The physical dimensions of the full-length CEA's are identical to those of the part-length CEA's. The staff. finds the replacement of the part-length CEA's by the full-length CEA's acceptable.
Since CE is the current fuel vendor, the large break and small break LOCA analysis will be performed by CE using their approved metholodology to show compliance with the 10 CFR 50.46 acceptance criteria. The CEA ejection accident will also be analyzed by CE using their approved methodology (Ref. 6). Therefore, references to CE methodology and analysis for both the LOCA and the CEA ejection events are acceptable.
The power-to-fuel centerline melt limit is being increased from 21 kw/ft to 22 kw/ft.
In response to the staff's request for justification of this increase (Ref. 7), OPPD submitted CEN-347(0)-P, Revision 1-P (Ref. 8). The power-to-fuel centerline melt limit was calculated using the NRC-approved FATES 3A fuel performance computer code. The analyses were performed for both fresh CE fuel to be introduced in Cycle 11 and burned ANF fuel being reused in Cycle 11. The new limit takes credit for the decrease in power peaking which is characteristic of highly burned fuel. The staff has previously performed audit calculations of this limit for other CE plants using 14 x 14 fuel (e.g., St. Lucie and Calvert Cliffs) and found it to be acceptable. Since Fort Calhoun fuel is of the same design, the staff finds the proposed increase in the power-to-fuel centerline melt limit from 21 kw/ft to 22 kw/ft acceptable.
In addition to the transient and accident analyses for Fort Calhoun reloads that the staff has previously found acceptable for OPPD to perform, two additional events will be analyzed using the CESEC-III computer code.
These are the loss-of-load to both steam generators and the loss-of-main feedwater events. For modifications to the plant which potentially degrade reactor coolant system (RCS) heat removal capability (including steam generator tube plugging), it is necessary to assure that the following i
acceptance criteria will still be met:
a)
The peak RCS pressure does not exceed 110% of design pressure.
b)
The transient minimum departure from nucleate boiling ratio (DNBR) is greater than the 95/95 confidence interval limit for the CE-1 correlation limit.
c)
The Peak Linear Heat Generation Rate (PLHGR) does not exceed 22 kw/ft.
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The staff has reviewed the proposed methodology for the loss-of-load to both steam generators and the loss-of-main feedwater events (Ref. 3).
The staff concludes that the Key Parameters and Analysis Assumptions contain sufficient conservatism to account for analysis uncertainties.
Any analyses performed with this methodology which do not exceed the above criteria, would provide adequate assurance that the criteria would not be exceeded during an actual event. Therefore, the staff finds that proposed methodology to be acceptable.
OPPD intends to utilize the CESEC-III computer code to analyze the loss-of-load to both steam generators and the loss-of-main feedwater events. The CESEC III computer code has not been generically approved by the staff for all CE designed plants. Therefore, the staff requested that OPPD demonstrate the applicability of the CESEC-III computer code to Fort Calhoun.
The staff has reviewed OPPD's and CE's responses (Refs. 9,10,11 and 12), including sample plant specific Fort Calhoun input data for the CESEC-III computer code. The staff concludes that OPPD has demonstrated that the CESEC-III computer code is applicable to Fort Calhoun.
In addition, the staff has previously reviewed OPPD's ability to correctly use the CESEC-III computer code (Ref. 4). Therefore, the staff finds it acceptable for OPPD to utilize the CESEC-III computer code to analyze the loss-of-load to both steam generators and the loss-of-main feedwater events for Fort Calhoun.
3.0 CONCLUSION
The staff has reviewed the proposed changes to the Fort Calhoun Reload Core Methodology Reports (Refs. 1, 2 and 3) submitted by 0 PPD on November 17, 1986, as supplemented by letters dated January 26, 1987, March 9 and 18, 1987, and finds them acceptable.
Date:
April 3,1987 Principal Contributors:
L. Kopp, J. Ramsey
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4.0 REFERENCES
1.
"0 PPD Nuclear Analysis Reload Core Analysis Methodology Overview,"
0 PPD-NA-8301-P, Rev. 01, June 1985.
2.
"0 PPD Nuclear Analysis Reload Core Analysis Methodology, Neutronics Design Methods and Verification," OPPD-NA-8302-P, September 1983.
3.
"0 PPD Nuclear Analysis Reload Core Analysis Methodology, Transient and Accident Analysis Methods and Verification," 0 PPD-NA-8303-P, September 1983.
4.
Letter from J. R. Miller (NRC) to W. C. Jones (0 PPD), " Reload Core Analysis Methodology Reports," May 11, 1984.
5.
Letter from R. L. Andrews (0 PPD), " Application for Amendment of Operating License," January 22, 1987.
6.
"CE Method for Control Element Assembly Ejection Analysis,"
CENPD-190-A, July 1976.
7.
Letter from R. L. Andrews (OPPD), " Cycle 11 Core Reload Methodology Change Questions," January 26, 1987.
8.
"0maha Batch M Reload Fuel Design Report, CEN-347(0)-P, Revision 1-P,"
January 1987.
9.
Letter from R. L. Andrews (OPPD), " Core Reload Methodology Changes for Cycle 11," March 9, 1987.
10.
Letter from A. E. Scherer (CE) to F. J. Miraglia (NRC), " Applicability of CESEC-III to the Fort Calhoun Station," February 27, 1987.
11.
Letter from R. L. Andrews (OPP), " Core Reload Methodology Changes for Cycle 11," March 18, 1987.
12.
Letter from R. L. Andrews (0 PPD), " Additional Information on Cycle 11 Core Reload Methodology," March 23, 1987.
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