ML20207J706
| ML20207J706 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 12/22/1986 |
| From: | Sells D Office of Nuclear Reactor Regulation |
| To: | Andrews R OMAHA PUBLIC POWER DISTRICT |
| References | |
| TAC-63967, NUDOCS 8701080563 | |
| Download: ML20207J706 (4) | |
Text
ODb DEC 2 21986 Docket No. 50-285 DISTRIBUT_ ION clocket File)
EJordan NRC PDR BGrimes t PDR JPartlow Mr. R. L. Andrews PBD#8 Rdg NThompson Division Manager - Nuclear Production FMiraglia Gray File Omaha Public Power District OGC-Bethesda PKreutzer 1623 Harney Street ACRS-10 DSells Omaha, Nebraska 68102
Dear Mr. Andrews:
SUBJECT:
CORE RELOAD METHODOLOGY CHANGES The staff has completed its review of your submittal dated November 17, 1986 dealing with your request for changes to the core reload methodology used for Fort Calhoun Station, Unit No. 1.
As a result of our review, the staff has determined that the information provided did not provide sufficient detail to complete the review.
The additional information required is identified in the enclosure to this letter.
In order to complete this review in a timely manner, consistent with your schedule for Cycle 11 reload, it is requested that you provide the requested information by January 30, 1987.
If you cannot meet this date, please contact me at 301-492-8144 to establish an acceptable date that will meet your requirements.
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I would also like to emphasize the need to receive your reload amendment application as early in the near year as possible.
Every effort should be made to submit that application by January 15, 1987 and certainly no later than January 30, 1987 to ensure timely processing.
The request for information contained in this letter affects fewer than 10 respondents; therefore, OMB clearance is not required under P.L.96-511.
Sincerely,
.co g w w ea %
Donaid E. Sells, Project Manager PWR Project Directorate #8 Division of PWR Licensing-B
Enclosure:
Request for Additional Information cc w/ enclosure:
See next page
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PBD#8 Pf PBD#8 PKt est2er 06 ells;cf AThadani 12/g/86 12/ff/86 12/22/86 8701080563 861222 UR ADOCK 05000285 POR
s Mr. R. L. Andrews Fort Calhoun Station Omaha Public Power District Unit No. I cc:
Harry H. Voigt, Esq.
LeBoeuf, Lamb, Leiby & MacRae 1333 New Hampshire Avenue, NW Washington, D.C.
20036 Mr. Jack Jensen, Chairman Washington County Board of Supervisors Blair, Nebraska 68008
)
Mr. Phillip Harrell, Resident Inspector U.S. Nuclear Regulatory Commission P. O. Box 309 Fort Calhoun, Nebraska 68023 Mr. Charles B. Brinkman, Manager Washington Nuclear Operations C-E Power Systems 7910 Woodmont Avenue -
Bethesda, Maryland 20814 i
Regional Administrator, Region IV U.S. Nuclear Regulatory Commission Office of Executive Director for Operations 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 Harold Borchert Director Division of Radiological Health Department of Health 301 Centennial Mall, South P.O. Box 95007 Lincoln, Nebraska 68509
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REQUEST FOR ADDITIONAL INFORMATION FORT CALHOUN STATION, UNIT NO. 1 CYCLE 11 RELOAD METHODOLOGY CHANGES i
1.
Although previous benchmarking of OPPD calculations using current method-ology (DIT-ROCS) over several cycles (Cycles 6, 7 and 8) of Fort Calhoun operation has shown acceptable results, the staff had previously stated that the data base in some instances was limited (Letter from J. R. Miller 1
to W. C. Jones, " Safety Evaluation by the Office of Nuclear Reactor Regulation, Omaha Public Power District, Fort Calhoun Station, Unit No. 1, DocketNo.50-285,"May11,1984). Therefore, the staff recomended con-tinuation of the ongoing benchmarking program in order to provide continuing assurance of the model applicability and the calculational accuracy. Based on this and on the statement in Section 5.7 that this program will provide additional verification data in the future, please coment on any new results of this verification program obtained from Cycles 9 and 10 and on the worthiness of updating OPPD-NA-8302 to include this data.
i 2.
Reference is made to " Omaha Batch M Reload Fuel Design Repcrt," CEN-347(0)-P, November 1986, which describes the CE Batch M reload fuel and in which the power-to-centerline melt limit has apparently been increased to 22 kw/ft from 21 kw/ft. Please provide the staff with copies of this fuel design report for review and evaluation.
3.
The enthalpy values given in Section 5.10.2 as fuel and cladding damage threshholds are not mentioned in Reference 5-5 for the CEA ejection ac-cident as stated. Also, in addition to clad failure, which is used to detennine the source term for dose calculations, the NRC limiting cri-terion for fuel fa1 lure, as stated in Reg Guide 1.77, is that the average i
fuel pellet enthalpy at the hot spot be less than or equal to 280 calories /
gram. The staff feels that this value provides a conservative maximum limit to ensure eration of press.that prompt r9pture of fuel pins and the consequent gen-ure pulses with resultant core damage does not occur.
l Therefore, Item I in Section 5.10.2 should refer to the 280 calories / gram criterion as it previously did. The 280 calories / gram criterion should also be mentioned in Sections 5.10.3 and 5.10.5.
4 I
4.
The CEA ejection event and the LOCA are being reanalyzed to allow for up to 6% tube plugging per steam generator.
Please comment on the effects of i
tube plugging on other events such as RCS depressurization, main steam i
line break, four-pump loss of coolant flow, seized rotor, CEA withdrawal.
l boron dilution, etc and the need for reanalyzing these events because of plugging.
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' 5.
Although analyses of the loss of load and loss of feedwater events were performed to account for steam generator tube plugging, no mention is made of the amount of tube plugging assumed in the description of these analyses. The staff recommends that this value be included in the key parameters and analysis assumptions.
6.
Chapter 14.9.1 " Loss of Load to Both Steam Generators" of the Fort Calhoun USAR states that the peak RCS pressure during a loss of load to both steam generators event is the transient initiated from 2122 psia. Table 5.12.4-1
" Key Parameters Assumed In The Loss Of Load To Both Steam Generators Analysis" of the proposed methodology assumes the transient begins at the minimum RCS pressure allowed by technical specifications (TS 2.20.4 requires minimum RCS pressure of 2075 psia).
Please provide justification for using minimum RCS pressure in proposed methodology when current Chapter 14.9.1 analyses state that transient initiated from 2122 psia is the most limiting case.
7.
Chapter 14.10.1 " Loss of Feedwater Flow" of Fort Calhoun USAR assumes an initial reactor pressure of 2053 psia for ar,alyzing a loss of feedwater Table 5.13.4.1 " Key Parameters Assumed In The Loss Of Feedwater event.
Flow Analysis" of the proposed methodology assumes the transient begins at the minimum RCS pressure allowed by technical specifications (TS 2.10.4 requires minimum RCS pressure of 2075 psia).
Please provide justification for using minimum RCS pressure in proposed methodology when current Chapter 14.10.1 analyses assume 2053 psia.
8.
Chapters 14.10.1 and 14.9.1 of Fort Calhoun USAR analyses both assumed beginning-of-cycle (BOC) kinetics paiameters. Tables 5.12.4-1 and 5.13.4-1 of the proposed methodology both assume end-of-cycle (E0C) kinetics para-meters.
Please provide justification for using EOC kinetics parameters when current Chapter 14.10.1 and 14.9.1 analyses assume BOC kinetics i
parameters.
9.
Chapters 14.10.1 and 14.9.1 of Fort Calhoun USAR analyses both assumed a 0.80 multiplier was applied to the Doppler Coefficient. Tables 5.12.4-1 and 5.13.4-1 of the proposed methodology assume a 0.85 Fuel Temperature Coefficient Multiplier.
Please provide justification for using a 0.85 multiplier in the proposed methodology when the current Chapter 14.10.1 and 14.9.1 ana,1yses assume a 0.80 multiplier.
- 10. Please note the following typos in OPPD-NA-8303, Rev. 02:
(a) Page 49-second line in second paragraph
" temperature" (b) Page 63A-seventh line from bottom-change " internal" to " interval"
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