ML15223A663
| ML15223A663 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 10/16/1980 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | Parker W DUKE POWER CO. |
| References | |
| NUDOCS 8011030057 | |
| Download: ML15223A663 (6) | |
Text
OCTOBER 1 6 1980 DISTRIBUTION:
NISC TNovak C-m y 3
ORBV4Rdg RTedesco L PDR NRR Rdg GLainas TERA-3 DEisenhut MFairtil'e Dockets Nos. 5-0-270 Rlngram and 50-287 RReid RPurple T8ovak J~oe ACRS-16 Mr. William 0. Parker, Jr.
1E3 Vice President -
Steam Production Gray Fie Duke Power Company HOBnstkwoo.
P. 0. Box 2178 41 akwo 422 South Church Street LLO is Charlotte, North Carolina 28242 ~DFieno
.. 1 '-WJohnston
Dear Mr. Parker:
In order to complete our revlIw of your Oconee Reload Design Methodology.
Technical Report Vf May 20,,1980,;we find that we need addi ti onal infor-1 mation. It is possible thatX yu maly,receiveadditional requests.
Kindly provide the requested' information with three signed originals and 37 additI onal copies within 30 da ys-of 'recei pt of 'thi's letter.
Sincerely, original signed Robed W. Reid Robert W. Reid, Chief Operating Reactors Branch #4 Division of Licensing
Enclosure:
Request for Additional Product Information cc w/enclosure:
4' See next page ny IE-3 6011030__ 5 OFFICE OB #4 :.D J.;
I.....
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NRC FORM 3:,8 (9-76) NRCM 0240
- US.
GOVE~RNMENT PRINTING OFFICE-' 1979--289-369
o0 UNITED STATES NUCLEAR REGULATORY COMMISSION qo WASHINGTON, D. C. 20555 October 16, 1980 Dockets Nos. 50-269, 50-270 and 50-287 Mr. William 0. Parker, Jr.
Vice President - Steam Production Duke Power Company P. 0. Box 2178 422 South Church Street Charlotte, North Carolina 28242
Dear Mr. Parker:
In order to complete our review of your Oconee Reload Design Methodology Technical Report of May 20, 1980, we find that we need additional infor mation. It is possible that you may receive additional requests.
Kindly provide the requested information with three signed originals and 37 additional copies within 30 days of receipt of this letter.
Sincerely, Robert W. Reid, Chief Operating Reactors Branch #4 Division of Licensing
Enclosure:
Request for Additional Information cc w/enclosure:
See next page
Duke Power Company cc w/enclosure(s):
Mr. William L. Porter Duke Power Company P. 0. Box 2178 422 South Church Street Office of Intergovernmental Relations Charlotte, North Carolina 28242 116 West Jones Street Raleigh, North Carolina 27603 Oconee Public Library 201 South Spring Street Walhalla, South Carolina 29691 Honorable James M. Phinney County Supervisor of Oconee County Walhalla, South Carolina 29621 Director, Technical Assessment Division Office of Radiation Programs (AW-459)
U. S. Environmental Protection Agency Crystal Mall #2 Arlington, Virginia 20460 U. S. Environmental Protection Agency Region IV Office ATTN: EIS COORDINATOR 345 Courtland Street, N.E.
Atlanta, Georgia 30308 Mr. Francis Jape U.S. Nuclear Regulatory Commission P. 0. Box 7 Seneca, South Carolina 29678 Mr. Robert B. Borsum Babcock & Wilcox Nuclear Power Generation Division Suite 420, 7735 Old Georgetown Road Bethesda, Maryland 20014 Manager, LIS NUS Corporation 2536 Countryside Boulevard Clearwater, Florida 33515 J. Michael McGarry, III, Esq.
DeBevoise & Liberman 1200 17th Street, N.W.
Washington, D. C. 20036
Enclosure REQUEST FOR ADDITIONAL INFORMATION OCONEE NUCLEAR STATION, RELOAD DESIGN METHODLOGY, REPORT NO. NFS-1001
- 1. Paragraph 3.2.5, Reactivity Coefficients and Deficits.
The described procedure for the calculation of the reactivity deficits involves PDQ07 or EPRI-NODE. However, it is not clear whether for widely different states that the reactivity difference due to the spectral component is also included. The same comment applies to the differential boron worth calculation.
- 2. Table 3-1, Shutdown Margin Calculation.
Provide a description of the manner in which the "Worth reduction due to burnup of poison material" has been calculated.
- 3. Paragraph 3.2.8, Kinetics Parameters.
Present a more detailed description of the DELAY code. Provide the source of the code, e.g., Duke Power Company.
- 4. Paragraph 8.3.2, Start-Up Accident.
Provide the variation of the total (and its components) reactivity for the start-up accident for the first ten seconds after accident initiation.
(These would complement Figures 14-1 and 14-2 of the Oconee FSAR Rev. 16.)
- 5. Paragraph 8.3.3, Rod Withdrawal Accident at Rated Power Operation.
Give the variation of the reactivity as in 4. above.
- 6. Paragraph 8.3, Discussion of Individual Accidents.
Have the computer codes used in accident analysis (summarized in Appendix A) been updated and revised since the Oconee FSAR was issued? If so, would the general conclusions of the accident analysis change if the anlaysis was to be performed with the updated codes? Justify your conclusion.
- 7. Paragraph 8.3.4, Moderator Dilution Accident.
"Additional Analysis" is claimed to demonstrate complete protection during refueling operations. Provide more information regarding this additional analysis.
- 8. Paragraph 8.3.6, Loss of Coolant Flow.
It is stated that the hot channel power peak augmentation factors, fuel densification, and rod bow effects are not expected to change for the reloads; however, it is not stated how this conclusion has been arrived at.
- 9. Paragraph 8.3.9, Steam Line Failure.
It is stated in the accident description that continued feedwater flow in
-2 the affected steam generator, combined with excessive heat removal and primary cooldown, the reactor may experience "a return to low power levels." There is not quantification of this power level, its potential consequences, or measures and actions for the return of the reactor to subcritical.
Under what conditions is there a minimum of rod worth which could have the most adverse effects?
- 10. Supplement 2, Figure 4-1 and Paragraph 3.1.1.1.
Figure 4-1, Supplement 2, appears to contradict the statement in Para graph 3.1.1.1 that reads:
"NON-fuel cross sections with the exception of burnable poison assemblies and control rods are also generated using EPRI-CELL. Cross sections for burnable poison assemblies and control rods for use in diffusion theory calculations are generated by matching reaction rates between the diffusion theory code PDQ07 and !CPM (a collision probability code)."
Give a more detailed description of the procedure for control rod and burnable poison cross section generation and the use of burnable poison cross sections in PDQ07-HAPONY depletion calculations.
- 11.
Supplement 2, Paragraph 3.2, Comparison of ARMP PDQO07 to Cold Criticals.
The two-dimensional simulation of the criticals has not been performed at Duke nor with PDQ07, yet it was concluded that the results would have been identical with the PDQ07 results. Justify the above conclusion.
- 12. Supplement 2, Paragraph 3.4, Conclusions.
The conclusions for the calculated results of the peak power are not tenable. There is no reason why the diffusion theory estimationby PDQ07 of the local radial peaking should be more conservative than those calculated with transport theory codes, or the measured values. This result must be regarded as fortuituous. For example (Figure 3-4), many fuel assembly maxima were underpredicted by PDQ07. Justify the conclusion that PDQ07 will always be conservative in peak power predictions and present physical arguments for this justification.
- 13. Supplement 2, Paragraph 4.2, Oconee Fuel Cycle Simulation.
It appears that the EPRI-NODE-P almost consistently underpredicts the assembly peak power for Cycles 2 and 3. Justify the conclusion in Paragraph 4.3 that the EPRI-NODE-P "yielded consistently good power distributions...
- 14.
Supplement 2, Figures 4-2 through 4-127.
The EPRI-NODE-P calculated power distributions for the first four cycles of operation of Oconee 1 consistently underpredicted the relative power in Assembly H-8, often by more than 10%. Is the reason for this anomaly known?
-3
- 15. Supplement 2, Paragraph 5.2, Normality Test Results.
All data sets have been used with the assumption of normal distri bution, yet some have failed the normality test. Justify the use of the data sets as normal.