ML20209G723

Forwards Rev 1-P to CEN-347(O)-P Omaha Batch M Reload Fuel Design Rept, & Responses to Cycle 11 Core Reload Methodology Change Questions for Review.Affidavit Also Encl. Rept Withheld (Ref 10CFR2.790)

ML20209G723
Person / Time
Site:	Fort Calhoun
Issue date:	01/26/1987
From:	Andrews R OMAHA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML19292G692	List: ML20209G723
References
LIC-87-042, LIC-87-42, TAC-63937, TAC-63967, NUDOCS 8702050444
Download: ML20209G723 (11)

Text

Omaha Public Power District 1623 Harney Omaha Nebraska 68102-2247 402/536 4000 7 4b LIC-87-042 gg., pg Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Gentlemen:

[

SUBJECT:

Cycle 11 Core Reload Methodology Change Questions

References:

(1) Docket No. 50-285 (2) Letter NRC (D. E. Sells) to OPPD (R. L. Andrews), dated December 22, 1986 (3) Letter OPPD (R. L. Andrews) to NRC (A. C. Thadani), dated November 17, 1986 (LIC-86-537)

(4) Letter OPPD (R. L. Andrews) to NRC (A. C. Thadani), dated December 15, 1986 (LIC-86-677) 7 In accordance with Reference (2), the following responses to your questions have been prepared and are submitted for your review. The information con-l tained in these responses was discussed with Mr. Walter Paulsen, Mr. Larry Kopp l and Mr. Jack Ramsey, of your staff in a telephone conversation on January 12, 1987.

Question 1 Although previous benchmarking of OPPD calculations using current methodology (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 to W. C. Jones,

" Safety Evaluation by the Office of Nuclear Reactor Regulation, Omaha Public Power District, Fort Calhoun Station, Unit No. 1, Docket No. 50-285," May 11, 1984). Therefore, the staff recommended continuation of the ongoing benchmark-ing 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 comment 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.

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I l Document Control Desk January 26, 1987 Page 2 Response 1 The District's ongoing benchmarking program contains verification results from Cycles 9 and 10. The program includes data collected from low power physics predictions and testing, at-power physics predictions and testing, and monthly core follow reports (i.e., power distribution comparisons between measured and predicted values and boron concentration versus burnup). Comparisons continue to show good agreement between ROCS-DIT predictions and measured values. The verification results from Cycles 9 and 10 will be included into the OPPD-NA-8302 update.

Questions 2 Reference is made to "0maha Batch M Reload Fuel Design Report," 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.

Response 2 The mechanical design report was submitted in Reference 4. This report, which was prepared by Combustion Engineering, Inc., was found to contain errors in references to drawings and did not provide an adequate description of the just-ification for the change from 21 kw/ft to 22 kw/ft. CE has revised this report and three non-proprietary copies and three proprietary copies of the revised report, CEN-347(0)-P, Rev. 1-P, are contained herein. Please note that pursuant to 10 CFR 2.790(b)(1), certain portions of the attached information has been deemed trade secrets and/or privileged commercial information by Combustion Engineering, Inc. Accordingly, please find attached an application for with-holding this information from public disclosure.

f Justification for the change from 21 to 22 kw/ft is based upon:

1. The fact that the Fort Calhoun Station 14 x 14 fuel is of the same design as the other CE designed units (e.g., St. Lucie and Calvert Cliffs) also using 14 x 14 fuel; these units currently have a 22 kw/ft limit.

2. The Fort Calhoun Station performance analysis / evaluation for CE and Exxon

( fuel indicates 22 kw/ft to be acceptable.

l Ouestion 3 The enthalpy values given in Section 5.10.2 as fuel and cladding damage thres-holds are not mentioned in Reference 5-5 for the CEA ejection accident as stated. Also, in addition to clad failure, which is used to determine the source term for dose calculations, the NRC limiting criterion for fuel failure, as stated in Reg. Guide 1.77, is that the average fuel pellet enthalpy at the hot spot be less than or equal to 280 calories / gram. The staff feels that this value provide a conservative maximum limit to ensure that prompt rupture of fuel pins and the consequent generation of pressure pulses with resultant core damage l

Document Control Desk January 26, 1987 Page 3 does not occur. 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.

Response 3 The CEA ejection analysis is performed by the fuel vendor. The damage thres-holds reported in the topical report OPPD-NA-8303 coincide with the NRC approved methods used by the fuel vendor. Revision 00 of the topical report referenced the methodology used by Exxon Nuclear Co. who performed the previous (i.e.,

Cycle 6) analysis. With a return to CE as the fuel vendor in Cycle 11 it was necessary to describe the methods used by CE. The threshold values reported are consistent with those last utilized by CE for Fort Calhoun in Cycles 4 and 5.

The reference to the CE CEA Ejection Topical Report CENPD-190-A (Reference 5-5) was incorrect. Reference 5-6 for the Cycle 5 reload submittal will be added.

The CE enthalpy acceptance criteria meets the intent of the 280 cal /gm level as indicated in Reg. Guide 1.77. The 200 cal /gm enthalpy level is indicative of clad failure. The 250 cal /gm is the average fuel centerline threshold which is similar to the 280 cal /gm radial average fuel threshold.

The CE enthalpy thresholds are set for dose consideration, assuming that cool-ability of the fuel will be maintained, whereas the Reg. Guide 1.77 threshold levels are set at clad disruption and fuel melting with primary emphasis on coolability of the fuel. Since the CE enthalpy acceptance criteria are set up for a lower level of fuel degradation and the analysis indicates that these levels are maintained, it is, therefore, concluded that the 200 cal /gm accept-ance criteria is as conservative as the 280 cal /gm requirement in Reg. Guide 1.77.

Question 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 tube plug-ging on other events such as RCS depressurization, main steam line break, four-pump loss of coolant flow, seized rotor, CEA withdrawal, boron dilution, etc.,

and the need for reanalyzing these events because of plugging.

Response 4 All events contained in Chapter 14 of the USAR were either reanalyzed or re-viewed as appropriate. A review constitutes an assessment of the significant parameters of the event, and an evaluation of the effect of the change of the parameters on the event. If it is determined that the event is bounded by the reference cycle analysis, no reanalysis is performed or if a reanalysis is al-ready underway it will be terminated with a negative 10 CFR 50.59 determination noted. For ease of classification, the Fort Calhoun Station USAR events were broken into the categories outlined in Reg. Guide 1.70. A summary of these events are contained below:

Document Control Desk January 26, 1987 Page 4

1. Increase in Heat Removal of Secondary System A. Excess Load (Reanalyzed - for TM/LP pressure bias term)

B. Excess Heat Removal Due to FW Malfunction (Not Reanalyzed)

C. MSLB (Reviewed - for reactivity cooldown curve)

There is a positive effect on these events due to degraded RCS heat removal - therefore, no reanalysis as a result of steam generator tube plugging is necessary.

2. Decrease in Heat Removal of Secondary System A. Loss of Load (Reanalyzed)

B. Loss of FW Flow (Reanalyzed)

There is a negative effect on events due to degraded RCS heat remo. val -

Reanalysis is necessary.

3. Decrease in RCS Flow Rate A. Loss of Coolant Flow (Reanalyzed)

B. Seized Rotor (Reviewed)

Possible increase in heat flux and temperature (negative effect) -

Reanalysis may be necessary. Events examined as indicated.

4. Reactivity and Power Distribution Anomalies A. Boron Dilution - Bounded by CEA Withdrawal B. CEA Withdrawal (Reanalyzed)

C. CEA Drop (Reanalyzed)

D. CEA Ejection (Reanalyzed)

Possible negative effect in approach to DNB limit - All events reanalyzed as indicated.

5. Increase in RCS Inventory A. Boron Dilution (Reanalyzed for dilution to critical time limit).

Events reanalyzed for other event classifications above.

6. Decrease in RCS Inventory A. Steam Generator Tube Rupture (Not Reanalyzed)

B. RCS Depressurization (Reanalyzed for TM/LP pressure bias tern).

No change as heat removal capabilities are maintained during event.

Document Control Desk i

January 26, 1987-

. Page 5 Ouestion 5 Although analyses of the loss of load and loss of feedwater events were per-formed 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.

i Response 5 The loss of load and loss of feedwater events were reanalyzed with 6% steam gen-

! erator plugging. A category for steam generator tube plugging will be added as a category in the methodology topical under key parameters and will include i analysis assumptions. As discussed with your staff on January 12, 1987, the analysis assumptions will be listed as " maximum number of tubes expected to be plugged" rather than a specific value, such as 6%.

h.

) Question 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 genera-tors event is the transient initiated from 2122 psia. Table 5.12.4-1 " Key Par-ameters 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 2074 psia). Please provide justification for using minimum RCS pressure.in pro-posed methodology when current Chapter 14.9.1 analyses state that transient ini-tiated from 2122 psia is the most limiting case.

f

! Response 6 i Per the January 12, 1987 telephone conversation, the responses to questions 6 7 through 9 will be summarized in the question 6 response.

The difference between the methodology submittal and the USAR Chapter 14 is due to the change from Exxon Nuclear Co. methodology (as described in the USAR) to Combustion Engineering methodology. The USAR will be updated after approval of

the reload application which will provide consistency between the methodology topicals and the USAR analyses. All of the changes reflecting the use of CE

, methodology are deemed to be acceptably conservative.

Question 7 i

, Chapter 14.10.1 " Loss of Feedwater Flow" of Fort Calhoun USAR assumes an initial reactor pressure of 2053 psia for analyzing a loss of feedwater event. Table

, 5.13.4.1 " Key Parameters Assumed In The Loss Of Feedwater 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.

Document Control Desk January 26, 1987 Page 6 Response 7 Refer to Response 6.

Question 8 Chapters 14.10.1 and 14.9.1 of Fort Calhoun _USAR analyses both assumed begin-ning-of-cycle (B0C) kinetics parameters. Tables 5.12.4-1 and 5.13.4-1 of the proposed methodology both assume end-of-cycle (E0C) kinetics parameters. Please provide justification for using EOC kinetics parameters when current Chapter 14.10.1 and 14.9.1 analyses assume B0C kinetics parameters.

i Response 8 Refer to Response 6.

Question 9

Chapters 14.10.1 and 10.9.1 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-

, I of the proposed methodology assume a 0.85 Fuel Temperature Coefficient Multi-plier. Please provide justification for using a 0.85 multiplier in the proposed methodology when the current Chapter 14.10.1 and 14.9.1 analyses assume a 0.80 multiplier.

Resoonse 9 Refer to Response 6.

Question 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" Response 10 Typos noted. Change will be made in OPPD-NA-8303, Rev. 01.

Upon completion of the revised methodology topical reports, which reflect the above commitments, we will transmit copies to you.

r Document Control Desk January 26, 1987 Page 7 If you have any further questions or require any additional information, please contact us.

Sin /1

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R. L. Andrews Division Manager Nuclear Production RLA/rh Attachment c: LeBoeuf, Lamb, Leiby & MacRae 1333 New Hampshire Avenue, N.W.

Washington, DC 20036 Mr. A. C. Thadani, Project Director Mr. W. A. Paulson, NRC Project Manager Mr. P. H. Harrell, NRC Senior Resident Inspector l

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AFFIDAVIT PURSUANT TO 10 CFR 2.790 Combustion Engineering, Inc. )

State of Connecticut )

County of Hartford ) SS.:

I, A.E. Scherer, depose and say that I am the Director, Nuclear Licensing, of Combustion Engineering, Inc., duly authorized to make this affidavit, and have reviewed or caused to have reviewed the information which is identified as proprietary and referenced in the paragraph immediately below. I am submitting this affidavit in conformance with the provisions of 10 CFR 2.790 of the Commission's regulations and in conjunction with the application of Omaha Public Power District (0 PPD) for withholding this information.

The information for which proprietary treatment is sought is contained in the following document:

Omaha Batch M Reload Fuel Design Report, CEN 347 (0)-P, Revision 1-P, January, 1987.

This document has been appropriately designated as proprietary.

I have personal knowledge of the criteria and procedures utilized by Combustion Engineering in designating information as a trade secret, privileged or as confidential commercial of financial information.

Pursuant to the provisions of paragraph (b) (4) of Section 2.790 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure, included in the above referenced document, should be withheld.

l

1. The information sought to be withheld from public disclosure concerns the analysis of the Fort Calhoun Batch M fuel assemblies, which is owned and has been held in confidence by Combustion Engineering.

2. The information consists of test data or other similar data concerning a process, method or component, the application of which results in substantial competitive advantage to Combustion Engineering.

3. The information is of a type customarily held in confidence by Combustion Engineering and not customarily disclosed to the public. Combustion Engineering has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. the details of the aforementioned system were provided to the Nuclear Regulatory Commission via letter DP-537 from F. M. Stern to Frank Schroeder dated December 2, 1974. This system was applied in determining that the subject document herein are proprietary.

4. The information is being transmitted to the Commission in confidence under the provisions of 10 CFR 2.790 with the understanding that it is to be received in confidence by the Commission.

5. The information, to the best of my knowledge and belief, is not available  !

in public sources, and any disclosure to third parties has been made pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence.

6. Public disclosure of the information is likely to cause substantial harm to the competitive position of Combustion Engineering because:

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a. A similar product is manufactured and sold by major pressurized water reactor competitors of Combustion Engineering.

b. Development of this information by C-E required thousands of man-hours of effort and tens of thousands of dollars. To the best of my knowledge and belief'a competitor would have to undergo similar expense in generating equivalent information.

c. In order to acquire such information, a competitor would also i

require considerable time and inconvenience related to the design and analysis of comparable fuel assemblies.

d. The information required significant effort and expense to obtain the licensing approvals necessary for application of the information.

Avoidance of this expense would decrease a competitor's cost in applying the information and marketing the product to which the information is applicable.

e. The information consists of fuel rod design and associated analysis, the applicaticn of which provides a competitive economic advantage. The availability of such information to competitors would enable them to modify their product to better compete with Combustion Engineering, take marketing or other actions to improve their product's position or impair the position of Combustion Engineering's product, and avoid developing similar data and analyses in support of their processes, methods or apparatus,

f. In pricing Combustion Engineering's products and services, significant research, development, engineering, analytical, manufacturing, licensing, quality assurance and other costs and expenses must be includea.

The ability of Combustion Engineering's competitors to utilize such

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information without similar expenditure of resources may enable them to sell at prices reflecting significantly lower costs.

g. Use of the information by competitors in the international marketplace would increase their ability to market nuclear steam supply systems by reducing the costs associated with their technology development.

In addition, disclosure would have an adverse economic impact on Combustion Engineering's potential for obtaining or maintaining foreign licensees.

Further the deponent sayeth not.

14 ' W A. E. Jefierer Director Nuclear Licensing Sworn to# before me this 06 'Ifay of putttcut , / 9 y>7,

.Notary Anamm Public e AnlSL s

$USANNE SMIT 5!, NOTARY PUBUC State of Connecticut No. 74143 Commission Exp!res March 31,1990 l

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