ML20092H692
| ML20092H692 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 02/12/1992 |
| From: | Gates W OMAHA PUBLIC POWER DISTRICT |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML19344C357 | List: |
| References | |
| LIC-92-020R, LIC-92-20R, NUDOCS 9202210278 | |
| Download: ML20092H692 (22) | |
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Omaha Public Power District 444 South 10th Street Mall Omaha. Netiraska 68102 2247 402/636 2000 i
february 12, 1992 LIC 92.D20R V. S. Nuclear Regulatory Connission ATIN: Document Control Desk Mail Station PI-137 Washington, DC 20555
References:
1.
Docket No. 50 285 2.
Letter from 0 PPD 27(,1991 (LIC 91-320A) OPPD (W. G W. G. Gates) to NRC (Document Control Desk) dated November 3.
Letter from NRC (D. L. Wigginton) to December 26, 1991 Gentlemen:
SUBJECT:
Additional Information Concerning fort Calhoun Station Cycle 14 Reload Attached are the Omaha Public Power District (OPPD) responses to the eight questions contained in Reference 2.
Also included is the response to an additional question raised at a meeting between OPPD and the NRC on January 13, 1992.
The responses provide the NRC additional information on OPPD's submittal for the proposed Technical Specification change on the negative limit for the Moderator Ternperature Coefficient (MTC) for Operating Cycle 14.
In response to Question 3 of the attachment, OPPD has referenced the report CE-CES-129, Revision 1-P. Twenty three (23) copies of this resort (copy numbers 29-
- 53) bustion Engineering) has determined that CE CES-129 are attached for your information.
Pursuant to l 0 CFR 2.790, ABB-CE (Com Revision 1-P contains proprietary information to be withheld from public disclosure.
Appropriate documentation is attached justifying this determination.
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-4 LIC 92 020R Page 2 if you should have any questions, please contact me.
Sincerely 0l' Division Man /M
. G. Gates ager Nuclear Operations WGG/sel Attachments LeBoeuf Lamb, Leiby & MacRae.
D.L.WIgginton,NRCProjectMan(w/oAttachments)hments) c:
ager, (with Attac S. D. Bloom, NRC Pro ect Engineer (with Attachments)(w/o Attachments)
R.P.Mullikin,RCReinnalAdministrator,RegionIV.entor Resident inspector.-(w/o At R. D. Martin N NRC a
I L.
Atto;hment LIC-92-020R I
Page 1 l
I i
Additional Information Concerning Fort Calhoun Station Cycle 14 Reload Application i
I
- 1. Explain why the beginning-of-cycle, hot sero power (HEP) steam line break accident is specified in section 5.1.1 as the most limiting in determining required shutdown margin.
From: Table y
5-1, both the moderator temperaturn coefficient and the doppler coefficient are most negative at ena-of-cycle (ROC) and, therefore,-the EOC event would appear to result in a
+
larger reactivity insertion with cooldown._.
l The. larger reactivity insertion rate at EOC would provide more
= limiting consequences;for the accident and forms the basis for-the required shutdown margin. However, the BOC conditions provide-the minimum available scram worth in assessing the margin-to the Technical _ Specification shutdown margin limit.
i (which is currently 4.0 %Ap),
This is shown in tho review of
-the scram worths available at-BOC and EOC.
The BOC scram worth is 5.0596 %Ap while the EOC scram worth is 5.9833 %Ap.
The rerponse to Question 2 provides additional discussion of shutdown margin and' scram worth.
- 2. Provide a table of CEA reactivity worths and allowances similar to Table 5-2 for Cycle 14 for EOC HZP conditions.
I The following table is similar to Table 5-2..lt includes the requested.EOC, HZP limiting values of reactivity worths and
. allowances.
The Table also illustrates the difference between
.BOC,_HZP.and EOC, HZP limiting CEA shutdown worths for all events, including the main steam line break accident.
From i
L the table.below, use of the BOC, HZP CEA worth value as the most limiting-value is appropriate-since the BOC, HZP excess -
t shutdown margin is 0.92 %Ap less than the EOC, HZP excess
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AttCchment LIC-92-020R Page 2 FORT CALHOUN UNIT NO.
1, CYCLE 14 LIMITING VALUES OF REACTIVITY WORTHS AND ALLOWANCES FOR llOT ZERO POWER BOC, HZP EOC, HZP 1%dd 14003 1.
Worth of all CEAs Inserted 7.52 8.86 2.
Stuck CEA Allowance 1.17 1.43 3.
Worth of all CEAs Less Worth of Host Reactive CEA Stuck out 6.35 7.43 4.
power Dependent Insertion Limit CEA Worth 1.19 1.33 5.
calculated Scram Worth 5.16 0.10 6.
phywies Uncertainty plus Bias 0.10' O.12' 7.
Het Available Scram Worth 5.06 5.98 8.
Technical Specification Shutdown Margin 4.00 4.00 9.
Margin in Excess of Technical specification shutdown Margin 1.06 1.99 1.96 % of calculated scram worth using Nodal Expansion Method (see Question 3, Reference 2).
Due to the applicability of BOC, HZP conditions for not only the snain steam line break accident but for all other events in determining the Cycle 14 minimum excess shutdown margin, a clarification is. required in the last paragraph of Section 5.1.1, page 21.
The revised paragraph should read as follows:
BOC, EP conditions for all evento ao the most limiting conditionn used in the determination of available shutdown margin for compliance with the Technical Specifications.
The sninimum available _ shutdown margin in 1.06 4Ap with reopect to the Technical Specification limit of 4.0 1Ap. Table 5-2 presents a summary of CEA shutdown wortha and reactivity allowances for Cycle 14.
The Cycle 14 CEA worth values, used in the calculation of minimum scram worth, exceed the minimum value required by Technical Specifications and thus provide an adequate shutdown margin.
AttCchment LIC-92-020R Page 3
- 3. Justify the reduction in the physics uncertainty and bias of the calculated scram worth to 1.96 percent as shown in Table 5-2 and reference all appropriate reports.
Use of upgreded reactor physics codes recessitates the use of uncertainties and biaacs consistent with the application of the new methods. Scram worths calculated using the new methods, with biases and uncertainties applied, provide comparable results to those obtained using the former methods (with biases and uncertainties included).
The NRC-approved reference cycle (i.e.,
Cycle 13) reload submittal utiliced the Higher Order Difference (HOD) method which is described in Reference 1.
For the Cycle 14 submittal, the Nodal Expansion Method (NEM), which is also described in Reference 1, was implemented which increases calculational accuracy of the nuclear design codes.
Specific changes incorporated into the new methods include:
1.
Implementation of NEM into the ROCS code;
- 2. Improvements in accountability of anisotropic scattering and higher order interface current angular distributions in the DIT code;
- 3. Introduction of assembly discontinuity f actors between the ROCS and DIT codes;
- 4. Update of biases and uncertainties applied to calculated parameters.
The revised biases and uncertainties associated with the application of NEF,1 are described in Reference 2.
Introduction of the improved methods required the re-evaluation of the biases and uncertainties.
The Combustion Engineering data base used to establish the biases and uncertainties was expanded to reflect recent reload cycles with low leakage and high burnup fuel management.
The data base was derived from the following sources:
AttCChment LIC-92-020R Page 4 Plant Cycle Humber of IMn}w Palo Verde 1 2
7 Palo Verde 2 2
7 Palo Verde 3 2
7 Palo Verde 1 3
7 Palo Verde 2 3
7 Calvert Cliffs 1 10 5
Calvert Cliffs 2 8
5 Fort Calhoun 13 6
Total Cycles:
8 Total Banks 51 i
In addition, Calvert Cliffs 2, Cycle 9 data was added later and found to be consistent with the above data base.
For calculating the FCS Cycle 14 (N-1) scram worths, the uncertainty plus bias terms used are found in Reference 2, Table C (Item C-2), page C-1.
The use of NEM in ROCS underpredicted scram worth by 4.32%, thus, calculated values must be-increased-by-4.32%.
The uncertainty term for ROCS-NEM scram worths is 6.28% which is applied in the conservative direction.
Since the bias term and the uncertainty term, when taken individually, are applied in different directions, the resultant bias plus uncertainty term is 1.96%.
Using the former ROCS-HOD method, the scram worths were overpredicted by 4%, thus the calculated value must be reduced by 4% to obtain the biased scram worth.
The uncertainty term for the ROCS-HOD method is 9%.
Since both terms must be applied in the same direction, the combined bias plus uncertainty term is 13%.
Therefore, in order for both NEM and HOD to produce similar not scram worths, the NEM combination of bias and uncertainty terms must be smaller in value than the HOD combination of bias and uncertainty terms.
To verify that use of-the HOD method or the NEM method produru similar results, a Cycle 13 scram worth model using ROCS-NEM was generated and compared to the NRC-approved Cycle 13 ROCS-HOD results.
These results are presented below along with results from Cycle 14 using ROCS-NEM:
AttCChment LIC-92-020R Page 5 FORT CALHOUN UNIT NO. 1, CYCLE 14 LIMITING VALUES OF REACTIVITY WORTilS AND ALLOWANCE FOR BOC, HZP ( %6p)
Cycle 13 cycle 13 cycle 14 1110P.1 JNML
.1HIlil.
1.
Worth of all cEAs Inserted 9.23 7.93 7.5?
2.
Stuck cEA Allowance 1.83 1.47 1.17 3.
Worth of all cEAs Less Worth of Most P,eactive cEA stuck Out 7.40 6.46 6.35 4.
Power Dependent Insertion Limit cEA Worth 1.23 1.10 1.15 5.
calculated scram Worth 6.17 5.36 5.16 P
6.
l'hysics Bias plus Uncertainty 0.80*
0.11**
0.10**
7.
11et Available Scram Worth 5.37 5.25 5.06 8.
Technical Specification Shutdown Margin 4.00 4.00 4.00 9.
Margin in Excess of Technical Specification Chutdown Margin 1.37 1.25 1.06 134 of calculated scram worth using Higher Order Difference OIOD) method.
1.964 of calculated scram worth using Nodal Expansion Method (11EM).
The results show the difference between the HOD and NEM methods for calculating the minimum Cycle 13 scram worth to be 0.12 1Ap, which is considered acceptably small.
It can also be dotermined from the above results that ROCS-NEM produced a more conservative net available scram worth than ROCS-HOD.
In summary, the application of the revised physics uncertainty and bias value documented in Table 5-2 of the reload application is based upon the use of ROCS-NEM methods rather than the ROCS-110D method,and both methods are described in Reference 1.
OPPD's application of the ROCS-NEM method in Cycle 14 was consistent wit the same method used in the derivation of the biases an uncertainties of Beference 2.
1 i
AttOchment LIC-92-020R
.Page 6 Referencest
- 1. *The ROCS & D1T Computer Codes for Nuclear Design",
CENPD-266-P-A, April 1983.
]
2.
" Physics Biases and Uncertainties", CE-CES-129, Revision i
1-P, August 1991.
- 4. Explain why there is no change in the maximum radial peaking factor or in the maximum ejected CnA worth between BOC and EOC conditions (Table 5-3).
The most limiting radial peaking factor was calculated (including uncertainties and biases) for the BOC and EOC conditions.
The peaking factors were then raised to a more bounding value and the largest value was transmitted to l
This value was then app?ied in a conservative manner to conditions during a cycle to ensure that the existing and future operating cycles would be bounded by the Westinghouse CEA ejection analysis.
- 5. Discuss in more detail the justification for using the CE fuel rod bow penalty for both the Westinghouse and CE fuel coresident in the e
.n.
The design basis for the amount of fuel rod bow allowed in the Westinghouse fuel and for the CE fuel design is the same.
Westinghouse has identified in the fuel mechanical design report that the-amount of deflection does not require a DNB penalty to be applied under Westinghouse analysis requirements.
Thus, the CE DNB penalty was applied to the Westinghouse fuel to ensure that the OPPD statistical combination of uncertainties was still valid and that conservative input assumptions were used in the analysis,
- 6. Section 6.1 implies that the steady-state DNBR analysis for Cycle 14 differs from that used in previous cycles because of the use of the TORC code rather than the CETOP-D' code.
However, this does not appear to differ from the methodology specified in the previous version (Rev. 3) of OPPD-NA-8301.
Please clarify this point and explain any DNBR methodology difference from the previous cycle in more detail.
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Attochment LIC-92-020R-l Page 7 L
The DNBR analysis applications and methods did not change from j
previous cycles, with the exception that the 70RC computer code was used to calculate the minimum DNBR rather than the
[
CETOP-D computer code.
Both codes are approved for use with the OPPD methods.
The CETOP-D code was developed to run much i
faster on the mainframe computer system than TORC.
With the large number of computer runs required during the reload process, there.was a considerable savings in using CETOP-D in the DNBR analyses.
I Since these calculations are now done on engineering workstations by OPPD, no cost savings are realized in the use of CETOP-D for DNBR work.
Any thermal margin lost in the tuning process for CETOP-D can nowlbe recovered by running l
TORC for DNBR calculations rather than the-CETOP-D.
The OPPD topical report was revised to show the use of TORC rather than
[
CETOP-D for calculating the MDNBR.- All of the other aspects i
for calculating MDNBR_ remain the same-as in previous versions-of the methodology topical report.
1
- 7. Explain why the critical boron concentration with'all rods out assumed in the Cycle 14 boron dilution event during refueling was 1180 ppm whereas the TS minimum refueling boron concentration is 1900 ppm.
Why have the critical _ boron concentration values for the.other modes decreased significantly from the previous cycle values and why have the inverse boron worths remained the same?
j The_1180 ppm value listed was used to determine the minimum boron concentration which in accordance with the Fort Calhoun Station Technical 1 Specifications must include at least a-5
%Ap shutdown margin. In addition the 30 minute dilution to i
critical _ time criterion must=be met.
The current TS value was i
compared to the 1180 ppm value adjusted by 5.0 %Ap and was found to be conservative.-The TS-value is adjusted as necessary for each cycle to ensure that the time to criticality meets'the minimum requirements for operator action. For Cycle 14 no adjustment was required and the margin noted in'the above question exists.
The use of the integral fuel burnable.abourber (IFBA) fuel design caused the large reduction in the critical boron o
- concentration requirements.
Since this is the first l
Westinghouse fuel to be loaded into Fort Calhoun Station it is anticipated that there will continue to be changes in the boron'_ requirements.for future cycles as more of the fuel displacing shims are replaced by IFBA rods.
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Attcchment LIC-92-020R Page 8 The inverse boron worths appear to remain the same since a bounding value is used to provide a limiting analysis for each cycle.
The table below compares the actual and analysis values of the inverse boron worth for Cycle 14.
FORT CALHOUN UNIT NO.
1, CYCLE 14 I!NERSE BORON WORTHS ACTUAL VERSUS ANALYSIS VALUES OPERATING ACTUAL VALUE ANALYSIS VALUE liODE IDDm/%Ap1
,(DraL%dpl Hot Standby (2)
-95.5
-90 Hot Shutdown (3)
-95.5
-55 Cold Shutdown t4e mi na u
-70.7
-55 Cold Shutdown ( 4 e n mum vo n= ) -70.4
-55 Refueling (5)
-79.7
-55
- 8. Since Table 5-2 specifies 5.06 percent as the net available scram worth at-HZP, why was 6.40 percent used in the HZP CEA withdrawal analysis?
The incorrect value was left in the Table from a previous draft.
The correct value, from the analysis document, is 5.048 %Ap for the CEA Withdrawal Analysis at HZP and should replace tne 6.40 1Ap value questioned.
The 5.06 %Ap in Table 5-2 is an input value for the Main Steam Line Break Analysis at liZP conditions.
- 9. In accordance with Appendix A of Standard Review Plan 4.2,
the NRC requires an evaluation of fuel assembly structural integrity considering the lateral effects of seismic and LOCA loads for transition cores consisting of different-fuel types using time history numerical techniques based on the plant specific safe shutdown earthquake (SSR).
Verify that this has been performed for Fort Calhoun Cycle 14 and that the results i
show that-all fuel types are structurally acceptable for the transition core.
The results should show that the grids will not buckle due to combined impact forces of a seismic /LOCA event, the core coolable geometry is maintained, and the stresses resulting from the seismic /LOCA induced deflections are within acceptable limits.
l l
l-
Attachment LIC-92-020R Page 9 In the discussions on January 13, 1992 the IJRC noted that for Cycle 8 there was a requirement in the SER (Reference 1) that "the licensee will be required to provide analytical results to t he !!RC within one year using approved E!JC 'nethodology to comply with fuel asserrbly structural acceptance criteria in Appendix A of SRP-4.2 for the design seismic event.*
lt should be noted that OPPD provided a response to the seismic analysis requirement in Reference 2.
The OPPD response stated:
- The District has reviewed the seirnd e analysis cont airwd in Appendix F of the USAR to determino the tyi.e of analysis performed for the first core.
This type of analysis would tw per f ormed on a core cont aining CE and E!JC or ENO f uel, since it is the licensing basis for the core oeismic analysis.
14ased on t he infotmation cont ained in Appendix F, the District concludes that a dynamic seismic analysis was not per f ezined f or the f uel assenblies.
It is the Dist rict's posit ion t hat an analysis to show compliance with the f uel assently st 2 uctural criteria in Appendix A of SkP-4.2 f or t he design seismic event is out side the scope of the design basis f or the ret t Calhoun station Unit flo. I and that an unreviewed saf ety question does not oxist f or a core of CE and ENO f uel or P!1C f uel wit h respect to the design seismic avent.
Therefore, it is the District's position that such an analysis is not r equir ed.
- Subsequent correspondence, Reference 3, indicated:
- You further stated that such an analysis is not required. We agree."
Thus, the seismic analysis was not required for a mixed core of CE and E!JC fuel.
As described in the fuel mechanical design report, there was a substantial offort made to ensure compatibility between the CE and Westinghouse fuel assembly design parameters.
The elevation of the grids in the CE and Westinghouse fuel assemblies are not matched on centerlines, but there is overlap between the adjacent grids.
TL fuel design also required that the grid crush strength be comparable between CE and Westinghouse fuel assemblies.
The crush strength was based on the LOCA blowdown loads and manufacturing tests by the fuel vendors.
A peripheral grid was found to have some deformation in the previous CE LOCA load analysis which required a coolable geometry study to be done.
In the Westinghouse LOCA analysis, two grids were assumed to fail as a conservative evaluation practice and the impact on coolable geometry and peak clad temperature was also assessed.
AttC;hanent LIC-92-020R Page 10 Both the CE and Westinghouse LOCA evaluations indicated that a coolable geometry was maintained based on !JRC approved acceptance criteria. The assembly and grid stresses were acceptable and the grids of one manufacturer will not crush the grids of the other due to impact loads from a LOCA.
f Ef erencer; j
- 1. Letter, E. G. Tourigny (!!RC ) to W. C. Jones (OPPD) dated March 15, 1983, Facility License Amendment #70.
- 2. Lotter(LIC-83-184)from W. C. Jones (OPPD) to Mr. Robert A. Clark (NRC) dated July 28, 1983.
- 3. Letter, James R. Miller ("RC) to Mr. W. C. Jones (OPPD) dated-August 29, 1983 y
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Omaha Pubtle Power District 1623 Hamey Omaha, Nebraska f> BIO 2 402/536 4000 mM f$h8sia[0' Q>4ESTiod N REFEf!EtJCE. 2.
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Mr. Robert A. Clark, Chief U. S. Nuclear Regulatory Commission OHice of Nuclear Reactor Regulation Division of 1.icensing Operatinq Reactors Branch No. 3 Washington, D.C.
20555
Reference:
Docket No. 50-285
Dear Mr. Clark:
Response to Concerns Contained in the Cycle 8 Reload SER and Cycle 9 Reload Schedule The Cycle 8 reload SER requested the District oddress concerns related to fuel assembly seismic analysis, worst case ECCS assumptions, high burnup fuel, and documentation of the District's reload methodology.
This letter addresses these concerns and also provides a schedule for submittals re-lated to the Cycle 9 reload.
Fuel Assembly Seismic Analysis The District has reviewed the seismic analysis contained in Appendix F of the USAR to determine the type of analysis performed for the first core.
This type of analysis would be performed on a core containing CE and ENC or ENC fuel, since it is the licensing basis for core seismic analysis.
Based on the information contained in Appendix F, the District concludes that a dynamic seismic analysis was not performed for the fuel assemblies.
It is the District's position that an analysis to show that compliance with the fuel assembly structural criteria in Appendix A of SRP-4.2 for the design seismic event is outside the scope of design basis for the Fort Calhoun Station Unit No. I and that an unreviewed safety question does not exist for a core of CE and ENC fuel or ENC fuel with respect to the design seismic event. Therefore, it is the District's position that such an analysis is not required.
Worst Case ECCS Assumptions The SER required that the assumption nf a worst single failure iri the ECCS be reviewed and verified as more limiting than an assumption of no single
- failure, in response to the District's request. Exxon Nuclear Company has performed a sensitivity study of the Fort Calhoun Cycle 8 ECCS analysis to determine which assumption provided the mst restrictive results.
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Hr. Robert A. Clark l
LIC 83184 Page Two j
The worst single failure assumption was the loss of a low pressure safety injection pump. When the ECCS anblysis was. redone, due to the Fort Calhoun ECCS configuration, the estimated full safety injection flow re-
-sulted in a higher reflood rate with no significant ef fect on containment l
pressure. Assuming the single f ailure, the analysis predicted lower re-flood rates and a higher peak cladding temperature.
Therefore, the Fort Calhoun Cycle 8 ECCS analysis provides the most limiting prediction with the assumption of a single failure.
High Burnup Fuel The Cycle 8 reload SER-stated that batch average burnups exceeding 38,000 HWD/MTV in future cycles would involve an unreviewed safety-question re-lated to radiological consequences.
Based on a May 16, 1983 telephone conversation between the Commission and District staffs, it is our under-
-standing that these high burnup concerns will be addressed in the extended burnup topicals submitted by the nuclear fuel vendors.
In response to t
your request made_ during the May 16, 1983 telephone conversation, the anticipated discharge batch burnups for future cycles are provided in Table 1.
i TABLE 1 FORT CALHOUN STATION UNIT NO. 1 ANTICIPATED BATCH DISCHARGE BURNUP (All fuel manufactured by ENC)
Anticipated Batch Average Cycle Shutdown Cycle No. of Discharge Burnup Discharged Date Loaded Assemblies (MWO/MTV) 9 Sept 1985 6
24 34,000 9
Sept 1985 7
19 35.500
-i 10 Mar 1987 7
17 40,000 10
_ Mar 1987 8
23 36,000 11 Sept 1988 8
5 39,000 11 Sept-1988 9
8 44,000 11 Sept 1988 9
31 39,500 Documentation of Reload Methodology t
The Cycle 8-reload SER requested the District submit methodology reports well in advance of the Cycle 9 reload application' date.
The scope of these methodology reports was discussed in a May 11, 1983 telephone con.
versation between members-of the Commission and District staffs. Based on-requests made _by members of the Commission Staf f during these conver-sations the District will submit methodology reports on reactor physics and transient analyses.
In addition the District intends to submit a
+
reload methodology report which will provide an overview of the analyses performed during a reload core analysis and the interfaces between these analyses.
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Mr. Robert A. Clark LIC-83-184 Page Three The District also intends to utilize a Statistical Combination of Uncer-tainties (SCU) in the Cycle 9 reload analysis.
The SCU analysis is being performed by Combustion Enginee -ing utilizing the methodology previously submitted and approved on the f alvert Cliffs and St. Lucie 1 dockets. The l
Schedule for submittals and ersots related to Cycle 9 reload licensing is given in Table 2.
l TABLE 2 CYCLE 9 RELOAD SCHEDULE Event Date Submit Reactor Physics and Transient September 23, 1983 Analysis 14ethodology Reports Submit Statistical Combination of Uncer -
October 21, 1983 taintdes Report Submit Cycle-9-Technical Specifications February 10, 1984 l
Start of Refueling-March 19, 1984 Cycle 9 Startup May 14,1984 The District believes this letter addresses all Commission staff concerns F
discussed in the Cycle 8 reload SER.
The submittal of the methodology re.
i ports will satisfy all requirements discussed in the SER.
Sincerely, W.[
MUf(.2._ s i
Jones Division Manager Production Operations t
WCJ/JKG:jmm cc:
LeBoeuf. Lamb, Leiby & MacRae 1333 New Hampshire Avenue, N.W.
Washington 0.C.
20036 Mr. L. A. Yandell,' Senior Resident inspector.
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9/7/03 - W[ll, GPB, WCJ,1100, KJ!i,%. HGG, TJii, JCB, JKG, DDW, Plis, FAT, TFF, DJii. TRR, RRM,
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NUCLEAR REGULATORY COMMISSION j y (,..
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'7 AUG 2 9 1983 Docket No. 50-285 L
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N D i io Manager Production kEfCO6NCb 3 Om a u Power District 1623 Harney Street Omaha, Nebraska 68102
Dear Mr. Jones:
We have reviewed your letter of July 28, 1983 in which you provided responses to our long range concerns contained in our Cycle 8 reload SER which was issued on March 15, 1983.
We required that if the District intends to use a safety analysis computer code to support reload licensing actions, it should demonstrate its pro-ficiency in using the code by submitting code verification.
This would best be accomplished by submitting methodology reports for our review and approval.
You have provided schedules for submittal of such reports and they are acceptable.
We required the District to provide analytical results to the NRC within one year using the approved ENC methodology to comply with fuel assembly structural acceptance criteria in Appendix A of SRP-4.2 for the design seismic event.
You have stated that an analysis to show that compliance with the fuel assembly structural criteria in Appendix A of SRp-4.2 for the design seismic event is outside the scope of design basis for the Fort Calhoun Station and that an unreviewed safety question does not exist' for a core of CE and ENC fuel or ENC fuel with respect to the design seismic event.
You further stated that such an analysis is not required. We agree, We noted that our Cycle 8 approval applies to the requested discharge average exposure of 37,200 MWO/MTU only and that significant increases in this or future cycles would involve safety questions related to radiological consequences.
You stated that it is your understanding that these high burnup concerns will be addressed in the extended burnup topicals submitted by the nuclear fuel vendors. We have no objection to addressing radiological consequences for high burnup fuel in vendor topicals as long as they apply to Fort Calhoun fuel and are addressed adequately and documented.
We required that for the large break LOCA, you must demonstrate the worst assumption for ECCS operation since it was shown that for some plants using Exxon fuel, maximum safety injection might be the worst case rather than loss of some ECCS capacity as was believed previously, You stated that the Fort Calhoun Cycle 8 ECCS analysis provides the most limiting prediction with the assumption of a single failure.
This is acceptable.
f Mr. W. C. Jones In suninary, your July 28, 1983 letter adequately addressed our long range concerns that were discussed in our Cycle 8 safety evaluation of March 15, 1983.
Sincerely, Chi > m Tawd$
Janes R. Miller, Chief Operating Reactors Branch #3 Division of Licensing cc: See next page
p s
Omaha Public Power District CC:
Harry.H. Voigt, Esq.
LeBoeuf, Lamb, Leiby & MacRae 1333 New Hampshire Avenue, N.W.
j Washington, D. C.
20036 i
Mr. Jack Jensen Chairman, Washington County Board of Supervisors Blair, Nebraska 68023
- U.S. Environmental Protection Agency Region Vl!
ATTN:
Regional Radiation-Representative 324 East 11th Street l
Kan,sas City Missouri 64106 Metropolitan Planning Agency ATTN:- Oagnia Prieditis 7000 West Center Road i
Omaha, Nebraska 68107 Mr. Larry Yandel.1 U.S.N.R.C. Resident inspector P. O. Box 309 Fort Calhoun, Neoraska 68023 Mr. Charles B. Brinkman Manager - Washington Nuclear Operations C-E Power Systems Combustion Engineering, Inc.
7910 Woodmont Avenue Bethesda, Maryland 20814 Regional Administrator Nuclear Regulatory Commission, Region IV Office of Executive Director for Operations 611 Ryan Plaza Drive Suite 1000 Arlington, Texas 76011
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AlILQbyIT PURSUANT TO 10 CPR 2.790 Combustion Engincoring, Inc.
)
l Stato of Connecticut
)
County of Hartford
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SS.:
I, S. A. Toollo, depose and say that I am the Managor,_ Nuclear
' Licensing, of Combustion Engineering, Inc., duly authorized to maho this affidavit, and,have reviewed or caunod to have reviewod tho information which is identified as propriotary and referenced in tho
- paragraph 'immediately below. :
I am submitting this affidavit in conformance with the provisions of 10 CPR 2.790 of the Commission's
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regulations and -in conjunction with Omaha Public Power District for withholding this information.
The information for which propriotary treatment is sought is contained in the following document:
Physics Blaces and CE-CES-129 Rov.
1-P, "Mothodology Manual Uncertainties," 1991.
- This. document han boon appropriately designated as propriotary.
I have. personal - knowledge. of the critoria and procedures utilized by Combustion Engineering in designating information as a
-trado secret,-privileged or as confidential commercial or financial
- information.
Pursuant to the provisions of paragraph (b) (4) of Soction 2.790 l'
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- of the Commission's regulations, the following is furnished for consideration by the Commission in datormining whether the information sought to be withhold from public disclosure, included in the above referenced documents, should be withhold.
1.
Tho informatior. sought to be withhold from public disclosure, which is owned and has boon hold in confidence by combustion Engineering, is physics biaces and uncertaintion applied to PWR nuclear design paramotors.
2.
The information consists of test data or other similar data concerning a procesa, method or component, the application of which results in substantial competitivo advantage to Combustion Engir coring.
3.
The information is of a type customarily hold in confidence by Combustion Engineering and not customarily disclosed to the public.
Combustion Engineering has a
rational basis for datormining the types of information customarily hold in confidence by it and, in that connection, utilizes a system to datormine when and whether-to hold cortain types of information in confidence.
The details of the aforementioned system woro 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 datormining that the subject document heroin is proprietary.
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The-information is being transmitted to the Commission in confidenco under the provisions of 10 CFR 2.790 with the understanding that it is to be rocolved in confidence by the Commission.
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5.
The information, to the best of my knowledge and bollof, is not available in public sources, and any disclosure to third parties has-boon mado pursuant to regulatory provisions or propriotary agreomonts which provide for maintenanco of the information in confidence.
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6.
Public disclosure of the information is likely to cause substantial harm to the competitivo position of Combustion Engincoring becauso
- 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 manhours and hundreds of thousands of dollars.
To tho bosu of my knowledge and boller, a competitor would have to undergo similar expenso in generating equivalent information, c.
In. order _to acquiro.such information, a competitor would als'o require considerable timo and inconvenienco related to the development of similar physien binsos and uncertainties that are applied to PWR nuclear design paramotors.
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d.
Tho-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 cosc in applying the information and marketing the product to which the information is a pp s '.c S hle.
e.
The-information consists of physics biases and i
uncertainties applied to PWR nuclear design parameters, the application of which provides a competitive economic 1
advantage.
The availability of such information to
= competitors would enable them to modify their p % Tuct to better compete with Combustion Engt cering, 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 rosearch, development, engineering, analytical, manuf acturing, licensing, quality assurance and other costs and expenses must'be included.
The ability of combustion Engineering's competitors to utilize such information without similar expenditure of resources may enable them to sell at prices reflecting significantly lower costs.
g.:
Use of-the information by conpetitors in.the international marketplace would increase their ability to market nuclear steam supply systems by reducing the costs associated with
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5-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.
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S. 4.
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.m S.
A. Toolle Manager Nuclear Licensing Sworn tg before me l /
this '7 Vii day of ( -h' O1t IB Al l 1992 D'I lg,.
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_q(i t i y e > ;.s.
I t i 16. p.
'-He'ta ry Pu'qlic S / E #~!
My commission expires:
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