ML20087N272
| ML20087N272 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 03/28/1984 |
| From: | William Jones OMAHA PUBLIC POWER DISTRICT |
| To: | John Miller Office of Nuclear Reactor Regulation |
| References | |
| LIC-84-087, LIC-84-87, NUDOCS 8404030287 | |
| Download: ML20087N272 (8) | |
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Omaha Public Power District 1623 Harney Omaha Nebraska 68102 402/536 4000 March 28, 1984 L1C-84-087 Mr. James R.
Miller, Chief U. S. Nuclear Regulatory Commission l
Office of Nuclear Reactor Regulation Division of Licensing Operating Reactors Branch No. 3 Washington, D.C.
20555 l
Reference:
Docke t No. 50-285
Dear Mr. Miller:
Fort Calhoun Station Cycle 9 Core Reload Submittal Attached is Omaha Public Power District's response to the seven (7) questions concerning the Fort Calhoun Station's Cycle 9 core reload submittal.
This information was requested by your staff to enable them to continue their review of the Cycle 9 reload faci-lity license change application.
Questions 1 through 6 were de-lineated in a letter from your office dated March 12, 1984.
Question 7 was received later in a telephone conversation between members of your staff and District personnel.
Sincerely,
!.~;
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W.
C.. Jones Division Manager Production Operations WCJ/JCB:jmm Attachment cc:
LeBoeuf, Lamb, Leiby & MacRae 1333 New Hampshire Avenue, N.W.
Washington, D.C.
20036 Mr.
E.
G.
Tourigny, Project Manager Mr.
L.
A. Yandell, Senior Resident Inspector 8404030287 840328 PDR ADOCK 05000205 f,II" P
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45.5124 Employment h qual Opportunity
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Attachment
- NRC Question 1:
What are the maximum expected peaking factors (FR and Fxy) i during Cycle 9?
t
Response
The maximum expected untilted values of FR and Fxy for Cycle 9 are:
1.63 FR
=
1.68 Fxy
=
NRC Question 2:
Is the B0C, HZP moderator temperature coefficient of +0.86 X 10-9 ap/*F given in Table 5-1 correct? If so, the Technical Specification value for power levels below 80 percent is violated and the values used in the safety analyses are non-conservative.
Response
The value is not correct. The BOC, HZP MTC reported did not include the DIT bias of -0.3 X 10-4 Ap/*F. The value which should have been reported is +0.56 X 10-4 Ap/*F.
Because this value still exceeds the Technical Specifica-tion limit of +0.50 X 10-9 Ap/*F, the District will use rod insertions to lower the critical boron concentration and thus the MTC, thereby insuring that the Technical Spec-ification li t is not exceeded.
. NRC' Question 3:
Since SCU procedures were used for DNBR and CTM calcula-tions, why was the nominal reactor coolant flow value given in Technical Specification.2.10.4(5) not used as input to the safety analyses? Explain how the value of 208,280 gpm was obtained.
i
Response
The best estimate RCS flow rate is used as input to the safety analysis of events in which the uncertainties are combined statistically as documented in Part III, Appendix C, of CEN-257(0)-P (Reference 1).
The Technical Specifica-tion LCO value of 202,500 gpm represents the minimum guar-anteed RCS flow rate including uncertainties and not the nominal flow rate. This LC0 flow rate less uncertainties is used as input to the safety analysis for events in which the uncertainties are combined deterministically.
The best estimate RCS flow rate is the 208,280 gpm value
.and was determined from a Cycle 3 through Mid-Cycle 8 data base of surveillance test measurements of the RCS flow rate.
NRC Question 4:
Please clarify the explanation of the changes made to Tech-nical Specification 2.10.4(5)(a)iii given in Table B-1 (change Number 18).
' Response:
Change Number 18 reflects a change in the method of appli-cation of uncertainties.. The Cycle 8 Technical Specifica-tion limit on RCS flow rate was 197,000 - gpm, which was an actual ' limit (not including uncertainties). The surveil-lance test limit used to monitor this LCO included the un-certainties. - An RCS ficw rate limit of 202,500 gpm was
' 'o
- used in this test.
In the proposed Cycle 9 Technical Spec-ifications, the flow measurement uncertainty has been stat-istically combined with other measurement uncertainties ar.d is not treated independently as was done in previous cycles. Therefore, the allowance for measurement uncer-tainty on the RCS flow rate should now be included in the Technical Specification LCO. The proposed value of 202,500 gpm is consistent with parameters in Technical Specification 2.10.4(5)(a) because all limits are now in-dicated values.
NRC Question 5:
What is the calculated net available scram worth at full power conditions including stuck CF.A allowance, PDIL allow-ance, and physics uncertainty and bias?
Response
The values of scram worths including the stuck CEA allow-ance, PDIL, and physics uncertainties are summarized be-low:
Reactivity (%Ap)
BOC EOC Worth of all CEA's less Worth of Most Reactive CEA Stuck Out 7.85 8.32 PDIL CEA Worth
-0.18
-0.26 Physics Uncertainty Plus Bias
-0.76
-0.80 Moderator Void Collapse Allowance 0.00
-0.10 Thennal Hydraulic Axial Gradient Allowance
-0.20
-0.40 TOTAL 6.71 6.76 NRC Question 6:
Please confinn that the analytical methods (e.g., creep collapse) used in the Cycle 9 or reference cycle fuel sys-tem design (1) have been reviewed and approved by the staff, (2) have been applied for the range of expected Cycle 9 conditions (e.g., burnup), and (3) consider both (vendor) fuel types in the Cycle 9 core.
Response
The CE fuel was analyzed using the methods used for the Cycle 5 analysis and reviewed and approved in Reference 2.
Mechanical design analysis, including creep collapse, has been performed on CE fuel in the Fort Calhoun reactor for an assembly burnup greater than 45,000 MWD /MTU. This anal-ysis adequately bounds the expected E0C 9 exposure for a CE fuel assembly of less than 40,000 MWD /MTU.
The ENC batch H and I fuel was analyzed using the methods discussed in XN-NF-79-70, Reference 3, which was submitted as part of the Cycle 6 analysis and reviewed and approved in Reference 4.
The mechanical design analysis contained
in XN-NF-79-70 was perfomed on ENC Batch H and I fuel in the Fort Calhoun reactor for an assedly burnup up to 40,000 MWD /MTV. This analysis adequately bounds the expected E0C 9 exposure of a Batch H or I fuel assedly of less than 40,000 MWD /MTU.
NRC Question 7:
For those events in wnicn a lower (RCS) flow rate results in a more conservative analysis, justify the use of 208,280 gpm rather than the Technical Specification value of 202,500 gpm.
Response:-
The CEA Drop and Loss of Coolant Flow events analyzed for Cycle 9 are the two events which fall into the category described in the question.
The analysis of the CEA Drop and Loss of Coolant Flow events for Cycle 9 is consistent with the methods des-cribed in the Statistical Combination of Uncertainties topical (CEN-257(0)-P).
In the topical (Appendix C of Part 3), a R0PM function is derived to account for the difference in R0PM's of the base coolant conditions and the worst coolant conditions. The base coolant conditions are represented by a core power of 100%, an inlet temper-ature of 545'F, a pressurizer pressure of 2075 psia, an FR of 1.75, and an RCS flow rate of 208,280 gpm. A sensitiv-ity study was performed on each of the aforementioned par-ameters for both the CEA Drop and Loss of Flow events by varying the parameter of interest, including RCS flow rate, from the base condition over the uncertainty range.
The largest value of R0PM (i.e., conservative) resulting from the sensitivity cases for any one of the parameters was denoted the worst condition for that parameter (The worst condition for RCS flow was a low flow corresponding to the lower bound 95/95 confidence interval). The worst conditions for all of the parameters were combined result-ing in the worst coolant conditions. The worst coolant canditions thus maximize the value of the R0PM. The di f-ference in R0PM between the worst coolant canditions case l
and base coolant conditions case represent tce AROPM func-tion and accounts for the combination of all parameter un-certainties. The AROPM function translates the parameter uncertainties into an equivalent AROPM. The AROPM func-i l
tion includes an accounting for the flow uncertainty at i
the 95/95 confidence interval. Since the parameter uncer-l tainties are invariant from fuel cycle to fuel cycle, the A R0PM function is applicable to any future cycle after its derivation.
The transient analysis for the SCU program included a com-parison between the SCU derived worst coolant condition R0PM and an R0PM derived using the detenninistic combina-tion of the uncertainties. The attached Figures 1 and 2 show this comparison and the SCU base coolant conditions R0PM. These figures show that the deterministically cal-culated R0PMs are always bounded by the SCU worst coolant conditions case.
It should also be noted that the SCU base coolant condition R0PMs are often more restrictive (i.e., larger) than the deterministically calculated val ues.
These Cycle 9 CEA drop and loss of coolant flow transient analyses use the best estimate flow of 208,280 and the flow uncertainty is accounted for by the 6ROPM function.
The LCO flow is set such that a deviation from the data base used to calculate the best estimate flow would be de-tected and that a minimum flow is guaranteed for those events analyzed using a deterministic methodology.
Based on Figures 1 and 2, it can be concluded that use of the 208,280 gpm RCS flow rate, for the CEA Drop and Loss of Coolant Flow events for Cycle 9, in conjunction with the SCU methodology, generates ASI dependent R0PM values which are conservative with respect to those which would be generated through the use of deterministic methods utilizing the RCS flow rate Technical Specification limit of 202,500 gpm. The use of the best estimate flow of 208,280 gpm is justified because the flow uncertainty is accounted for in the AR0PM function which is added to the R0PM calculation which used the best estimate flow (i.e.,
base coolant conditions) and because the R0PM for the worst coolant condition case (base coolant condition case plus AROPM function) is conservative with respect to the deterministic method case.
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REFERENCES 1.
CEN-257(0)-P, " Statistical Combination of Uncertainties", November, 1983.
2.
Letter from R. W. Reid (NRC) to T. E. Short (OPPD) dated December 5, 1978.
3.
! XN-NF-79-70, " Generic Mechanical Design Report for Exxon Nuclear Fort Calhoun 14 x 14 Reload Fuel Assenbly", September,1979.
4.
Letter from Robert A. Clark (NRC) to W. C. Jones (OPPD) dated August 15, 1980.
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