ML20138Q172

Safety Evaluation Supporting Amend 92 to License DPR-40

ML20138Q172
Person / Time
Site:	Fort Calhoun
Issue date:	11/29/1985
From:	Office of Nuclear Reactor Regulation
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ML20138Q153	List: ML20138Q149 ML20138Q160 ML20138Q172
References
NUDOCS 8512270078
Download: ML20138Q172 (11)

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8 jog UNITED STATES 8 o NUCLEAR REGULATORY COMMISSION

. E < wAsmNGTON, D. C. 20555

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.92 TO FACILITY OPERATING LICENSE NO. DPR-40 OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION,_ UNIT NO. 1 l DOCKET NO. 50-285

1.0 INTRODUCTION

, By letter dated September 6,1985 (Ref.1), Omaha Public Power District

, e (OPPD) submitted a reload application for Cycle 10 of the Fort Calhoun I reactor. In addition by letter dated June 13, 1985 (Ref. 2) OPPD proposed certain reload methodology changes for Cycle 10 which are discussed below.

L The major review areas for this reload are these methodology changes and i the reduction of MDNBR safety limit from 1.22 to 1.18.

I t The fuel system, nuclear, and thermai-hydraulics design evaluations are

, } presented herein. In addition, those transients for which a reanalysis has been performed are evaluated in the Safety Analyses Section. An evaluation k of the proposed Technical Specification changes is also presented.

I 2.0 METHODOLOGY CHANGES i By letter dated June 13,1985 (Ref. 2), the Omaha Public Power District l (OPPD) submitted reload methodology changes which affect the previously j approved report entitled " Reload Core Analysis Methodology Overview",

1 OPPD-NA-8301-P. These methodology changes are due to the following modifications and fuel system design changes which will be made to the Cycle 10 core of Unit 1 of the Fort Calhoun Station:

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[ a. Part length poison rods will be inserted in peripheral assemblies j to further reduce the flux to the reactor vessel beltline welds.

b. The Thermal Margin / Low Pressure (TM/LP) calculators will be converted from the "early system" type which did not monitor the  :

, axial shape index to the " standard system" type units which l include axial shape dependency.

c. Mini-CECOR/Better Axial Shape Selection System (BASSS) will be installed on the plant com DNB and Linear Heat RateLHR) (puterLimiting for incore monitoring Conditions for of the Operation (LCO).

These changes are incorporated in the revised methodology overview topical report, OPPD-NA-8301-P, Revision 1 (Ref. 3) and are described further in Reference 4.

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i These changes are the subject of a separate review by the staff (Ref. 6) and have been found to be acceptable. However, this review' concludes that ,

mini-CECOR must be validated prior to its use at the site. In Reference 7, OPPD submitted their procedure for validating mini-CECOR and proposed to complete this validation and notify NRC of the results by December 1,1985.

The proposed schedule and procedures for validating mini-CECOR are acceptable to the staff. But the mini-CECOR/BASSS system must not be utilized until this validation is reviewed and approved by the NRC.

I 3.0 FUEL SYSTEMS DESIGN The mechanical design for the Batch L reload fuel is identical to that of the Batch K fuel described in the Cycle 9 reload submittal and reviewed and

, { approved in Reference 8. The fuel system design and analysis for ENC fuel l

in the Fort Calhoun reactor is described in Reference 9.

! The ENC Batch H and I assemblies were analyzed using the methods presented in-XN-NF-79-70 which were reviewed and approved for the Fort Calhoun Cycle 6 analysis. The mechanical design analysis contained in XN-NF-79-70 was performed on ENC Batch H and I fuel in the Fort Calhoun reactor for an assembly burnup of 40,000 MWD /MTU. This analysis, therefore, adequately i bounds the expected E0C 10 exposure.

I 4.0 NUCLEAR DESIGN l

4.1 Core Characteristics f

i The Cycle 10 fuel management uses a low-radial leakage design, with twice, thrice and fourth burned assemblies predominately loaded on the periphery of the core. This low-radial leakage fuel pattern is utilized to minimize the flux to the pressure vessel welds. In addition, selected assemblies located adjacent to critical welds will contain part-length poison rods L

to further shield the welds. While this type of fuel management results in reduced pressure vessel flux over a standard out-in-in pattern, the

radial peaking factors are increased.

g The Cycle 10 loading pattern incorporates 44 fresh Batch L assemblies with

[ an enrichment of 3.80 w/o. Twelve 4-cycle burned Batch G assemblies and f 9 thrice burned Batch H assemblies, all of which were removed at EOC8, are combined with 32 once burned Batch K assemblies 8 once burned Batch J assemblies, and 28 twice burned Batch J assemblies to produce a Cycle 10 l pattern with a cycle energy of 12,500

• 500 MWD /T. The Cycle 10 core characteristics have been examined for a Cycle 9 tennination between 12,500 MWD /T and 13,500 MWD /T and limiting values established for the safety

analysis. The loading pattern is valid for any Cycle 9 endpoint between tnese values.

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4.2 Moderator Temperature Coefficients ~

The Technical Specifications require that Le mo cc, efficient {gTC) be less positive than Eap/cr .5x10 andgerator less negative temperature than -2.7x10 ap/*F at all times during Cycle 10.

Calculations (+.25to-2.41)haveshowntuttheselimitsaremetforall operating conditions. Since acceptable methods have been used and

appropriate values incorporated in the safety analyses, the range of MTC for Cycle 10 is acceptable, i'

4.3 Power Distributions 1 -

{- Hot full power (HFP) fuel assembly relative power densities calculated for

, t beginning-of-cycle (BOC), middle-of-cycle (M0C), and end-of-cycle (E0C)

I conditions show that the maximum expected peaking factors for Cycle 10 are within the proposed Technical Specification limits for F rand F xy of 1.80 i and 1.85, respectively, including uncertainties and an allowance for azimuthal tilt. Comparisons of the radial peaks given in the calculated i power distributions with the allowable values shown in the Technical

. ) Specifications demonstrate the adequacy of the results given in the safety i analyses. The power distribution measurement uncertainties applied in

Cycle 10 are consistent with the values approved in our review of j CENPD-153-P (Ref. 5), and are, therefore, acceptable for Cycle 10.

I 4.4 Control Requirements i

The value of the required shutdown margin is determined by the E0C steamline break analysis occurring at HZP and remains at 4.0%a k/k for Cycle 10. Based on this value of required shutdown margin and on calculated available scram reactivity including a maximum worth stuck CEA and appropriate calculational uncertainties, sufficient excess exists between available and required scram

reactivity for all Cycle 10 operating conditions. These results are derived by approved methoos and incorporate appropriate assumptions and are, therefore, i acceptable.

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5.0 THERMAL-HYDRAULIC DESIGN i

j 5.1 DNBR Analysis The thermal-hydraulic design methodology used by 0 PPD for Cycle 10 reload analysis was previously submitted to the NRC in the Cycle 9 reload application for Fort Calhoun and was approved for OPPD use. This includes the steady-state DNBR analysis using the TORC /CETOP/CE-1 methodology. In addition, the statistical combination of uncertainties (Ref. 10) associated with the thermal-hydraulic analysis has been reviewed and approved (Ref. 11). Using this methodology, the engineering hot channel

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. factors for heat flux, heat input, rod pitch and clad diameter are combined statistically with other uncertainty factors to arrive at an equivalent DNBR minimum limit of 1.18. This limit ensures with at least 95%7robability and at least a 95% confidence level (95/95 probability / confidence) that the limiting fuel pin will avoid departure from nucleate boiling-if the predicted MDNBR is not below the 1.18 limit.

t The minimum acceptable DNBR limit has been decreased from 1.22 to 1.18 for Cycle 10. This reduction is consistant with the staff's approval of a 1.15 CE-1 correlation limit (rather than the 1.19 interim value) (Ref.14) and the i continued employment of SCU initiated during Cycle 9. Therefore, the 1.18 DNBR limit is acceptable.

5.2 Fuel Rod Bowing i

f The fuel rod bow penalty accounts for the adverse impact on MDNBR of random l variations in spacing between fuel rods. The methodology for detennining l [ rod bow penalties for Fort Calhoun was based on the NRC approved methods presented in CE topical reports on fuel and poison rod bowing (Ref.13).

l The penalty at 40,000 MWD /MTU burnup is 0.5% in MDNBR. This penalty is

applied directly to the new MDNBR design limit. Thus, the new 1.18 MDNBR i limit contains an allowance for a 0.5% rod bow penalty as well as allowances l for uncertainty in the CHF correlation and system parameters and uncertainty

in the TORC code prediction.

.  ! 6.0 SAFETY ANALYSES t

OPPD has reviewed the parameters which influence the results of the transient i require a and accident reanalysis. Theanalyses for Cycle review entailed 10 to determine a comparison betweenwhich, currentif any,(Cycle 10)

[ values of key safety parameters and their bounding values. For those

! current cycle values of key parameters for a particular event which were

! Lonservatively bounded by the reference cycle values, no reanalysis was l required and the results and conclusions quoted in the reference cycle r.nalysis are valid for Cycle 10.

Table 1.0 lists the design basis events which were reanalyzed for i Fort Glhedn, Cycle 10 and the results of these analyses are presented.

The staff has reviewed these results and finds them acceptable for operation of Cycle 10.

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Itshouldbenotedthatthelicensee(Ref.1)hasstatedthatthesite boundary dose for the seized rotor event is well within 10CFR100 limits with 11% failed fuel based on MDNBR values using approved methods.

Current SRP criteria for this event require that the dose for the event is a small fraction of 10 CFR 100 limits. However, in Reference 12 the licensee has stated that if less than 1% of fuel rods in the core fail I radiological releases are a small fraction of the 10CFR100 guidelines for i

the Ft. Calhoun site. Therefore, this event _. sets current staff criteria.

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1 7.0 TECHNICAL SPECIFICATION CHANGES  !

I i The licensee has proposed a number of changes to the Technidl Specifications for Cycle 10 which are listed in Table 3. The staff has revi_ewed these changes and has found that they are properly incorporated in the supporting physics and safety analyses for Cycle 10 using approved methods and are acceptable.

8.0 ENVIRONMENTAL CONSIDERATION

The amendment involves a change in the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.

The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase

! in individual or cumulati 2 occupational radiation exposure. The 1

Connission has previously published a proposed finding that the amendment involves no significant hazards consideration and there has been no public l connent on such finding. Accordingly, the amendment meets the eligibility r criteria for categorical exclusion set forth in 10 CFR 551.22(c)(9).

t Pursuant to 10 CFR 551.22(b), no environmental impact statement or environmental L .v" - " assessment need be prepared in connection with the issuance of the amendment.

9.0 CONCLUSION

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! The staff has reviewed the information presented in the Fort Calhoun

Cycle 10 reload report and in OPPD responses to our requests for additional infonnation. We find the proposed reload and the associated modified Technical Specifications acceptable. However, the mini-CECOR/BASSS system must not be utilized until it is validated and this validation is

, .. reviewed and approved by the staff.

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^ We have concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the f

. p(ublic will not be endangered by operation in the proposed manner, and2) su regulations, and the issuance of the amendment will not be inimical to the connon defense and security or to the health and safety of the public.

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i Principal Contributors l G. Schwenk E. Tourigny ~

Date: November 29, 1985 7......_ .. . . . _ _ _ . . _ _ _ . , . . _ , _ _ _ _ . _ _ _ _ , , , _

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REFERENCES

1. Letter, LeBoeuf, Lamb, Leiby & MacRae to H. R. Denton, dated-September 6, 1985 _

2. Letter from R. L. Andrews (OPPD) to E. J. Butcher (NRC), " Core Reload Methodology Changes for Cycle 10", LIC-85-237, dated June 13, 1985.

3. " Reload Core Analysis Methodology Overview", OPPD-NA-8301-P, j Rev. 01, Omaha Public Power District, June 1985.

4. "CE Setpoint Methodology", CENPD-199-P, Rev. 1-9, Combustion Engineering, March 1982.

' 5. " Evaluation of Uncertainty in Nuclear Power Peaking Measured by the Self-Powered, Fixed In-Core Detector System," CENPD-153-P, Rev. 1-P, May 1980.

j 5 6. Letter, E.J. Butcher (NRC) to R. L. Andrews (OPPD) " Core Reload Methodology Changes for Cycle 10", dated August 26, 1985.

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7. Letter, R. L. Andrews (OPPD) to E. J. Butcher (NRC), " Schedule for CECOR Program Validation", dated October 4,1985

! 8. Memo, L. S. Rubenstein to G. Lainas, " Fort Calhoun Cycle 9 Reload

, [ Analysis", dated April 20, 1984.

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9. " Generic Mechanical Design Report for Exxon Nuclear Fort Calhoun l 14x14 Reload Fuel Assembly", XN-NF-79-70-P, September 1979.

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10. " Statistical Combination of Uncertainties, Part 1, 2, and 3,",

CEN-257(0)-P, November 1983

11. Memorandum for G. C. Lainas from L. S. Rubenstein, "SER for Fort Calhoun SCU", dated March 27, 1984,

12. Letter, W. C. Jones (OPPD) to R. A. Clark (NRC), " Fort Calhoun i

Station Cycle 8 Reload Amendment Application," December 29, 1982.

13. " Fuel and Poison Rod Bowing," CENPD-225-P, October 1976.

14. Memo. L. S. Rubenstein to D..M. Crutchfield, " Review of CENPD-207, October 23, 1984.

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1 TABLE 1.0 DESIGN BASIS EVENTS REANALYZED TQR f0RT CALHOUN CYCLE 10 l Reason for Acceptance Summary

i. Event Reanalysis Criterion of Results (changes relative l to reference cycle) t' l Baron Dilution Increased critical boron Dilution to critical Acceptance criteria i concentration from Cycle time limits of 30 minutes met. See Table 2.

I 9. for refueling and 15

l. minutes for all other j* I subcritical modes must be met.

Excess Load Change in TM/LP trip func- -

Pb 'as = 35.5 psia tion (P wh'ch is more limiting Re-evalldfe)P trip equation.

term. (as in Cycle 9) than bias the RCS Depressuriz-l [' ation.

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RCS Depressurization Re-evaluate Phias term'. -

Pbias = 20.7 psia which is less limiting than that of Excess

.. ., _ , , Load event.

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[. Suquential CEA Group Withdrawal Increased Tech. Spec. Minimum DNBR greater MDNBR = 1.39

, limits on radial peaking than 1.18 using CE-1 PLHGR < 21! kw/ft j . factors. correction. Transient j

! , PLHGR < 2T kw/ft I i

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DESIGN BASIS EVEN[ REANALYZED FOR FORT CALHOUN CYCLE 10 1

Reason for Acceptance Summary Event Reanalysis Criterion of Results l

[ (changes relative j- to reference cycle)

] Loss of Coolant Flow Increased Tech. Spec. Minimum DNBR greater Minimum DNBR = 1.50 l limits on radial peaking than 1.18 using CE-1 g factors, correlation.

Full Length CEA Drop Increased Tech. Spec. Mininaum DNBR greater f limits on radial peaking than 1.18 using CE-1 Minimum DNBR = 1.47 9

, g factors. correlation.

I i l Seized Rotor Increased Tech. Spec limits Stte boundary dose a small Site boundary dose on radial . peaking factors. '

acceptable. Less frqctigns 9f.;10CFR100 -

i g . limits .; - - -

. than 1% failed fuel i ,

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FORT CALHOUN CYCLE 10 RESULTS OF THE BORON DILUTION EVENT

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Criterion for Minimum Time to Lose Time to Lose Prescribed Shutdown Prescribed Shutdown

Mode Margin (Min) Margin (Min) i I Cycle 9 Cycle 10 t

, Hot Standby 92.7 93.8 15 h

r Hot Shutdown 45.2 45.8 15 L

Cold Shutdown - Normal .

$ RCS Volume 39,3; 38.2 15 L Cold Shutdown - Minimum ,

RCS Volume 16.4 18.2 15 l

J. , Refueling 35.0 31.2 30 t

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, TABLE 3.0 .

.) Explanation for Cycle 10 Technical Specification Changes 1

l Tech. Spec. Number Changes keasons  !

, 1. 1.1,1.3(2),1.3(4), Change the minimum The CE-1 limit for 14x14

? 1.3(8),1.3(9),2.1.1 DNCR value from 1.22 fuel was approved by the Pg. 1-2, 1-7, 1-8, to J.18 NRC at 1.15. The SCU 1 1-9, 2-2b analysis was revised to ,

reflect the approved limit. l

2. 1.1 Change the total un- The unrodded planar radial F Pg. 1-2 rodded planar radial peak is being raised for j peak from 1.78 to 1.85 Cycle 10.

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3. 1.1 Change the unrodded The unrodded integrated Pg. 1-2 integrated radial radial peaking factor is peaking factor from being raised for Cycle 10.

1.73 to 1.80 L 4. Figure 1-1 Replace Figure 1-1 with The TM/LP safety limits with enclosed Figure have been changed to i 1-1 reflect changes in peaking factors and inclusion of

ASI input into the TM/LP

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calculators.

5. Figure 1-3 Replace Figure 1-3 with The TM/LP trip LSSS equation r

enclosed Figure 1-3 has been adjusted to reflect ASI input into the i

TM/LP calculators. (Item 11, Reloati Evaluation).

6. 1.3.(2) - Pg. 1-7 Delete references to The Fort Calhoun License

" less than 4-Pump i's limited to 4-Pump operation. operation.

r 7. 1.3(4) Include axial shape The modified TM/LP Pg. 1-8 index as a thermal- calculators monitor ASI hydraulic parameter.

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8. Table 1-1 Delete references to 3- The Fort Calhoun License

! No. 1 and 2 and 2- Pump Operation is limited to 4-Pump i f Pg. 1-10 operation.

i l 9. Figure 2-6 Replace Figure 2-6 The LHR excore LCO has been t

with Enclosed Figure changed to reflect higher p 2-6 radial peaking factors.

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Tech. Spec. Number Changes Reasons I

10. Figure 2-9 Replace Figure 2-9 The F T and F with enclosed Figure haveUEenchanlTlimits ed to 2-9 reflect higher peaking i factors. The asymmetric

& loading pattern impacted i t

the shape of the F*YT limit line.

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11. 2.10.4(2) Change limit to 51.73 The F T changes have Pg. 2-57a to $1.80 and with F,T been$adetoreflect p 21.73 to with F RT 2I.80 proposed changes to i Tech. Spec. 1.1.

t l 12. 2.10.4(3) Change limit to 51.78 The F T changes have been l Pg. 2-57a to limited to 51.85 and madel5reflectproposed with FRT 21.78 to with changes to Tech. Spec. 1.1.

7 FRT 21.85.

13. 2.10.4(4) Mini-CECOR/BASSS Adds a change in tilt y Pg. 2-57b Tilt Limits when Mini-CECOR/BASSS is being used to monitor Technical Specifications l 14. 2.10.4(5)(a)(1) Change $545'Fto The Cold Leg Temperature Pg. 2-47c 5540 F limits have been changed from 545'F to 540*F.

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15. 2.10.4(5)(a)(iv)

Add *** footnote to The Fort Calhoun" Station Pg. 2-57c ider.tify incore non- has added incore DNB LCO itoring with Better monitoring system in Shape Selection System. addition to the excore LCO.

16. 3.10(6)a DELETE two symmetric Asymmetric fuel loading 3

Pg. 3-36b ' safety channels and pattern for Cycle 10 two symmetric control prevents this combination

channels. from properly detecting azimuthal tilts.

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