ML20215M169

Safety Evaluation Supporting Amend 109 to License DPR-40

ML20215M169
Person / Time
Site:	Fort Calhoun
Issue date:	05/04/1987
From:	Office of Nuclear Reactor Regulation
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ML20215M141	List: ML20215M137 ML20215M149 ML20215M169
References
NUDOCS 8705130110
Download: ML20215M169 (12)

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT N0.109 TO FACILITY OPERATING LICENSE NO. DPR-40 OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION, UNIT NO. 1 DOCKET NO. 50-285

1. 0 INTRODUCTION By letter dated January 22, 1987 (Ref. 1), Omaha Public Power District (OPPD) submitted an application for an amendment to Facility Operating License No. DPR-40 that would modify the Fort Calhoun Station, Unit No. 1, Technical Specifications to support Cycle 11 operation. In addition, by letter dated January 16, 1987 (Ref. 2), OPPD submitted revised large and small break loss of coolant accident (LOCA) analyses as well as a reanalysis of the control element assembly (CEA) ejection event. These reanalyses were performed by the Cycle 11 fuel vendor, Combustion Engineering, Inc. (CE). The January 16, 1987 submittal was referenced in the licensee's January 22, 1987 amendment application.

The fuel system, nuclear, and thermal-hydraulic design evaluations are presented herein. In addition, those transients and accidents for which a reanalysis has been performed are evaluated in the Safety Analyses Section. An evaluation of the proposed Technical Specification changes is also presented.

2.0 METHODOLOGY CHANGES By letter dated November 17, 1986 (Ref. 3), OPPD submitted reload method-ology changes which affect the previously approved reload core analysis methodology reports (Refs. 4, 5 and 6). These changes are the subject of a separate review by the staff and have been found to be acceptable (Ref. 7). They have been incorporated in the revised methodology topical reports (Refs. 8, 9 and 10).

3.0 FUEL SYSTEM DESIGN The mechanical design for the Batch M reload fuel is slightly different from the Batch L fuel described in the Cycle 10 reload submittal due to the change back to CE as the fuel vendor. It is, nevertheless, similar in design to the Batch G fuel previously supplied by CE for Cycle 5 and is mechanically, thermally, and hydraulically compatible with the Advanced Nuclear Fuels Corp. (ANF) supplied fuel remaining in the core.

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y CE fuel has been analyzed for previous Fort Calhoun cycles using approved methods. Mechanical design analyses, including fuel cladding collapse, irradiation induced dimensional changes, cladding strain and fatigue analysis, maximum fuel rod internal pressure and fuel rod corrosion, have been performed on CE fuel in Fort Calhoun for assembly burnups greater than 45,000 MWD /MTU. Therefore, the expected end-of-cycle (E0C) exposure of well under 20,000 MWD /MTU for a CE fuel assembly in Cycle 11 is adequately bounded.

An extended burnup analysis (Ref. 11) was performed for the Batches J, K and L ANF-supplied fuel and allows for a peak assembly burnup of 43,000 MWD /MTU. It is anticipated that an EOC peak assembly burnup of 44,500 MWD /MTV will be achieved for the Batch K fuel. Therefore, ANF has agreed to revise the mechanical design report for Batches K and L fuel in order to envelop the future batch burnups. The licensee has committed to submit this completed analysis for NRC approval prior to exceeding a peak assembly burnup of 43,000 MWD /MTU in order to demonstrate compliance with the appropriate design criteria at these higher exposures (Ref. 12).

4.0 NUCLEAR DESIGN 4.1 Core Characteristics The Cycle 11 fuel management uses a low radial leakage design with fuel assemblies which have been irradiated over several cycles of operation predominantly loaded on the periphery of the core. This low radial leakage fuel pattern is utilized to minimize the flux to the pressure vessel 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. However, the peaking factors for Cycle 11 are consistent with those in previous cycles in which low radial leakage patterns have been utilized and have been conservatively used in the safety analyses and the reactor trip setpoint analyses.

The Cycle 11 loading pattern incorporates 44 fresh Batch M fuel assemblies with an enrichment of 3.8 weight percent U-235. One Batch H assembly, which was irradiated for three cycles and removed at the end of Cycle 8, is being returned to the core along with 16 previously irradiated Batch J assemblies, 28 previously irradiated Batch K assemblies, and 44 previously irradiated Batch L assemblies to produce a Cycle 11 pattern with an anticipated maximum cycle length of 13,600 500 MWD /MTU. The Cycle 11 core characteristics have been examined for a Cycle 10 termination between 11,500 MWD /MTU

< and 12,500 MWD /MTU and limiting values established for the safety analysis. The loading pattern is valid for any Cycle 10 endpoint between these values.

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4.2 Moderator Temperature Coefficients TheTechnicalSpecificationsrequirethatthemogeratortemperature coefficient (MTC)belgsspositivethan+0.5x10 Ap/ F and less negative than -2.7x10 ap F at all times during Cycle 11.

Calculations have shown that these limits are met for all operating conditions. Since acceptable methods have been used and appropriate values incorporated in the safety analyses, the range of MTC for Cycle 11 is acceptable.

4.3 Power Distributions Hot full power (HFP) fuel assembly relative power densities calculated for beginning-of-cycle (BOC), middle-of-cycle (M0C), and end-of-cycle (E0C) conditions show that the maximum expected peaking factors for Cycle 11 are within the proposed Technical Specification limits for F and F of 1.80 and 1.85, respectively, including uncertainties and aballowaXceforazimuthaltilt. Comparisons of the radial peaks given in the calculated power distributions with the allowable values shown in the Technical Specifications demonstrate the adequacy of the results given in the safety analyses. The power distribution measurement un-certainties applied in Cycle 11 are consistent with the values approved in the staff review of CENPD-153-P (Ref. 13), and are, therefore, acceptable for Cycle 11.

4.4 Control Requirements The vaYue of the required shutdown margin is determined by the ECC steamline break analysis occurring at hot zero power (HZP) and remains at 4.0% ak/k for Cycle 11. 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 11 operating conditions. These results are derived by approved methods and incorporate appropriate assumptions and are, therefore,

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acceptable.

5.0 THERMAL-HYDRAULIC DESIGN 5.1 DNBR Analysis The thermal-hydraulic design methodology used by 0 PPD for Cycle 11 reload analysis was previously approved by the NRC for OPPD use. This includes the steady-state DNBR analysis using the TORC /CETOP/CE-1 methodology. In addition, the statistical combination of uncertainties (Refs. 14 and 15) associated with the thermal-hydraulic analysis has been reviewed and approved by the staff. Using this methodology, the engineering hot channel factors for heat flux, heat input, rod pitch and clad diameter are combined statistically with other uncertainty factors to arrive at an equivalent departure from nucleate boiling

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ratio (DNBR) minimum limit of 1.18. This limit ensures with at least 95% probability 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 minimum DNBR (MDNBR) is not below the 1.18 limit.

5.2 Fuel Rod Bowing The fuel rod bow penalty accounts for the adverse impact on MONBR of random variations in spacing between fuel rods. The methodology for determining 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. 16). The penalty at 45,000 MWD /MTU burnup is 0.5% in MDNBR. This penalty is applied oirectly to the MDNBR design limit. Thus, the 1.18 MDNBR limit contains an allowance for a 0.5% rod bow penalty as well as allowances for uncertainty in the CHF correlation and system parameters and uncertainty in the TORC code prediction.

6.0 SAFETY ANALYSES -

OPPD has reviewed the parameters which influence the results of the transient and accident analyses for Cycle 11 to determine which, if any, require a reanalysis. The review entailed a comparison between current (Cycle 11) values of key safety parameters and their bounding values. For those current cycle values of key parameters for a particular event which were conservatively bounded by the reference cycle values, no reanalysis was a required and the results and conclusions quoted in the reference cycle t analysis are valid for Cycle 11.

In order to justify steam generator tube plugging beyond the current 1%

} limit, CE performed revised large and small break LOCA analyses. The CEA ejection event was also reanalyzed by CE using their methodology to-

. evaluate the effect of Cycle 11 data for ejected CEA worth and post ejected radial peaks in addition to the assumption of 6% steam generator plugging.

s Table 1.0 lists the design basis events which were reanalyzed for Fort Calhoun, Cycle 11 and the results of these analyses are presented. The staff has reviewed these results and finds them acceptable for Cycle 11 operation.

It should be noted that the licensee (Ref. 1) has stated that the site boundary dose for the seized rotor event is well within 10 CFR Part 100 limits with <1% failed fuel based on MDNBR values using approved methods. Current Standard Review Plan (SRP) criteria for this event require that the dose for the event is a small fraction of 10 CFR Part 100 limits. However, in Reference 17 the licensee has stated that, if less than 1% of fuel rods in the core fail, radiological releases are a small fraction of the 10 CFR Part 100 guidelines for the Fort Calhoun site. Therefore, this event meets current staff criteria. -

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7.0 TECHNICAL SPECIFICATION CHANGES The licensee has proposed a number of changes to the Technical Specifi-cations for Cycle 11 which are listed in Table 3.0. The staff has reviewed these changes and has found that they are properly incorporated in the supporting physics and safety analyses for Cycle 11 using approved methods and are acceptable.

8.0

SUMMARY

The staff has reviewed the information presented in the Fort Calhoun Cycle 11 reload report and in OPPD responses to requests for additional information.

The staff finds the proposed reload and the associated modified Technical Specifications acceptable.

However, it is anticipated that the ANF-supplied Batch K fuel assemblies will reach an E0C burnup of 44,500 MWD /MTU. Since the present extended burnup analysis for ANF fuel allows for a peak assembly burnup of 43,000 MWD /MTU, the licensee has committed to submit a revised fuel mechanical design report for NRC approval prior to exceeding a peak assembly burnup of 43,000 MWD /MTU in order to demonstrate compliance with the appropriate fuel design criteria required by SRP Section 4.2 at these higher exposures. This report should be submitted in a timely manner in order to allow sufficient time for NRC staff review prior to reaching 43,000 MWD /MTU in any assembly.

9.0 ENVIRONMENTAL CONSIDERATION

This 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 cumulative occupational radiation exposure. The

- Commission has previously published a proposed finding that the amendment

- involves no significant hazards consideration and there has been no public comment on such finding. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR S51.22(c)(9).

Pursuant to 10 CFR SSI.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

10.0 CONCLUSION

The staff has concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations, and the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Date: May 4,1987 .

Principal Contributor: L. Kopp

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11.0 REFERENCES

1. Letter from R. L. Andrews (0 PPD) to USNRC, dated January 22, 1987.

2. Letter from R. L. Andrews (0 PPD) to USNRC, "CEA Ejection and LOCA Analyses Submittal," January 16, 1987.

3. Letter from R. L. Andrews (OPPD) to A. C. Thadani (NRC), " Core Reload Methodology Changes for Cycle 11," November 17, 1986,

4. "0 PPD Nuclear Analysis Reload Core Analysis Methodology Overview,"

OPPD-NA-8301-P, Rev. 01, June 1985.

5. "0 PPD Nuclear Analysis Reload Core Analysis Methodology, Neutronics Design Methods and Verification," 0 PPD-NA-8302-P, September 1983.

6. "0PDD Nuclear Analysis Reload Core Analysis Methodology Transient and Accident Analysis Methods and Verification," 0 PPD-NA-8303-P, September 1983.

7. Letter from W. A. Paulson (NRC) to R. L. Andrews (0 PPD), " Core Reload Methodology Changes for Cycle 11," April 3, 1987.

8. "0 PPD Nuclear Analysis Reload Core Analysis Methodology Overview,"

0 PPD-NA-8301-P, Rev. 02, November 1986.

9. "0 PPD Nuclear Analysis Reload Core Analysis Methodology, Neutronics Design Methods and Verification," OPPD-NA-8302-P, Rev. 01, November 1986.

10. "0 PPD Nuclear Analysis Reload Core Analysis Methodology, Transient and Accident Analysis Methods and Verification," 0 PPD-NA-8303-P, Rev. 01, November 1986.

11. " Fort Calhoun Design Report Extended Burnup Analysis," XN-NF-82-61, October 1982.

12. Letter from R. L. Andrews (OPPD) to USNRC, Cycle 11 Reload Evaluation, LIC-87-119, dated February 24, 1987.

13. " 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.

14. " Statistical Combination of Uncertainties, Part 1, 2, and 3" CEN-257(0)-P, November 1983.

, 15. " Statistical Combination of Uncertainties Methodology for Fort Calhoun," CEN-257(0)-P, Supplement 1-P, August 1985.

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

17. Letter from'W. C. ' Jones'(0 PPD) to'R. A. Clark (NRC), " Fort Calhoun Station Cycle 8 Reload Amendment Application," December 29, 1982.

Attachments:

1) Table 1 - Design Basis Event Reanalyzed for Fort Calhoun Cycle 11

2) Table 2 - Results of the Boron Dilution Event for Fort Calhoun Cycle 11

3) Table 3 - Explanation for Cycle 11 Technical Specification Changes i,

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. TABIE 1.0

! l DESIG BASIS EVDTP REANALYZED FOR FORF CAD 10UN CYCLE 11 P m for Acceptanos sumary j

Event Peanalysis Criterion. of Results ,

. (changes relative to reference cycle)

Baron Dilutica Incramaeri critical boren Dilutien to critical Acceptance criteria canoentrations from Cycle tima limits of 30 mimtes met. See Table 2.0

10. for refueling and 15 i

mimtes for all other subcritical miaa mst I be met.

i ir Ioss of Ioad Evaluate steam generator Peak RCS Pran-e $ 2750 Ppc3 = 2570 psia  :

tube plugging for 5 6% psia eh pluggecVSG.

h Ioad Oiange in 'DyIP trip func- - = 61.0 psia j tion (P ) trip equation. E is more limit-

) Reevalu5a gia, tem. 4 ing (as in Cycles 9 and j 10) than the RCS De-I pressurization.

4 Ioss of Feedwater Flow Evaluate steam generator Peak RG Pressure 5 2750 Ppc3 = 2487 psia psia

! tube plugging for S 6%

) tubes pluggetySG.

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! PCS Depressurizaticn Reevaluate Pbias term.

- Pya, = 26.8 psia i ufi1Ts is less limiting than that of Excess Ioad event.  ;

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j TABU 1.0 (Continued) i ,

IESIGi BASIS EVDfP REANALYZED FOR IbRP CADOUN CYCIE 11 Paa m for Acceptance Sunenary Event Reanalysis Criterion 3

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- of Results i

Sequential GA Croup Withdrawal Incraad 7%cil. Spec. Mininam IMR greater HDiBR = 1.31 limit on core inlet than 1.18 using s -1 PU CR < 22 kw/ft.

temperature ard steam correlation. Transient

' generator tube plug- PUCR < 22 kW/ft.

ging 5 6%/SG.

Imme of Coolant Flow Increased Tedt. Spec. , Mininum INBR greater Mininam INBR = 1.45 j

limit on oore inlet than 1.18 using 3 -1 tenperature and steam correlation.

i generator tube plug-ging 5 6%/SG.

Mill Iength EA Drop Incr=a w Tech. Spec. Minlmm INBR greater Minimum INBR = 1.43 limit on core inlet than 1.18 using G -1 tenperature and steam . oorrelation, generator tube plug- -

ging 5 6%/SG.  ; -

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l CEA Ejecticrt Increased ejected A Site boundary dose within Site boundary dose worth and steam gener- ' ' 10 CFR 100 limits, spec- acceptable. Imss

ator tube plugging 5 ifically less than 1% than 1% failed fuel.

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6%/SG. failed fuel.

LOCA (Lar.ge Break) Evaluate steam generator Peak clad temperature less Peak clad temp.= 18871 tube plugging for s6% than 2200"F local clad oxi- Local oxidation = 3.3t i

tubes plugged /SG. dation less than 17%. core Core wide

' wide clad oxidation less oxidation = 4 0. 5' than 1%.

l LOCA (Small Break) Evaluate steam generator (Same as for Large Break Peak clad temp. = 18954 tube plugging for 610% LOCA) local oxidation = 2.24' l tubes plugged /SG. Core wide 1 oxidation = <0.26 4

4. %,,0-TABLE E 0 FORT CA1110UN CYCLE 11 .

RESULTS OF THE BORON DILUTION EVENT Criterion For Minimum Time to lose Time to Lose Prescribed Shutdown Prescribed Shutdown li2df Martin (Mini Martin (Mini Cvele 10 Cvele 11 Hot Standby 93.8 91.6 15 Hot shutdown 45.8 44.7 +

15 Cold Shutdown - Normal '

RCS Volume 38.2 36.2 15 Cold Shutdown - Minimum RCS Volume 18.2 15.2 15 Refueling 31.2 31.8 30 o

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TABLE 3.0 Explanation for Cvele 11 Technical Specification Chances Tech. Spec. No. Chance Reasons

1. Pg. 5 Replace part-length The part-length CEA's CEA's definition with have been replaced with non-trippable CEA's (non trippable) CEA's.

definition.

2. P S. 6 Change part-length to The part-length CEA's non-trippable in have been replaced with definition of shutdown non-trippable CEA's.

margin.

3. Figure 1-3 Replace Figure 1-3 with The Cold Leg Temperature enclosed Figure 1-3. limit has been changed from 540*F to 545'F.

This has resulted in a change to the y term of the TM/LP equation.

4 Pg. 1 6 "strady" to " steady" Corrected typo.

5. 1.3.(8). Change the peak linear To maintain consistency 1.3.(9), heat rate from 21 KW/ft with the standard CE 14 Pg. 1-9 to 22 KW/ft. x 14 fuel assembly design limits.

6. 2.10.l(3) Replace part-length with See 1.

non-trippable.

7. 2.10.2(5) Replace part-length Part-length CEA's will Pg. 2-50c CEA's with full-length soon exceed their design non-trippable CEA's. life. Since Technical Specifications prevent their use during power operations and no shut-

. down margin is credited, full-length CEA's will

be used as replacements.

8. Figure 2-6 Replace Figure 2-6 with The LHR LCO has been enclosed Figure 2-6. changed to reflect addi-tional margin available with the symmetric core loading pattern.

! 9. Figure 2 8 Replace Figure 2-8 with Both CE and Exxon analy-enclosed Figure 2 8. ses have shown the 14 x

14 fuel augmentation is

! 1.0 for all axial loca-tions, j

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p TABLE 3.0 (Continued)

Exclanation for Cvele 11 Technical Soecification Channes q Tech. Socc. No. Chance Reasons

10. Figure 2-9 Replace Figure 2-9 with The F T and F R

enclosed Figure 2-9. limitNasafunctionof power have been revised to maintain consistency with the changes to Fig. 1 ure 2-6.

11. 2.10.4(2) Replace part-length with See 1.

2.10.4(3) non trippable.

Pg. 2-57a 12, 2.10.4(5) Change "$ 540*F" to "5 The Cold Leg Temperature (a)(i) 545'F." limit is being chan6ed Pg. 2 57c back to 545'F for increased unit l

, efficiency.

13. Footnote ** Change "

540*F" to '

The Cold Leg Temperature Pg. 2 57c "545'F" and "542*F" to limit is being changed "547'F." back to 545'F to in-crease unit efficiency.

14. 3.10(2)3 Expand the window for The current requirements Pg. 3-63a the MTC test, are one-sided and too 3

restrictive with daily plant activities. No sensitivity associated with test window.

15. 4.3.2 Change the figure. number Reflect the updated Pg. 4-3 from FSAR to USAR. refe rence.

16, 4.3.2 . Replace part-length Part-length CEA's will

Pg. 4-3 CEA's with full-length soon exceed design life.

non trippable CEA's. Will replace with full-length CEA's which will also be non-trippable,

, 17. 4.3.3, 4.5 Change the reference Pg. 4-3 Reflect the updated from FSAR to USAR. reference.

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