ML20034A074
| ML20034A074 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 04/04/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20034A071 | List: |
| References | |
| NUDOCS 9004190275 | |
| Download: ML20034A074 (9) | |
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UNITED STATES e c'
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',g NUCLEAR REGULATORY COMMISSION 5
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.126TO 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 26, 1990, OmahaPublicPowerDistrict(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 (TSs) to support Cycle 13 operation.- By letter dated February 6,1990, OPPD submitted an additional change associated with the amendment request. The January 26, 1990, submittal reflected a lowering of the allowable peak liner heat rate which was indicated on Figure 2-5 of the TSs but was inadverter,tly not changed on TSs page 2-57. The February 6,1990., letter made this correction.
2.0 DISCUSSION The reload methodology, fuel system, nuclear, thermal-hydraulic, safety analyses, and startup tests are discussed herein.
2.1 Reload Methodology In their amendment application, OPPD has identified reload methodology topical reports (References 1, 2, and 3) to be used for Cycle 13. This methodology was used for the prior Cycle 12 reload and remains acceptable.
2.2 Fuel System Design The mechanical design for Batch P reload fuel is essentially (CE), as the same as that of the Batch N fuel supplied by Combustion Engineering described in the Cycle 12 reload submittal.
The Batch P fuel is similar in design to the Batch G fuel supplied by CE in Cycle 5 and is mechanically,-
thermally, and hydraulicelly compatible with the Advanced Nuclear Fuel (ANF) fuel remaining in the core. The.CE fuel has been analyzed for previous cycles using approved methods and adequately bounds the.
end-of-cycle (E00) exposure. The analysis approved for the residual ANF
. supplied fuel remains valid.
2.3. Nuclear Design 2.3.1 Core Characteristics The Cycle 13 fuel management uses a low-radial leakage design with I
once, twice, and thrice burned assemblies predominately loaded on the l
periphery of the core.
This fuel pattern is utilized to minimize the fl 9004190275 900404 ADOCK 0500g;t]5 L
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! fluence to the pressure vessel welds and achieve the maximum in neutron econory. This fuel management, however, results in higher radial peaking factors than the standard out-in-in pattern. The peaking factors for Cycle 13 are consistent with previous cycles in i
which the low radial leakage patterns have been utilized.
The Cycle 13 loading pattern incorporates 40 fresh Batch P fuel
. assemblies (32 shimmed and 8 unshimmed) with an average enrichment of 3.66 weight percent U-235. Eight thrice burned Batch L assemblies, initially loaded.into the core for Cycle 11 are being returned to the core along with 41.twice burned Batch M assemblies, and 44 once burned Batch N assemblies to produce a pattern with a cycle energy of 14,250 +/- 1000 MWD /MTV. The Cycle ~13 core characteristics have been examined for a Cycle 12 termination between 10,000 MWD /MTU and 12 MWD /HTU and limiting values established for the safety analysis.,000 The loading pattern is valid for any cycle endpoint between these values.
2.3.2 Power Distributions The all rods out (ARO) planar radial power distributions at Beginning-of-Cycle (B00), Middle-of-Cycle (MOC),andEnd-of-Cycle (E00)have been calculated and are an average of the low and high burnup ends of the Cycle 12 shutdown window. The high burnup end of this shutdown window tends to increase the power peaking in the high power region of the core for Cycle 13. The calculated power densities also shew that the maximum expected peaking factors for Cycle 13 are within the proposed Technical Specification limits.
For both the departure from nucleate boiling and linear heat rate safety and setpoint analyses in either the rodded or unrodded configurations, the power peaking values used are higher than those actually expected to occur at any time during Cycle 13.
The range of allowable axial peaking is defined by the limiting conditions for operation (LCO) covering the axial shape index (ASI).
The maximum three-dimensional or total peaking factor anticipated in Cycle 13 during normal base load, ARO operation at full power is 1.82, not including uncertainty allowances.
Based on the above, the radial peaking limits and the axial shapes are acceptable.
2.3.3 Moderator Temperature Coefficients The Technical Specifications reo,uire the moderator temperature coefficient (MTC) be less positive than +0.2 x 10~4 delta-rho /'F
(+0.50x10~ delta-rho /*F for power lev power) and less negative than -2.7 x 10~gls below 80 percent of rated delta-rho / F at all times during Cycle 13.
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Calculations have shown that these limits are met for the operating conditions. Since acceptable methods have been used and appropriate values incorporated in the safety analyses, the range of MTC for Cycle 13 is acceptable.
2.3.4 Control Requirement t
The value of the required shutdown margin is determined by the EOC steam line break analysis occurring at hot zero powe.' and remains at 4.0 percent ' delta-k/k for Cycle 13.
Based on this value and the available' scram reactivity, including allowance for a maximum worth stuck Control Element Assembly and appropriate calculational uncertainties, sufficient excess reactivity exists between the available and the required scram reactivities for all Cycle 13 operating conditions. The results were derived by approved methods and incorporate appropriate assumptions and are, therefore, acceptable.
2.4 Th,e,rma,1.-Jiydraulic, _ Design 2.4.1 DeparturefromNucleateBoiling(DNBR) Analysis The thermal-hydraulic methodologies for Cycle 13 were serformed using computer codes previously approved by the NRC for use >y OPPD. The steady state DNBR analyses used the TORC computer code, the CE-1 critical heat flux correlation, and the CETOP-D computer code. -This con.bination was used in the previous cycles starting with the Cycle 8 reload analyses. The calculational factors, including the engineering heat flux factor, the engineering factor on hot channel heat input, rod pitch and clad diameter factor, were statistically combined with other uncertainty factors to arrive at the equivalent DNCR limit of 1.18.
The statistical combination of factors has been approved by the NRC. This limit ensures that, with at least a 95% probability and at least a 95% confidence level, the limiting fuel pin will avoid departurefromnucleateboilingifthepredictedminimumDNBR(MDNBR) is not below the 1.18 limit.
2.4.2 fuel Rod Bowing The fuel rod bow penalty accounts for the adverse impact on MDNBR of random variations in spacing between fuel rods. The methodology for determining this penalty was based on NRC approved methods. The penalty at 45,000 MWD /MTU burnup is 0.5% in MDNBR, and was applied directly~to the MDNBR design limit of 1.18 in the statistical combination ofuncertainties(Ref.4).
2.5 S,af,e,ty,,Ana,1ysi s OPPD has reviewed the pararacters which influence the results of the transient and accident analyses for Cycle 13 to determine which, if any, would require reanalysis. With regard to non-l.0CA safety-analysis at 1500 MWt for Cycle 13, the design basis events'(DBE) were considered, and a comparison of core
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t parameters to bounding values was made. No reanalysis was performed for i
those DBEs in which the key transient input parameters were within the bounds of the reference cycle values (Refs. 5 and 6).
1 The results and conclusions for those DBEs quoted in the reference cycle analysis remain valid for Cycle 13. All the events were evaluated for an assumption of 6% steam generator tube plugging.
Table 1.0(Attachment 1)liststheDBEsthatwerereanalyzedandthe reasons for the reanalysis, the acceptance criterion used in judging the results, and a summary of the results obtained. The NRC has reviewed
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these results and finds them to be valid for Cycle 13 operation.
'The Cycle 11 Large and Small Break Loss of Coolant Accident analyses were perforud using the methodology previously approved by the NRC (Ref.1).
The Cycle 11 revised Emergency Core Cooling System performance analysis was verified to be valid for Cycle 13 given the bounding input assumptions.
The peak linear heat generation rate of 15.2 kw/ft was conservatively reduced to 14.4 Lw/ft to ensure that CE fuel mechanical design requirements were valid for the entire Cycle 13 operation. This change was shown in the initial submittal on January 26, 1990, for TS Figure 2-5, but the change on TS page 2-57 was inadvertently not changed.
By letter dated February 6,1990, the change on TS page 2-57 was made.
2.6 Startup Tests The startup testing program for Cycle 13 is identical to that used in 3
Cycle 12, and with two exceptions is the l
TheseexceptIonsare:sameasthatusedintheCycle6 Reload Application.
(1)theCEAexchangetechnique (Ref.7)forzeropowerrodworthmeasurementswillbeperformedin accordance with Reference 7 replacing the boration/ dilution method, and (2) the low power CECOR flux maps and pseudo-injection rod measurements will be substituted for the full core syrnetry checks. The NRC has reviewed the acceptance criteria for these tests and finds them acceptable (Ref.8).
3.0 QA,LUATION 3.1 Te ch n i,c a,1_Sp e c i f i ca t i o n C h a ng e s OPPD has proposed changes to the Technical Specifications for Cycle 13 which arelistedinTable2.0(Attachment 2). The NRC has reviewed these changes and finds that they are properly incorporated in the supporting physics and safety analyses for Cycle 13 using the approved mM.tods. The changes, and the additional changes to correct typographical errors, are acceptable.
3.2 Flndings The NRC staff has reviewed the information presented in the Cycle 13 reload application. The staff finds the proposed reload and the essociated modified Technical Specifications acceptable.
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i 4.0 ENVIRON!tENTAL CONSIDERATION The amendment involves a change to a requirement with respect to 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 exposures. The Commission has previously issued 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 categoricalPursuant to 10 exclusion set forth in 10 CFR Section 51.22(c)(9.
no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
5.0 CONCLUSION
The staff has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of-ihe 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 issuance cf the amcodment will not be inimical to the common defense and security or to the health and safety of the public.
Date: April 4, 1990 Principal Contributor:
M. McCoy
-Attachments:
- 1. Table 1.0
- 2. Table 2.0 J
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TABLE 1.0 i
i FORT CALHOUN UNIT NO. 1 DESIGN BASIS EVENTS REANALYZED FOR CYCLE 13 i
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Reason for Acceptance Sassary 2
Event Reanalysis Criterion of Results Sequential CFA Group Increased reactivity insertion Minim e DNBR greater' PDNBR =1.72-Withdrawal rate and rod shadowing thm* i.18 using the CE-1 PLNGR< 22kW/ft 1
factor.
correlation. Transient i.
PLHGR <22.kW/f t.
CEA Drop Increase in distortion Minimum DNBR greater-MDNFR = 1.36 factors nonconservative than 1.18 using CE-1 PLHGR (22 kW/ft.
with lower rod worth, correlation and Transient.
PLHGR< 22 kW/ft.
Excess Load Reduced minimum scram Transient PLHGR< 22 kW/ft MDNBR = 1.22 i
- worth, and Minimum DNBR greater:
PLMGR< 22 kW/ft.-
than 1.18.
Boron Dillution Increase in Critical' Boron Criterion'for minimum time Actual time to lose:
Concer.tration.
tol lose prescribed shutdown mergin >
shutdown mergin. ~(15 Miniumum time.
minutes for. Modes 3, 4' and 5, 30 minutes for Mode.6)
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TABLE 2.0 Explar,ation for Cycle 13 Technical Specificetion Changes Item Tech. Spec.
No.
Page/ Figure No.
Change Reasons 1
Page 1-2 Change total unrodded The revised values'are planar radial peak from conservative with respect 1.00-to 1.75 and the
.to the previous values. The unrodded intcgrated total reduced value will provide radici peak from 1.80 to additional operating-1.70, margin.
2 Figure 1-3 Replace Figure 1-3 with The gema term of the the enclosed figure 1-3.
P equation has been cMgedto'reflectan increased p term and.
reducedpowN"lofuel design limits on DNBR.
3 Figure 2-5 Replace figure 2-5 with The allowable peak linear the enclosed Figure 2-5.
heat rate LOCA limit has been reduced to provide additier,61 operating targin. The 14.4 kw/ft it
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conservative with respect to the previous value.
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4 Figure 2-9 Replace Figure 2-9 with
.The F and F limits the enclosed Figure 2-g.
asaNnctionOfpowerhave been revised to maintain consistency with changes te itens 1 end 6.
5 Fage 2-57a Change total unrodded The revised values are planar radial peak from conservative with respect 1.80 to 1.75 ar.d the to the previous values.
unrodded integrated total The reduced values will radial peak from 1.80 to provide additional 1.70._
operating margin.
6 Page 1-2 Change "Section-2-10" to Correction of typographical "Section 2.1C" in para errors.
graph 3, line 2 and change "that" to "than" in paragraph 3, line 7.
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-TABLE 2.0 Explanetion for Cycle 13 Technical Specification Changes item Tech. Spec.
No.
Page/ Figure No.
Change Reasons-7 Page 2-57a Change-" vale" toi"value" Correction of a typographical'
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in2.10.4(3) paragraph'1, error.
line 1.
Add"(F T " on para-Administrative changes to.
graph 37Y)ne4, follow--
8 Page 1 11 define unrodded planar;
.ing the words " total radial peak and update USAR
.unrodded planar radial.
references.
Reference (3) peak." _Also change currently corresponds to Reference (1)fromUSAR USAR Section 3.6.6 and-Section "3.6.7".to-Reference (3)isnowthe "3.6.6" and delete sameasReference-(1).
Reference (3) ir.'the
" References.
As a result, chan e "less than 1.18 (3
- to "less than 1.18 (1 "'in para-graph 4,-line 10.
-9 Fage 2-57_
Change the denominator The allowable-peak linear value in the expressier.
heat rate in' Figure 2-5 has contained in Section been reduced from 15.5 2.10.4(c)to14.4.
Lw/ft to 14.4 kw/f t.
The 14.4 kw/f t value is conser-vative with respect to the previous value.
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- c, o-I REFEREttCES 1.
" Omaha Public Power District Reload Core Ar.alysis Methodology - Transient and Accident Methods and Verification," OPPD-NA-8303-P, Revision 2,. dated April 1988.
2..
" Omaha Public Pcwer District Reload Core Analysis Methodology Overview,"
=0 PPD-NA-8301-P, Revision 3, dated April 1988.
3.-
" Omaha.Public Power-District. Reload Core Analysis Vethodology -
Neutronics Design Methods and~ Verification," OPPD-NA-8302-P, Revision 2, dated April'1986.
-l 4.
1" Statistical Combination of Uncertainties, Part 2," Supplement 1-P, CEN-257(0)-P,datedAugust1985.
S.
Fort Calhcun Station Updated Safcty. Analysis Report, dated July.1988.
6.
"Ao endt.;ent tc Operating' License.DPR-40, Cycle 11 License Application,"
j Decket No. 50-285, dated May 4, 1987.
7.
" Control Rod Group Exchenge Technique," CEN-319, dated November 1985.
D.
" Acceptance for Referencing of Licensing' Topical Report Cell-319 - Control l
Rod Grcup Exchange Technique," with letter, D. H. Crutchfield (NRC to R.
W.-Wellt(Chairman--CEOwnersGroup),datedApril 16, 1986.
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