ML20043B195
| ML20043B195 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 05/16/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20043B194 | List: |
| References | |
| NUDOCS 9005250055 | |
| Download: ML20043B195 (7) | |
Text
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UNITED STATES
[.
NUCLEAR REGULATORY COMMISSION 5
W ASHINGT ON, D. C. 20555
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.34 TO FACILITY OPERATING LICENSE NO. NPF-51, ARIZONA PUBLIC SERVICE COMPANY, ET AL.
PALO VERDE NUCLEAR GENERATING STATION, UNIT 2 DOCKET NO. STN 50-529 i
1.0 INTRODUCTION
ByletterdatedNovember6,1989(Ref.2).theArizonaPublicService Company (APS) on behalf of itself, the Salt River Project Agricultural Improvement and Power District Southern California Edison Company, El Paso Electric Company, Public $ervice Compeny of New Mexico Los Angeles Departme:it of Water and Pcwer, and Southern California Pub 1Ic Power Authority { licensees), requested changes to the Tectoien1 Specificationt for the Palo Verde Naclear Cenerating)Statica, Unit 2 (Appendix A to i
Facility Operatina, License No M -51. The proposed changes would revise thase Nrtions of the Technh.c.1 Specifications regarding Shutdown Margin, Conti0161emnt Auesiy in:ortion Limits,in, in suppori. of Cycle 3 Axial Shere Index and Departu*e Fron Nucitate Boiling Rstto Metg ope:mtion for Palo Verds, Utif t 2.
In support nf both tte Technical Specifin. tion changes and Cycle S operation, the licensees submitted a Reload Analysis Report by letters dated October 24, 1989(Ref.2),andApril 16,1990(Reference 3).
The staff's evaluation of the reload analysis is presented in Section 2.0 through 5.0 below. The evaluation of the specific change to the Technical Specification is presented in Section 7.0 below.
The Cycle 3 core will consist of 241 fuel assemblies. Sixty-nine Batch B and 28 Batch C assemblies will be removed from the Cycle 2 core and replaced by 96 unirradiated Batch E assemblies. All Batch D assemblies and 36 Batch C assemblies from the Cycle 2 core will be retained.
In addition, one Batch B assembly discharged at end of Cycle 1 will be reinserted.
Burn-up distribution is based on a Cycle 2 length of 420 effectivefullpowerdays(EFPD). Cycle 3 control element assembly patterns and in-core instrument locations remain the same as in Cycle 2.
2.0 EVALUATION OF FUEL DESIGN 2.1 Mechanical Design The 96 Batch E assemblies to be added to the Cycle 3 core are identical in fD0052500ss 90033,3
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i design to the Cycle 2 Batch D assemblies except for changes to the poison I
rod assembly, the lower end fitting, and center guide tube.
The poison rod assembly was increased in overall length from 160.918 inches to 161.168 inches to improve burnup capability and reduce end-of-life internal pressure.
The two-piece lower end fitting was replaced by a one piece casting with a recess for the center guide tube. The length of the center guide tube was increased from 163.715 inches to 163.965 inches to make it compatible with the redesigned lower end fitting.
The above design changes represent minor improvements which do not affect i
the fuel mechanical design basis. We therefore find these changes acceptable.
1 Also, based on previous staff reload evaluations, clad collapse analyses of new C-E manufactured fuel do not need to be performed because of the absence of gaps between fuel pellets.
2.2 Thermal _ Design i
The thermal performance of Cycle 3 fuel was aaalyzed using the NRC-approved FATES 3A code and composite fuel pins that entelope the pins of Batches B.
C D, and E.
A powe history that er.veloped the pcwer 6 tid burrup levels of the puk pin e.t each burnup interval, from the beginning of cycle to the end of burnup, was used. The mayimum peak Din burnup analy2ed bounds that which is spected at the end of Cycle 3.
3ased on this e.n61ysis, the internal pressure in the most limiting fuel rod will stay below the nominal reactor coolant system (RCS) pretsure of 225C psi.
Because this satisfies Standard Review Flan (SRP) Section 4.2 criteria, we find the thermal design of the Cycle 3 core to be acceotable, i
3.0 EVALVATION OF NUCLEAR DESIGN 3.1 Fuel Management A general description of the Cycle 3 core is given in Section 1.0.
The Cycle 3 core uses a low-leakage fuel management scheme where previously burned Batch C assemblies are placed on the periphery and most of the
-fresh Batch E assemblies are located throughout the core interior in a pattern which minimizes power peaking. The highest Batch E enrichment is 4.03 weight percent U-235; the PVNGS fuel storage facilities are approved for a maximum enrichment of 4.05 weight percent U-235.
Expected Cycle 3 lifetime is 430 EFPD. A comparison of the Cycle 3 nominal characteristic physics parameters with those of Cycle 2 shows very little deviation between the two cycles.
3.2 Power Distribution Calculated "all-rods-out" relative assembly power densities have been resented for beginning of cycle (BOC), middle of cycle, and end of cycle p(E00). Relative assembly power densities are also given at BOC and EOC for rodded configurations allowed by the power dependent insertion limit at full power. These configurations consist of part length CEAs, Bank 5, and Bank 5 l
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l plus the part length CEAs. The Cycle 3 nominal axial peaHng factors are estimated to range from 1.21 to 1.11, at BOC and EOC, respectively. As approved for the Cycle 2 reload, densification augmentation factors have i
been eliminated for Cycle 3 because the same manufacturing process is used in the fuel fabrication.
Physics and power distribution calculations are based on the NRC-approved ROCS and MC codes employing DIT code generated neutron cross-sections.
The power distribution calculations are, therefore, j
acceptable.
3.3 Control Requirenents i
i The value of the required shutdown margin varies throughout core life with the most restrictive value occurring at E00 hot zero power (HZp) conditions.
1 This minimum shutdown margin of 6.5% delta k/k is required to control the reactivity transient resulting from the RCS cooldown associated with a steau line break accident at these conditions.
For operating teriperatures below 350'F, the reactivity transients rerv1 ting from any(postulated accident are minimal and a c 0 delta k/k shutdown margin revised from.3 value of 3.5 for Cycle 2) provides adequate protection.
SutTicient boration capability and net tve!1able CEA worth, including a minintum worth stuct CEA and appropriate calculational uncertainties, exist to meet these shutdown margin requirements. These respits were derived by approved methods and incorporate appropriate assumptions and are, therefore, acceptable.
4.0 EVALUATION OF THERM!.-HYDMULIC DESIGN Steaoy state thermal-hydrulic entlysis for Cycle $ is perforrcd using the approved thermal-hydraulic code TORC and the CE-1 critical heat flux (CHF) ewrelation. The design thermal margin analysis is performed with the fast running variation of the TORC code, CETOP-D. The CET0p-D model has been verified to predict minimum departure frcm nucleate boiling ratio (DNBR) conservatively relative to TORC.
The uncertainties associated with the system paranieters are combined statistically using the NRC-approved modified statistical ccmbination of unct:rtainties methodology. Using this methodology, the engineering hot channel factors for heat flux, heat input, fuel rod pitch, and cladding diameter are combined statistically with other uncertainty factors to arrive at overall vocertainty penalty factors to be applied to the DNBR calculations performed by the core protection calculetors (CPCs) and the Core Operating Limit Supervisory System (COLSS). When used with the Cycle 3 DhER limit of 1.24, these overall uncertainty penalty factors provide assurance with a 95/95 confidence / probability that the hottest fuel rod will not experience DNB.
L The 1.2A value incorporates all applicable penalties, such as for rod bow, the 0.01 DNBR for HID-1 Srids, and the penalties specified in the statistical l
combination of uncertainties. The rod bow value used in the analysis is 1.75% M DR, for burnups up to 30,000 MWD /MTU.
For burnups higher than 30,000 MWD /MTV, sufficient margin exists to offset the rod bow penalty due to lower
e
. i radial power peaks in these higher burnup assemblies and rods. Therefore, l
the rod bow penalty is adequate for all anticipated burnups.
Because the thermal-hydraulic design analyses were performed using approved codes and took into account all cpp11 cable penalties, we find these analyses acceptable.
5.0 EVALUATION OF NON-LOCA SAFETY ANALYSl$
The design basis events (DBEs) considered in the safety analyses are categorized into two groups: anticipated operational occurrences (A00s) and postulated accidents (limiting faults). All events were reviewed by the licensee to assess the need for reanalysis as a result of the new core configuration for Cycle 3.
The DBEs were evaluated with respect to the following four criteria:
fuel performance (DNBR and centerline melt) RCS pressure, loss of shutdown margin, and offsite dose. The limiting fault events corresponding to each criterien were reanalyzed.
Plant response to the DSEs was simulated using the same mathods and computer programs which were used and approved for the Cycle 2 ar,alysos.
l These include the CESEC III, ST31 KIN Ils TORl and HERM!TE computer programs.
for some of the reanalyzed CBLs, certain in'.tial core perameters nere i
assurned to be more limiting than the calevicted Cycle 3 values in order to bound future cycles.
All of the events reanal
=>lthin NRC acceptence criteria and, therefore,yzed have results WMch dre are acceptatle.
Tw3 o? the reanalyzed events, however, were not bounded by the Cycle 2 analnes.
l These are the inadvertent opening of a stesr generator sefety salve or atmospheric dump valve with loss of offrite power, and the singit reactor l-coolant pu:np shaf t seizure / sheared shaf t event.
For the former event, the l
amount of predicted failed fuel increased from 8% to 12% as a result of I
more adverse nuclear power distributions.
However,thebasecase(i.e.
l without loss of offsite power) is bounded by the Cycle 2 analyses.
For the latter event, an increase in predicted fuel failure from 3.79% to 4.5%
occurs. The resulting radiological consequences are within 10 CFR 100 guidelines and are acceptable.
6.0 EVALUATION OF ECCS ANALYSIS l
l An ECCS analysis was performed for the limiting break size LOCA (a double-ended guillotine break with a 1.0 discharge coefficient) for Cycle 3 to demonstrate compliance with the requirements of 10 CFR 50.46. The methodology is the same as for the Cycle 2 analysis. The analysis justifies a 13.5 kw/ft peak linear heat generation rate.
Because there have been no significant changes in hardware characterisitics for Cycle 3, only fuel rod clad temperature and oxidation calculations were performed. The code STRIKIN-II was used for this purpose and the fuel performance data were generated using the FATES-3A fuel evaluation code.
It was demonstrated that burnup with the highest initial fuel stored energy was limiting.
The ECCS analysis methods employed have been previously approved and are acceptable.
The results of the limiting break LOCA analysis for Cycle 3 are bounded by the results obtained in the Cycle 2 analysis, i.e., a peak clad temperature
i o
e of 2091'F, a maximum local clad oxidation of 9.0%, and a core wide clad oxidation of less than 0.80%.
These values are within the 10 CFR 50.46 4
limits of 2200'T 17.0% and 1.0%, respectively and are, therefore, acceptable. SimIlarly,,a review of Cycle 3 fuel and core data has confirmed that the small break LOCA analysis results are bounded by the Cycle 2 analysis and is acceptable.
7.0 TECHNICAL SPECIFICATION CHANGES TS Figure 3.1-1A The proposed change increases the required shutdown margin from 3.5 to 4.0%
delta k/k for the RCS cold leg temperature range zero to 350'F when any full-length CEA is fully withdrawn.
Tht-incre6 sed shutdown margin will ensure that the TS are consistent with the safety onelyses perfomed for the Cycle 3 core and that the consequences of DBIs e.nd enticipeted operational occurrences are bounded by these analyses.
The proposed change is therefore accettable.
TSicbles;bl2.3.1-3_and3.1-5 These t:bles provide frequencies for monitoring RCS boron concentration in the nert that one or both startup channel high neutron flux alarmi are inoperable.
The proposed changes are more restrictive in that certain inonitoring frt.quenciet.are increased to ensure that the TS are consistent with the safety analyses performed for the Cycle 3 core and that, in the event of an inadvertent boron dilution sufficient time will be available to terminatetheeventpriortolossofshutdownmargin. The proposed changes are, therefore, acceptable.
TS Figures 3.1-3 and 3.1-4 Figures 3.1-3 and 3.1-4 provide regulating group CEA insertion limits when the COLSS is in service and out of service, respectively. The proposed change to Figure 3.1-3 will prohibit insertion of regulating group 3 CEAS above 20% of rated thermal power. This is permitted under the existing TS. The proposed change to Figure 3.1-4 will permit slightly increased insertion of regulating group 3 CEAs bc?. ween J5% and 20% of rated thermal power.
The proposed revisions are necessary to ensure consittency of the TS with the safety analyses performed for the Cycle 3 core.
The5e analyses demonstrate that reactor operation in accordance with the revised insertion limits will ensurethattheSpecifiedAcceptablefuelDesignLimits(SAFDLs)willnotbe exceeded during the most limiting anticipated operational occurrence. The proposed changes are, therefore, acceptable.
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TS 3.?.7a i
TS 3.2.7a ensures that the actual value of the core everage Axial Shape Inder (ASI) remains within the range of values used in the safety analyses when the COLSS is operable. The proposed change revises the limits of core 6verage ASI from 28 SI
.28 to.27 SI
.27 to make the TS consistent with the safety analyses performed for the Cycle 3 core. The j
proposed charge is, therefore, acceptable.
TS Figures 3.2-2 and 3.2-2A i
Figure 3.2-2 provides DNBR margin limits when at least one Control Element Assembly Calculator (CEAC) is operable and the COLSS is out of service.
Figure 3.2-2A provides the additional DNPR margin necessary when COLSS and both CEACs are out of service. Reactor operation within these limits ensures that the SAFDLs will not be violated during an anticiptted op> rational occurrence.
The proposed chcopes are necessary to enture consistency of the TS with the tefety analyses perforred for the Cycle 3 core and are, therefore.
6Cceptable.
8.0 STAP.YUF 1EST1Na l
Tbc licensee has presented a brief description of the icw power physles tests and the power ascer,sion testing to be performed during Cycle $
startup. The described tests will verify that core performance it consistent with the engineering design ami safety analyses.
If the acceptrnce criterion of any of the startup physics tests are not met, an evaluation will be performed by the licensee.
Resolution will be required prior to subsequent power escaletion.
If an unreviewed safety question is involveo, the HRC will be notified.
The staff has reviewed the proposed startup test prograr for Cycle 3 and finds that it conferns to accested practices and adequately supplements norraal surveillance tests whic1 are required by the plant Technical Specifications.
9.0 EVALVATI0t
FINDINGS l
The staff has reviewed the fuels,le 3 reload report. physics, and thermal-hydraulics l
presented in the pvt:GS Unit 2 Cyc Also reviewed were i
the Technical Specification revisions, the startup test procedures, and the safety reanalyses.
Based on the evaluations given in the preceding l-sections, the staff finds the proposed reload acceptable, i
10.0 CONTACT WITH STATE OFFICIAL l
1 l.
The Arizona Radiation Regulatory Agency has been advised of the proposed determination of no significent hazards consideration with regard to these changes, flo comments were received.
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11.0 ENVIRONMENTAL CONSIDERATION
i The amendment involves changes in requirements with respect to the instal-lation or use of facility components 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 amount, and no significant change in the type, of any effluent that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued proposed findings that the amendment involves no significant hazards consideration, and there has been no public comment on suc1 finding. Accordingly, the amendment meets the i
eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuantto10CFR51.22(b)noenvironmentalimpactstatementorenviron-mental assessment need to be prepared in connection with the issuance of f
the anendment.
12.0 gfjLUSION 7he staff has concluded, based on the considerations ciscussed above, that (1) there is reasonable assurance that the health and safety of the pet:lic will not be emiaogered by operctior, in the proposed racr.er, (2) such activities will h condJCted in CompM$nce With the CDmission'%
regulations, and (3) the is n ance of the amendment will r.ot be inimical io tht common defense and security or to the health and safety of the tublic. We, therefore, conclude that the proposed changes are acceptable.
13.0 REFERENCES
1.
Reload Safety Analysis Report for Palo Verde Nuclear Generating Station Unit 2, Cycle 3, submitted by letter from W. F. Conway (APS), dated October 24, 1989.
1 2.
Proposed Reload Technical Specification Changes for Palo Verde Nuclear Generating Station Unit 2, Cycle 3, submitted by letter from W. F. Conway (APS),datedNovember6,1989.
3.
Revision to Reload Analysis Report for Palo Verde Nuclear Generating Station Unit 2, Cycle 3, submitted by letter from W.F. Conway (APS),
dated April 16, 1990.
Principal contributor:
H. Abelson Dated: May 16, 1990
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