ML20077D454
| ML20077D454 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 05/20/1991 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20077D452 | List: |
| References | |
| NUDOCS 9105300224 | |
| Download: ML20077D454 (8) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT N0. 26 TO FACILITY OPERATING LICENSE NO. NPF-74 ARIZONA PUBLIC SERVICE COMPANY, ET AL.
PALO VERDE NUCLEAR GENERATING STATION, UNIT N0. 3 DOCKET NO. STN 50-530
1.0 INTRODUCTION
By letter dated February 21, 1991, Arizona Public Service Company, the licensee for the Palo Verde Nuclear Generating Station Unit 3 (PVNGS3), submitted a reload safety analysis report in support of a request to reload and operate PVNGS3 for a third cycle at 100 percent ated core power of 3800 MWt. The licensee also submitted proposed changes to the Technical Specifications (TS) to support Cycle 3 operation.
The Cycle 3 core will consist of 241 fuel assemblies.
Seventy-three Batch B and 48 Batch C assemblies will be removed from the Cycle 2 corr. and replaced by 88 unirradiated Batch E assemblies. One-hun 6 ed and four Batch D assemblies and 16 Batch C assemblies from the Cycle 2 core will be retained, in addition, 33 Batch 8 assemblies discharged at end of Cycle 1 will be reinserted.
Burnup distribution is based on a Cycle 2 length of 436 effective full powe: days (EFPD). Cycle 3 control element assembly patterns and in-core inst"onant locations remain the same as in Cycle 2.
The staff has reviewed the licensee's submittal of February 21, 1591, aid has prepared the following evaluation of the proposed TS changes, the fuel design, nuclear design, thermal-hydraulic design and accident / transient analyses associated with the Cycle 3 core.
2.0 EVALUATION OF FUEL DESIGN 2.1 Mechanical Design The 88 Batch E assemblies to be added to the Cycle 3 core are identical in design to the Cycle 2 Batch D assemblies except for changes to the poison rod assembly, the lower end fitting, and center guide tube.
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The poison rod assembly was increased in overall length from 160.918. inches to l
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 i
was increased from 163.715 inches to 163.965 inches to make it compatible with j
the redesigned lower end fitting.
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2 The above design changes represent minor improvements which do not affect the fuel mechanical design basis. The staff tierefore, finds these changes acceptable. Also,basedonpreviousstaffreloadevaluations,cladcollapse analyses of new C-E manufactured fuel do not need to be performed because the time to clad collapse is in excess of any prsctical core residence time.
2.2 Thennal Design The thermal performance of Cycle 3 fuel was analyzed using the NRC-approved FATES 3A code and :omposite fuel pins that envelope the pins of Batches B, C, D, and E.
A powea history that enveloped the power and burnup levels of the peak pin at each tarnup interval, from the beginning of cycle to the end of cycle, was used. The maximum peak pin burnup analyzed bounos that expected at the end of Cycle 3.
Based on this analysis, the internal pressure in the most limiting fuel rod will stay below the nominal reactor coolant system (RCS) pressure of 2250 psi.
Because this satisfies Standard Review plan (SRP)
Section 4.2 criteria, the thermal design of the Cycle 3 core is acceptable.
3.0 EVALUATION 0* NUCLEAR DESIGN 3.1 Fuel Management A general description of the Cycit. 3 core is given in Section 1.0.
The Cycle 3 core uses a low-leakage fuel management scheme where previously burned Batch B assemblies are placed on the periphery and most of the fresh Batch E assemblics are located throughout the core interior in a pattern which minimizes power peaking. The highest Batch E enrichment is 3.96 weight percent U-235; the PVNG5 fuel storage facilities t.re approved for a maximum enrichment of 4.05 weight percent U-235. Expected Cycle 3 lifetime is 390 EFPD. A comparison of the Cycle 3 nominal characteristic physics parameters with those used in the safety analyses shew that the latter are conservative in all cases.
3.2 Power Distribution Calculated "all-rod-out" relative assembly power densities have been presented for beginning of cycle (BOC), middle of cycle, and end of cycle (E00). Relative assenbly power densities are also given at BOC and E0C for rodded configurations allowed by the power dependent insertion limit at full power. These configura-tions consist of part length CEAs, Bank 5 and Bank 5 plus the part length CEAs. TheCycle3nominalaxialpeakingfactorsare4.stimated to range from 1.22 to 1.08, at BOC and EOC, respectively, physics and power distribution calculations are based on the NRC-approved ROCS and MC codes e@loying DIT code generated neutron cross-sections. The power distribution calculations are, therefore, acceptabic.
3.3 Control Requirements The value of the required shutdown margin varies throughout core life with the most restrictive value occurrir.g at EOC hot zero power (HZP) conditions. This minimum shutoown margin of 6.5 percent delta k/k is required to control the reactivity transient resulting from the RCS cooldown associated with a steam
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-3 line break accident at these conditions.
For uperating temperatures below 350*F, the reactivity transients resulting from any postulated accident are minimal and a 4.0 percent delta k/L shutdown margin (revised f rom a value of 3.5 for Cycle 2) provides adequate protection.
Sufficient boration capability and net available CEA worth, including a minimum worth stuck CEA and appropriate calculational uncertainties, exist to meet these shutdown margin requirements.
These results were derived by approved methods and incorporate appropriate assumptions and are, therefore, acceptable.
4.0 EVALUATION OF THERMAL-HYDRAULIC DESIGN Steady-state thermal-hydraulic analysis for Cycle 3 is performed using the approved thermal-hydraulic code TORC and the CE-1 critical heat flux (CHF) correlation.
The design thermal margin analysis is performed with the f ast running variation of the TORC code, CETOP-D.
Se CETOP-D model has been verified to predict minimum departure fror te boiling ratio (DNBR) canservatively relative to 10RC.
The uncertainties associated with the system parameters are combined statisti-j cally using the NRC-approved modified statistical combination of uncertainties 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 f actors to arrive at overall uncertainty penalty factors to be applie. to the DNBR calculations performed by the core protection calculators (CKs) and the Core Operating Limit Supervisory System (COLSS). When used wit
- the Cycle 3 DNBR limit of 1.24, these overall uncer-tei. ty penalty f actor', provide assurance with a 95/95 confidence / probability that the hottest f Wl rod will not experience DNR.
The 1.24 value incorporates all applicable.oenalties, such as for rod bow, the 0.01 DNBR fo: HID-1 grids, and the penaltie's specified in the statistical combinatie: of uncertainties. The rod bow value used in the analysis is 1.75 percent DNER, for burnups up to 30,000 MWD /MTV.
For burnups higher than 30,000 MWD /MMl, sufficient margin exists to offset the rod bow penalty due to lower radial power peaks in these higher burnup assemblies and rods.
Therefore, 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 applicable penalties, the staff finds these analyses acceptable.
5.0 EVALUATION Of NON-LOCA SAFETY ANALYSIS The design basis events (DBEs) considered in the safety analy(A00s) and postu-ses are catego-rized-into two groups:- anticipated operational occurrences i
lated accidents (limiting f aults). All events were reviewed by '.he licensee to assess the need for reanalysis as a result of the new core configuration for Cycle 3.
The DBEs were evalurted with respect to the following four criteria:
fuel performance (DNBR and centerline melt), RCS pressure, loss of shutdown i
margin, and offsite dose.
The limiting fault events corresponding to each criterion were reanalyzed, u..
, plant response to the DBEs was simulated using the same methods and computer These include programs which were used and approved for the Cycle 2 analyses.
For the CESEC 111, STRIKIN-11, CETOP-D, TORC, and HERMITE computer programs.
some of the reanalyzed DBEs, certain initial core parameters were assumed to be more limiting than the calculated Cycle 3 values in order to bound future cycles. All of the events reanalyzed have results which are within NRC accep-tance criteria and, therefore, are acceptable. Two of the reanalyzed events, however, were not bounded by the Cycle 2 analyses. These are the inadvertent opening of a steam generator safety valve or atmospheric dump valve (ADV) wita loss of offsite power and the single reactor coolant pump shaft seizure /shea'ed shaft event with loss of offsite power and a single active failure of the ADJ to c'ese. This single failure for the latter event maximizes the radiological For the latter event, an increase in predicted fuel failure froc consequences.
3.79 percent to 4.5 percent occurs. The resulting radiological consequences are within 10 CFR 100 guidelines and therefore, meets the appropriate dose criteria and are acceptable.
For the former event, the amount of predicted failed fuel increased from 8 percent to 12 percent as a result of more adverse nuclear power distributions.
The maj?r parameter of concern is the number of fuel rods which experience DNB.
This pa"ameter is used to determine if fuel cladding degradation might be An anticipated and determines the source for the resulting dose calculations.failure ADV may be inadvertently opened by the operator or may be open due to :
of the iontrol system which operates the valve. The worst single failure for this eve nt is the loss of of fsite power concurrent with a turbine trip (LOP) since ttis combines the greatest decrease in DNBR after initiation of a reactor trip signal with the lowest possible pretrip DNBR. The loss of flow due to the 4 pump coastdown, which results from the assumption of LOP following turbine trip, causes a greater decrease in DNBR after reactor trip than other possible single iailures.
In addition to the assumed single failure of loss of offsite power, the most reactive CEa is assumed to be stuck in the fully withdrawn position following reactor trip. The licensee indicated that the ADVs are air Nr the ADV to operated and are spring loaded to fail closed on loss of air.
of DC open and remain open, there must be 6 failures involving 2 channels In order for an inadvertently opened ADV to remain open due to. M ani-power.
cal binding, the valve would need to seize up so firmly that it could not De closed neither by air pressure, spring nor manual handwheel operation.
The FSAR used a deterministic method of predicting fuel rod failure in which any fuel rod falling below a DNBR limit of 1.24 was assumed to experience cladding failure with all of the activity in the fuel-clad gap released to the primary coolant. The Cycle 3 analysis used a statistical convolution approach in which the probability of being in DNB at a given DNBR is taken into As in the deterministic method, a fuel rod is still assumed to fail account.
if it e neriences DNB. This approach had been found acceptable by the staff for analysis of limiting faults such as the locked rotor, sheared shaft and CEA ejection accidents at PVNGS. These are occurrences that are not expected to occur but are postulated because their consequences would include the potential for the release of significant amounts of radioactive material.
Although the inadvertent opening of an ADV with loss of offsite power is not 1
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, categorized as a limiting fault but rather as an infrequent incident which may be expected to occur during the lifetime of a plant the staff concludes that the use of statistical convolution to determine fuel f ailures for this event is acceptable for PVNGS. This conclusion is based on the unique design of the PVNGS ADVs and the resulting low probability of failure to close which was discussed previously.
It is also based on the conservative transient assump-tions mentioned previously as well as the conservatisms inherent in the metho-dologyly inlet flow equal to 73% of the core average.such as the fact that all DNBR va assemb A realistic flow distri-bution would result in a much smaller number of f ailures.
For this event, the staff practice has been to restrict the resulting two hour site boundary dogs to a "small fraction" (10% or less) of 10 CFR 100 guidelines of 300 rem to the thyroid and 25 rem to the whole body. The resultant offsite dose for this event using the statistical convolution method has been calculated by the licensee to be 30 rem thyroid and less than 2 rem whole body, thereby meeting the staff's criterton, in conclusion, the staff finds the results of the inadvertent opening of a stcam generator ADV with a loss of offsite power using a statistical convolution of DNB are acceptable for PVNGS 3 Cycle 3.
Since the dose consequences of this meet the acceptable limit of 10% of 10 CFR 100, this analysis is acceptable as the reference analysis for PVNGS. All future reload analyses performed for this event must use the same assumptions and methods used for the Unit 3. Cycle 3 analyses described above. Changes to these assumptions and methods should be submitted to the staff including a discussion of why the statistical convolution methodology remains acceptable for this event.
6.0,(VALUATION OF ECCS ANALYSIS 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 demon-strate 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 characteristics for Cycle 3, only fuel rod clad temperature and oxidation calculations were performed. The code STRIKIN-Il 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 bmployed 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 of 2091*F, a maximum local clad oxidation of 9.0 percent, and a core wide clad oxidation of less than 0.80 percent. These values are within the 10 CFR 50.46 limits of 2200*F,17.0 percent, and 1.0 percent, respectively, and are, there-fore, acceptable.
Similarly, a review of Cycle 3 fuel and core deta has con-firmed that the small break LOCA analysis results are bounded by the Cycle 2 ana lys is.
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7.0 TECHNICAL SPECIFICATION CHANGES y
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TS figure 3.1-1A The proposed change increases the required shutdown margin from 3.5 to 4.0 percent delta k/k for the RCS cold leg temperature range zero to 350*F when any full-length CEA is fully withdrawn The increased shutdown margin will ensure that the TS are consistent with the safety analyses perf ormed for the Cycle 3 core and that the consequences of DBEs and anticipated operational occurrences are bounded by these analyses.
The proposed change is therefore acceptable.
TS Tables 3.1-2, 3.1-3, and 3.1-5 These tables provide frequencies for monitoring RCS boron concentration in the evert that one or both startup channel high neutron flux alarms are inoperable.
The proposed changes are more restrictive in that certain monitoring frequencies 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 inadverter.: buroh dilution, sufficier.t time will be availdble to terminate the event prior tc loss cf shutdown margin.
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 crd out of service, respectively.
The proposed change to figure 3.1-3 will prohibit insertion of regulating group 3 CEAs above 20 percent 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 grcup 3 CEb between 15 percent and 20 percent of rated thermal power.
The proposed revi:, ions are necessary to ensure consistency of the TS with the safety analyses perforrned for the Cycle 3 core.
These analyses demonstrate that reactor operation in accordance with the revised insertion limits will ensure that the Specified Acceptable fuel Design Limits (SAT 0Ls) will not be exceeded during the most limiting anticipated operational occurrence.
The croposed changes are, therefore, acceptable.
TS 3.2.7a TS 3.2.7a ensures that the actual value of the core average Axial Shape Index (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 average ASl from between.28 to +.28 to between. 27 to 4.27 to m6Le the TS censistent with the safety analyses performed for the Cycle 3 core.
The proposed change is, therefore, acceptable.
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TS Figures 3.2-2 and 3.2-2A 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 DNBR 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 anticipated operational occurrence.
The prcposed changes are necessary to ensure consistency of the TS with the saf ty analyses rcrforud for the Cycle 3 core and are, therefore, acceptable.
t 8.0 STARTUP TEST!NG The licensee has presented a brief description of the low power physics tests and the power ascension testing to be performed during Cycle 3 startup. Tre described tests will verify that core perfort.ance is consistent with the engineering design and safety analyses.
If the acceptarce 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 escalation.
If an unreviewed saf tty quotion is involved, the NPC will be notified.
The staff has reviewed the pra)osed startup test program for Cycle 3 and finds that it conforms to accepted practices and adequately supplements normal surveillance tests which are requircd by the plant Technical Specifications.
9.0 EVALUATION FINDINGS The staff has reviewed the fuels, physics, and thermal-hydraulics information presented in the PVNGS3 Cycle 3 reload report.
Also reviewed were the Technical Specification revisions, the startup test procedures, and the safety reanalyses.
Based on the evaluations given in the preceding sections, the staff finds the proposed reload acceptable.
10.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Arizona State official was notified of the proposed issuance of the amendment. The State official had no comments.
11.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a recuirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no signifi-cant 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 Cormission has previously issued a proposed finding that the amendrtent involves no significant hazards censideration,_ and there has been no public concent on such finding.
Accordingly, the amendment rocets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR
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, 31.22(b) ro environmental impact statenient or environmental assessment rieed be prepared in connection with the issuance of the amendment.
12.0 CONCLUSION
The Comission 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 ranner, (2) such activities will be conducted in compliance with the Comission's regulations, and (3) the issuance ci S:te amendment will not be inimical to the comon defense and security F.- to the health and safety of the public.
Principal Contributors:
L. Kopp K. Eccleston Date: May 20,1991 I
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