ML20004G022

Safety Evaluation Supporting Amend 24 to License NPF-6

ML20004G022
Person / Time
Site:	Arkansas Nuclear
Issue date:	06/19/1981
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20004G021	List: ML20004G020 ML20004G022 ML20004G023
References
NUDOCS 8106260498
Download: ML20004G022 (53)

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......f SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 24 TO FACILITY OPERATING LICENSE NO. NPF-6 ARKANSAS POWER AND LIGHT COMPANY ARKANSAS NUCLEAR ONE, UNIT NO. 2 DOCKET NO. 50-368 OG260

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TABLE OF CONTENTS

- 1.0 Introduction 2.0 Discussion and Evaluation 2.1 Cycle 2 Fuel Design 2.1.1 Thermal Performance Analytical Methods 2.1. 2 -

Cladding Creep Collapse 2.1. 3 '

Fuel Rod Bowing 2.1.4 Fuel ' Assembly Shoulder Gap 2.1.5 CEA and Fuel Assembly Guide Tube Integrity-2.1.6 Programmed CEA Insertion 2.1.7 Fuel Failures 2.1.8 General Fuel Assembly Surveillance and Grid Strap Damage 2.1.9 Surveillance Reporting Requirements 2.1.10 Demonstration and Test Fuel 2.1.11 Fuel Design Conclusion 4

2.2 Nuclear /.nalysis 2.2.1 Nuclear Parameters 2.2.2 Uncertainty in Nuclear Peaking Factors 2.3 Thermal Hydraulic Design 2.3.1 Statistical Combination of Uncertainties 2.3.2 CETOP-D Computer Code

? 3.3 CETOP-2 Computer Code 2.3.4 CE-1 Correlation 2.3.5 CPC/CEAC Sof tware Modifications and Phase II Test Results 2.3.6 The Most Limiting A00 2.3.7 Comparison of Cycle * ~ o Cycle 2

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-C-2.4 Accidents and Transients i

2.4.1 Ar.ticipated Operational Occurrences 2.4.1.1 CEA' Withdrawal 4

2.4.1.2 Full and Part Length CEA Drop 2.4.1.3 Fuel Misloading

- 2.4.1.4 Closure of One MSIY

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2.4.1.5 Baron Dilution 2.4.1.6 Loss of Load / Loss of Condenser Vacuum / Turbine Trip 2.4.1.7 Loss of Coolant Flow 2.4.2 Postulated Accidents 2.4.2.1 CEA Ejection 2.4.2.2 Seized RCP Shaft 2.4.3 Loss of Coolant Accident 2.4.4 Radiological Consequences of Postulated Accidents 2.4.4.1 Seize'd RCP Shaft

. 2.4.4.2 CEA Ejection 2.5 Reactor Protection System

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.2.5.1 Verification and Validation of CPC Sof tware Modifications 2.5.2 TS to Control Modifications to CPC Addressable Constants 2.5.3 Data Links Between CPC/CEAC and the Plant Computer 2.5.4 Monitoring of CPC Room Temperatures

2.6 License Conditions 4

2.6.1

'Fuc1 Performance 1

2.6.2 Instrument Trip Setpoint Drift A110wcnce l'

2.6.3 RCS Overpressure Mitigating System'-

i 2.6.4

'CEA Guide Tube Surveillance Program.

2.7 Other Matters 2.7.1 Co'ntainment Pressure, Temperature and Humidity.

2.7.2 High Pressurizer Pressure Trip Setpoint 3.0 Technical Specification Chunges

- 4.0 Physics -Testing 5.0 Environmental Consideration

- 6.0' Conclusions 7.0 References l

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. '1.0 INTP.0000TIOu By application'cated February 20, 1981 and March 5,1961, and supplemental information-as listed in the reference section of tnis report, *rkansas Power & Light Company (AP&L Co. or the licensee) recuestad ar. agendment to Facility Operating License No. NPF-6 for the Arkans** Nuclear One-Unit No.- 2 plant ( ANO-2_ or the facility). The amendment request consists of-

. Appendix A (Safety) Technical Specification.(TS) chances resulting from the analysis of the Cycle 2 reload fuel and other matters as discussed -in this report.

. Proposed changes to 't' e reactor protection system's core protection h

calculator system computer software to accommodate new methodo.ogy

.for calculating departure from nucleate boiling' ratio trip limits.

The associated specific TS changes are described in section 3.0 of this report. In addition this report addresses our evaluation of:

. The completion of the requirements of conditions to the license. elated to Fuel Performance, Instrument Trip Setpoints Drift Allowance, Over-

. pressure Mitigation: System, and the CEA Guide lube Surveillance Program.

These evaluations are presented in section 2.6 of this report.

. The issuance of TS changes for matters not necessarily related to the l

review of the reload analyses but whichLmay be conveniently addressed in this evaluation. These include TT, changes on (1) limiting the containment pressure, temperature and relative humidity so as to control containment differential pressure in the event of an inadvertent actuation of containment spray, (2) the high pressurizer pressure trip setpoint. These evaluations are presented in section 2.7 of this report.

l The information provided to support the staff's review of this reload and other issues included in this report are listed in the reference section (7.0) of this report.

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2-2.0 DISCUSSION AND EVALUATION We have reviOWed tne information provided in supper: cf the ANO-2 Cy:le 1 reloa to determine whether the design Objectives continue to be met and to determine whether the proposed reload changes have resulted in a reduc-tion of previously' approved design. margins. 0ur evaluations, as described in the following sections of this report, are complete for the-purposes of authorizing Cycle 2 ope. ation at the licensed full power level of 2815 l

MWt except for certain detailed matters within the thermal hydraulic review. The status of the thermal hydraulic review is discussed as follows.

By letter dated December 1,1980 (Ref. 2.3-1), AP&f Co submitt3d new methodology on the statistical combination of uncertainties in the calculation of the minimum DNB ratio, prepared by Combustion Engineering, Inc..(CE), for use in Cycle 2 and future ANO-2 reloads.

By letter dated January 9,1981, AP&L Co. submitted descriptions of revised software for the CPC/CEAC: system to implement the CE-1 departure from nucleate uailing ratio correlation for Cycle 2 and future cycles. These reports, in conjunction with rther-information submitted and in support of the Cycle 2 Reload Report, provide the basis for the Cycle 2 Limiting Conditions for j

. Operation (LCOs).

The staff has determined that insufficient time is available to complete all details of the review of these reports prior to the scheduled attain-ment of core criticality for Cycle 2 operation. AP&L Co. has been requested to provide additional -information to-enable the_staf_f to complete its review of the remaining details.

The nature of. the staff's concerns relates to whether or not sufficient margins have been represented in the core l

protection calculator system software changes for Cycle 2 to account for j

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the uncertainties ~ associated with the following:

(1) the CE-l DNBR l

correlation, (2) the CETOP-D code, (3) the CETOP-2 code, e d (4) the statistical combination of uncertainties. While the staff's review of all det.'

of these matters has not yet been completed, we have examined j

l' these issues in depth,'we judge the basic changes to be reasonable, and conclude that the completion of our review will not reveal the application of these changes for ANO-2 Cycle 2 operation to be significantly in error.

l Since these questions relate primarily to the adequacy of available thermal margins to account for anticipated operational occurrences (A00s) at full l

power conditions,we 'have concluded that it is acceptable for the plant to start up and operate at a reduced power. level for a short period pending the completion of our reviews. Operation at a reduced power level will provide additional thermal margins to account for the uncertair. ties discussed above while we complete our review. The licensee has submitted additional information on the Linear Power Level - High Trip required to limit operation to seventy percent of the licensed full power level of 2815 MWt.

Further details regarding these matters are presented in Sei Mon 2.3 of this l

repcrt. On the basis of the information discussed above, including the licensee's Linear Power Level - High Trip value which will provide additional protection to the plant from A00s at the reduced power level, we conclude that operation during this interim period at the reduced power level is acceptdble. Upon completion of our revicw of these matters,

.another SE will be issued.

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. 2.1 CYC! F 2 PJEL DESIGN 1

The ANC-2 Cycle 2 core will be comprised of 177 fuel assed11es of the 16x16 wunetry that were manufactured by Combustior. Engineering, the original MSSS vendcr. The major changes to the core for_ Cycle 2 are the removal of 60 Batch A fuel assemblies. These assemblies will be replaced.by 40 Batch D:assenblies and 20 Batch D* assedlies. The Cycle 2 core loading inventory is given in Table 1.

The Cyclu 1 fuel management pattern (Refs. I and 2) was developed to accomodar.e an E0C-1 core-average exposure of 12.5 GWd/t, which was the actual exposure achieved (Ref. 3). After the reload,,the B0C-2 exposure-will be 7.9 GWd/t, and the E0C-2 exposure is predicted to be 19.0 GWd/t.

The maximum EOC-2 exposure of any individual assembly will be 25.2 GWd/t.

Two Batch D fuel assemblies will serve as carriers for 42 DOE high-burnup -

demonstration rods (Ref. 4). Among the test rods are designs such ~as annular fuel pellets, large-grain-sized pellets, graphite coatings on cladding inner surfaces, ard. segmented fuel rods.

It is anticipated that tne performance information to be obtained from these tast rods will contribute to establishing the bases upon which future batch-average l exposures may be ir. creased to as much as 53 GWd/t.

All other fuel conprising Cycle 2 is of the standard FSAR ' design except 4 C-E test rods (Ref. 5) of a proprietary design. These rods are contained in a Batch 4 fuel assembly that was previously burned in Cycle:1.

Evaluation of the C-E 16x16 fuel mechanical design is based on engineering analyses, tests, and a substantial amount of in-reactor operating experience

- with previous 14x14 and 15x15 fuel designs.

In addition,.the performance of the design is subject to ' continuing surveillance'of operating reactors by C-F and _ licensees having C E NSSS plants. These programs continually provide current performance information.

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- 2.1.1 THERMAL PERFORMANCE ANALYTICAL METHODS The C-E fuel performance evaluation model called iATES is presented in the i

C-E topical report CENPD-139, " Fuel Evaluation Model " (Ref. 6). This model was used to calculate fuel temperature, stored energy, linear thermal output,

- and augmentation (power spike) factors.

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r ' In 1975, fter the ' approval' (Ref. 7) of CEMPD-139. information was made p

availabla to the NRC. that lead us to questi:- t.ic validity of fission gas release calculations in the_C-E model for fuel pallet burnups greater than l

20 ~GWd/t. ;Corbustion Engineering was inforrad I,wef. 8) of this concern I

and provided with a method of correcting fission gas release calcul.ations

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burnups greater than 20 GWd/t. Also, the ANO-2. license (Ref. 9) was

. conditioned to require resolution of this issue prior to the cycle in is which a pellet burnup of 20 GWd/t was achieved.

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'In. response to our concern, AP&LCo submitted (Rafs. I and 2)L a Cycle 2 reload

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ana',ysis -in which the NRC correction method has oeen_ used~ (Ref. 3) in l-all FATES analyses including that for the LOCA. Also, AP&LCo has performed l

. (Ref.- 10) a rod internal pressure analysis using the.present C-E fuel,

performance model with the _NRC correction for enhanced fission-gas release.

.The results with the NRC correction method show that-(a)-the ANO-2 fuel will-not exceed the LOCA acceptance criteria of 10 CFR 50.46, (b)~ rod internal pressure will' remain below nominal coolant system pressure throughout Cycle 2, and (c) other burnup-dependent analyses have implicitly accommodated

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. enhanced fission gas release.

We, therefore, conclude that enhanced fission gas release has been appro-priately considered for ANO-2 Cycle 2 operation.

2.1.2 CLADDING CREEP COLLAPSE Combustion Engineering has written a computer code that calculates time-to-collapse of Iircaloy cladding in a pressurized water reactor ' environment.

This code has been approved by the NRC 'and is described in the report CENPD-187, "CEPAN Method of Analyzing Creep Collapse of Oval Cladding"

-(Ref. 'l).

For Cycle' 2 operation, C-E has' performed time-to-cladding-collapse cal-(

culations using CEPAN and conservative input values of internal rod pressure,-

cladding dimensions, cladding temperature, and neutron flux. The results of i

this analysisishowed that the minimum time-to-collapse is in excess of the

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design batch-average discharge lifetime of the fuel, which will not be

exceeded during Cycle 2 operation.

The cladding collapse analysis' for the

'D0E demonstration and C-E test rods were included in the analysis discussed above.

We, therefore', conclude that the fuel rod cladding collapse analysis for LANO-2 Cycle 2 operation is acceptable.

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. 2.1.3' FUEL R0D BOWING Because fuel rod bowing _in pressurized water reactor' Jfects neutronic and thermal-hydraulic safety margins, AP&LCo na; analyzed the anticipated cr. tent cf red bewing in Cycle 2.

In the :.asysis, AD&LCo has referenced tne C-E topical report CENPD-225, " Fuel and Poison Rod Bowing" (Ref.12).

l The' staff has not yet approved the CENPD-225 report. Accordingly, it is i

the staff position that the rod bow compensation currently specified in Technical Specification 4.2.4.4 shall remain applicable for initial Cycle 2 operation. We estimate that the peak Sundle average burnup will be 20.2.GWD/t by the end of November 1981 when the rod bow compensation review is expected to be complete.

The rod bow compensation required for that burnup is 11.4 percent, compared to the proposed 2%, of the ENBR limit value.

The difference in DNBR limit due to rod bow compensation methodology should by compensated by an equivalent increase in. power uncertainty factor BERR 1.

Based on the sensitivity study provided in the response to NRC questions 492,66, a relationship between the BERR 1 and DNBR limit is established.

Using the most conservative value of -0.6 for the derivative of-percentage BERR 1 with respect to the percentage ENBR, we estimate the.BERR 1 value should be increased by 5.6% to account for the rod bow compensations.

We anticipate our approval of the topical report CENPD-225 by November l

1981. At that time, AP&LCo may amend their Technical Specifications to l

' reflect any reduction in the rod bowing penalty that is possible from the application of the CENPD-225 methodology.

We thus conclude that the effects of fuel rod bowing have been adequately addressed for Cycle 2 operation.

2.1.4 FUEL ASSEMBLY SHOULDER GAP During irradiation, fuel rods and fuel assembly guide tubes undergo axial l

growth at different rates. To ensure that an adequate design shoulder gap exists for the fuel assemblies that will comprise the Cycle 2 core, AP&LCo has made a calculation (Ref. 3) on the lead-burnup fuel rod in a Batch B fuel assembly.

The calculation of the minumum shoulder gap in the Batch B fuel assembly was

.gerformed with the methods ' described in the C-E topical report, CENPD-198, l

Zircaloy Growth In-Reactor Dimensional Changes in Zircaloy-4 Fuel Assemblies" (Refs.17,18, and 19).

neutron fluence to 4X10',The calculation was made for axially averaged fast j

neutrons per square centimeter, which corresponds l

to a maximum assembly exposure of.22.5 GWd/t, as specified in our approval i

(Ref. 20) on the CENPD-198 methodology.

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For calculating differential growth at exposures beyond.2?.5 GWd/t, a more conservative method, which is acceptable (Ref. 21), was utilized. The results showed that no interference Detween fuel rods and the upper end fitting is l

predicted for Cycle 2 operation. Furthemore, during the current refueling l

outage, shoulder gap inspection of the Cycle 1 characterized fuel assemblies verified the acceptability.of the gap calculation.

Therefore, we conclude that an adequate fuel assembly shoulder gap will be maintained for Cycle 2 operation.

2.1.5 CEA AND' FUEL ASSEMBLY-GUIDE TUBE INTEGRITY Fretting wear has been observed (for example see Refs. 22, 23, 24, and 25) l.

in irradiated fuel assemblies taken from operating C-E reactors. These observations revealed unexpected degradation of guide tubes that were under control element assemblies.

It was concluded that coolant turbulence was responsible for vibration of the normally fully withdrawn control rods and, where these. vibrating rods were in contact with the inner surface of the guide tubes, wearing of the guide tube walls took place.

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As a remedy, AP&LCo has installed scupper extensions on the upper. guide structure flow channels and stainless-steel sleeves in all fuel assembly guide tubes to be used in CEA positions. The 4 Batch A unsleeved test assemblies (Ref. 26) that resided under CEAs in Cylce 1 will not be used in Cycle 2.

Our review of the scupper extensions and the sleeving program has been documented in the ANO-2 safety evaluation (Ref. 26).. Our prior safety evaluation concluded that scupper extensions and guide tube sleeves will perform their function of mitigating fretting wear in fuel. assemblies.

Furthermore, to provide assurance of guide tube and sleeve integrity. th's licensee perfomed an EOC-1 guide tube surveillance. program (Ref. 27).

Eddy current testing was performed on all of the 9t:ide tubes in 5 fuel assemblies from Batch A and 5 fuel assemblies from Batch C.

These assemblies were spatially ' selected from the Cycle 1 core on the basis of where maximum wear might be expected to occur. The results (Ref. 28) indicate that the total amount of wear is negligible and that sleeves remain intact.

We conclude that the sleeved guide tubes in the Cycle 2 fuel assemblies continue to meet their design functions and are therefore acceptable.

1 Based on the reported favorable surveillance results and continued use of guide tube sleeves under all CEAs, we consider the issue of guide tube wear resolved for future cycles of ANO-2.

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l While the stainless steel. sleeves nave orecluded guide ~ tube wear, they

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have. probably increased the cladding wear.that occurs on the control rods tnemselves. Therefore, during the Cyc.16 2 outage, eddy current' testing with-an encircling probe was performed on 8 CEAs. The results (Ref. 28) were l'

consistent with similar measurements un CEAs from C-E NSSS reactors -

using 14X14. fuel designs af ter one operating cycle. Since the measured l

wear is within the limits for continued CEA operation, it is therefore acceptable.

l To date, no inspections have revealed CEA cladding' wear: rates that would' indicate a-potential for the loss of CE.'s henniticity in the near future. I t, nevertheless, remains uncertain as to whetner wear degradation to CEAs could ultimately' reduce the CEA design lifetime. 'We can,. however, conclude I.

that for Cycle 2 operation, fretting wear to CEA cladding will remain at acceptably low levels.

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'2.1.6 PROGRAMMED CEA INSERTION' During ANO-2 Cycle 2 operation, AP&LCo will _ continue programmed CEA

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- i nsertion.~ -This program was approved (Ref. 26) and instituted during Cycle 1 operation so that the magnitude of guide tube wear at any one

lacation would be-reduced by repositioning fully withdrawn-CEAs. -

i j Specifically, the. full _-out insertion limits for the CEAs are extended l

3 inches into the core.)

l We believe that' this method of apportioning the wear is acceptable, l

though not necessary, because the guide tube sleeve wear reported (Ref. 28) in ANO-2 is insignificant and the Cycle 2 core will contain no_unsleeved assemblies. Because increased axial peaking of about' 4% occurs when all of the CEAs are inserted to the 3-inch full-out insertion limit, AP&LCo might want to consider discontinuing this program in future cycles.

2.1.7 _F_UEL FAILURES l

In January 1930 AP&LCo determined from primary coolant activity that L

a limited number of fuel rods had perforated in ANO-2.

The failures were detected.over a 1 to 3 day period while the plant was in the initial power ascension program. Prior to the occurrence of these failures,._ the-testing program included preconditioning at 80% power, dropped-control-rod testing at 50% power, and then a ramp to 651, i

power. Following the ramp to 65% power, xenon oscillations were

-observed coincident with increased coolant activities.

Dur 1g Cycle 1, the licensee and fuel vendor were unaware of the specific nature of tce failures, but had tentatively ascertained that the. fuel was operated within. the recommended operating restrictions inasmuch as (a) the fuel was preconditioned prior to the occurrence of failures and -(b) the rate of power ramp just prior to failure occurrence was substantially less than that allowed by the operating limits.

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. During the. outage all 177 assennites in the core were sipped and a

-total of 7. assemblies were found to contain leaking fuel rods (Ref. 29).

i These leakers were distributed among.2 Batch A assemblies, 3 Batch B.

assemblies,-and 2 Batch C assemblies. Since visual -inspection was not successful-in locating the failed rods, each of the rods in the leaking Batch B and C fuel assemblies was removed and eddy current

. tested to identify the failed rods.

-A. total of 14 failed fuel rods-

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were found in these 5 assemblies, which were planned for reuse during Cycle 2..In addition, one poison rod in a Batch C assembly was found to be perforated. Also, 52 additional fuel rods showed questionable j

eddy current indications.

Consequently,' the 14 failed and 52 questionable fuel rods were replaced by 66 fuel rods that were extracted from a sound Batch A fuel assembly.

The _ perforated poison rod was replaced with a solid Zircaloy "dumny" rod. - Finally, prior to: reinsertion into the core, all 5 reconstituted 4

fuel assemblies were sipped to ensure the absence of leakers.

Our interest in this is' sue is based on three fundamental concerns.

First, that coolant activity. levels be kept as low as reasonably achievable.and within the Technical Specification limit and safety analysis assumptions. Second, that the cause of the failures be investigated so that such failures can be minimized or eliminated.-

Third, that NRC receive.progt notification-of such failures-so that (a) operators of similar plants can be informed and (b) NRC can watch for conmon trends and generic problems.

In regard to the first concern, the licensee has replaced the failed fuel rods-with non-failed fuel rods of lesser enrichment and the failed j

' poison rod with a solid Zircaloy rod. Since the licensee has determined that these substitutions do not violate the Cycle 2 physics analysis, l'

we find these actions to be appropriate.

l In-regard to the second concern, the licensee has not cogleted the investigation.and, therefore, has not determined the cause of all failures.

. Preliminary indications are that (a) some of the fuel rod failures were caused by fretting wear from foreign material lodged between lawer end fitting flow

~ lates and bottom Inconel grid structures, and (b) the poison rod failure pmay be due to fretting. The license 3 is continuing the investigation and will report the findings in a writtea report that was scheduled for submittal in 90 days (see further discussion in Section 2.1.9 with regard to this schedule). The second concern is, therefore, being adequately addressed.

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-3 In. regard to the. third concern,.AP&LCo'nas agreed to issue (a)'3 report 1 describing the present damaged' fuel and (b) a letter to NRC during Cycle 2 ; operation. if additional fuel failures are inferred from variations

. in the equilibrium' primary coolant activity level. Consequently,

' AP&LCo has satisfied the third concern.

2.1.8 GENERAL FUEL ASSEMBLY SURVEILLANCE AND GRID STRAP DAMAGE' The fuel-surveillance' program that was' described in Section 4.2.1.1.10

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of the FSAR included the visual examination of all the initial core fuel assemblies. ~ Approximately 15 fuel assemblies were to be' inspected

. prior. to reactor startup, and the visual examination of the balance

-of the discharged' assemblies was to be perfomed later..-

The licensee reported (Ref. 28) preliminary results of the 15 fuel assemblies that were inspected by TV camera or periscope. There were no abnormalities' observed from these. assembly inspections.

However, in later performing the visual examinations of t.he remaining 45 discharged Batch A. fuel-assemblies, AP&LCo observed 5. assemblies

.having grid strap damage. ' This information was verbally conveyed '

l to NRC on May 27 and followed up by letter of June 4,1981 (Ref. 29).

Of the 5-assemblies having grid strap damage, 2 had:relatively minor damage chat was confined to missing tabs, while.each of the other 3 had

-significar.t damage that consisted of a missing section of one of the grid perimeter straps on that assembly. The damage apparently occurred during the Cycle 2 outage because (a) the fracture surfaces were shiny and not oxidized like the remainder of the undamaged grid surfaces and (b)'fuelirods t adjacent to the missing grid strap sections had abrasion markings that corresponded tr finger spring locatlons thus indicating the presence of intact. grid, traps during Cycle 1 operation.

. Because the grid strap damage was not detected until af ter the core was reloaded, the number of assemblies with damaged grids - remaining in the core

-in unknown, but estimated (assuming a random damage distribution) by C-E to be limited.to 16.

~ Our interest in this issue is based on three concerns. First, that the cause of the grid strap damage be determined and eliminated or the effects be reduced. Second, that any grid strap damage present during

' the next cycle of operation not result in unacceptable damage such as additional fuel failures. Third, that NRC receive notification of such occurrences in *.he future.

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In regard to the first con-m : M".LCn has not yet determined the exact cause of grid strap tearing.

iicniever, as discussed above, AP!.LCc is confident that the damage occurred during the Cycle 2 outsge fuel shuffle.

Thisiconclusion is1further sucoorted by AP&LCo's review of refueling experiences and procedures u::::d during the outage.

From personnel interviews, it was learned that refueling machine overlrad/ underload trips-occurred frequently during withdrawal-and replacement of fuel assemblies. Since the overload / underload trip set points were believed-to be adequate to prevent fuel damage, routine' procedure was followed-Lafter such trips occurred. That orocedure consisted of reestablishing normal. loading by raising or lowering the assembly and then manually shaking the hoist cable. -(This procedure had been found successful for.

similar occurrences in ANO-1. )

The licensee is continuing this investigation to (a) quantify post-trip

- fuel assembly loading that occurs due to system momentum, (b) determine the effect of. cable shaking, and (c) determine the loading required to ~

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cause grid. damage.

The first concern is thus being adequately addressed.

In regard to the second concern,. if-grid strap damcge is present in the Cycle 2 core, there are two topics of interest:

(1) the potential for fretting or fatigue damage to fuel and poison rods that might be inade-quately' supported in the vicinity of damaged grid sections and (2) the potential-for: problems associated with loose grid pieces, including the possibIlit'y.of flow blockage with attendant departure from nucleate boiling.

Based on C-E out-of-pile' tests on 16X16 fuel bundles,. the licensee does not' expect rods that are inadequately ' supported at one grid elevation to fail. We.are not familiar with the specific tests to which AP&LCo

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has referred; however, it is most unlikely that these tests employed simulated grid damage.. Consequently, we have' no opinion on the applicability of these tests.

Nevertheless, C-E has conservatively estimated that there are less than 24 fuel rods which are adjacent to damaged grid sections. Hence, the number of potential failures is limited.

Concerni.ng potential problems due to loose grid pieces, the licensee has-postulated that-limited fuel rod failure could occur due to fretting at locations where grid pieces might become lodged in the fuel region. We agree that such a failure mechanism is conceivable although this type of fretting wear has not previously been observed and would be local and confined to a few rods at most.

Since (a) the rate of fuel failures due to fatigue or fretting would be slow and detectable by the letdown monitors or periodic primary coolant l

sanpling, (b) the number of rods involved is small, and (c) this hypo-thetical assessment seems conservative, we conclude that the second' concern is satified for Cycle 2 operation.

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lInLreaard.'to the last concern, the licenses has agreed (as discus:ed above -in' Section 2.1.7) to notif.y NRCL by. ietter in the event that fuci failures are detected' during Cycle 2 operation. - Also, with respect lto' the general fuel assembly inspections.at EOC-2, tM licensee has agreec to; performing these visual inspections prior tof sealing up the reactor vessel-(the FSAR commitment reads only " prior to reactor re-rtartup").

Performing _these inspections ~ prior to. sealing up the system t=ich_was 21:0

the case for this' outage) will assure flexibility :for inspecting additional-

' fuel assemblies in the event 1that such'is' warranted from the'observaticns on 7

-the selected 15 (or-more) fuel assemblies. We, therefore, conclude that f

.the third: concern will. be handled appropriately.

sin ' light of the;above discussions,Lwe-find that the licensee has satisfied the Cycle 2xoutage. surveillance requirements for pre-startup reporting of f fuel' assembly inspections.

1

~ 2.1 '. 9 SURVEILLANCE REPORTING REQUIREMENTS l

According to a commitment in the FSAR, AP&LCo is also to provide a. final

- report-describing the results of fuel inspections within _90 days following

. However,:in light of (a) the unexpected' damage (i.e.,~ torn refueling.

gridJstraps, fuel and poison rod perforations) observed after Cycle 1, (b) the preliminary nature of the conclusions. attained to date, -(c)' the need for additional inspections to: establish conclusive damage. mechanisms that' were operative during Cycle 1.and the refueling outage, and (d) the time required to fonnulate preventative measures to be' employed in the future, we will relax the 90 day reporting requirement for-this specific

~

i nformation. Nevertheless, AP&LCo should strive to submit this'information as soon as it is available and certainly no later than 90 days prior to the next refueling outage.

'2.1.10 DEMONSTRATION AND TEST FUEL We find the use of the 00E demonstration fuel rods (Ref. 4) in the 2 L

Batch 0 fuel ' assemblies ~ acceptable since (a) they contain few fuel rods I

in number and thus constitute a small portion of the Cycle 2 core,

~

(b) they are to be placed in non-limiting positions, and (c) they have

- been analyzed with approved methods as were the standard fuel which will comprise the core. Furthmore we encourage such demonstration and test L

programs because they tend to lead to improved design and safety analyses of fuel performance.

Upon the same bases,'we find the continued use of the C-E test fuel rods (Ref. 5)-in the 2 Batch C fuel assemblies acceptable as well.

2.1.11 FUEL DESIGN CONCLUSION i

l

.We have reviewed the AP&LCo reload analysis (Refs.1 and 2) and supporting information (Refs. 3,10, 27, 28, and 29) submitted as justification for Cycle 2 i

l-

- operation of ANO-2. We have determined that all applicable requirements related to the reactor fuel design have been met. Therefore, we conclude i

that AP&LCo's application is acceptable.

lL. --

_..._...____..~..-.__._._._.._.._.._..,_..__a,__,_._..

A

- Ti -

TABLE 1 ANO-2. CYCLE 2 CORE LOADING INVENTORY-1 Initial BOC-EOC Assembly Number of Enrichment - Burnup Average Burnup Average Designation Assemblies w/o U235 (GWd/t)

(GWd/t)

A 1

1.93 13 2 21.0 B

60 2.27 14.1

'24.5 C

56 1.94 9.7 21.6 0

40 3.48 0

9.7 3.03 D*

20

_3.03

.0 13.5 2.73 177 l

l I

l

~~

.- 2.2 NUCLEAR ANALYSIS 4

The nuclear design analycis used in Cycle 1 (reference cycle) nas oeen

_ erformed with the P0007 (fine mesh) co@ uter code. The Cycle 2 analysis p

is based on the ROCS (coarse mesh) code in a manner consistent witn Calvert l

Cliffs Units 1 and 2 and St. Lucie Unit 1.

The ROCS code is considered as i

a " state-of-the-art" code which has been used extensively by Combustion

. Engineering, Inc. (CE) ano is acceptable.

The use of ROCS has l

been limited to the calculation of three dimensional effects while local power peaking'is calculatea with PDQ. Few-group cross sections for input to both codes have been cc:=uted using the DIT code, a nultigroup transport theory code. -The following safety parameters were calculated:

- Critical Baron Concentrations,

- Boron Worths,

- Moderator Tegerature Coefficient,

- Doppler Coefficient,

- Total Delayed Neutron Fraction, S,ff,

- Neutron Generation Time, 2,*,

- Available CEA Worths, and

- Required Worth Allowances.

2.2.1 NUCLEAR PARAMETERS l

l The expected Cycle 1 termination burnup is 12,500 MWD /MT and the corres-l ponding expected Cycle 2 full power operation burnup is 10,500 MWD /MT.

For both rodded (partial length control element assemblies (CEAs), bank 6) and unrodded configurations, the maximum power peaks occur at the 80C-2.

l Augmentation factors for Cycle 2 have been calculated and they include l

the effects of fuel densification, radial pin power distribution, single l-gap peaking factors, and burnup.

l

.The Cycle 2 planar radial peaking factor uncertainty is 5.3 percent and is based on the topical report CENPD-153-P, Rev.1-P-A which is an NRC approved report and is, therefore, acceptable. This value is conservative with respect to the maximum value of the reference cycle.

The Cycle 2 moderator coefficient is calculated to be

.5 X 10-" ao/*F at B0C and -2.3 X 10-" ap/*F at E0C. These values are bounded by the reference cycle values (i.e.,

.5 and -2.5 X.10-* ao/*F). The fuel temperature coefficient-(Doppler) values for Cycle 2 are slightly more negative than the values of Cycle 1.

However, the extended pointwise Doppler feedback technique-has been used which involves use of iterations of pointwise l

power distribution and pointwise fuel te@erature instead of using l

precalculated fuel tegeratures.

It is estimated that the Cycle 2 Doppler coefficient values are more accurate.

We find the moderator and

. the fuel te@erature coefficient values to be acceptable.

L

. At. the beginning of Cycle 2 (B0C-2), the reactivity worth of all CEAs inserted (assuming the highest worth CEA is stuck out) is 9.3 percent ao.

The reactivity worth required for shutdown including the power cefect from hot full power to hot zero power and the CEA bite (i.e., the fact that CEAs

.may be slightly inserted instead of being fullly withdrawn) is 2.5 percent ao.. The excess CEA worth available for normal shutdown is, therafore, 6.8 percent ao. At the end of ~ Cycle 2 (E0C-2) the corresponding excess CEA worth is 6.6 percent. The required shutdown margin is 5 percent ac.' hence, the available margin is negative and more than. adequate to account for possible uncertainities. We find these shutdown margins acceptable.

The consequences of a dropped CEA were analyzed. The limiting safety analysis values for dropped CEA increase in radial peaking factor is 17 percent for Cycle 2 compared to 27 percent for the reference cycle.

~

However. the Cycle 2 value is conservative compared to the actual calculated values and is acceptable.

t L

2.2.2 UNCERTAINTY IN NUCLEAR PEAY.ING FACTORS l

.Incore detector measurements are used to compute the core peaking factors

~

l using the INCA code. The methodalogy, the required coefficients and the reduction are described in the approved topical report. As mentioned above, the planar! radial power distribution measurenent uncertainty is 5.3 percent l'

and will.t>e applied in Cycie 2 to COLSS'and the CPC on-line calculations.

On this-basis we find these measurement uncertainties to be acceptable.

l i

l i

i l

l l

.. L 2.3' THERMAL AND !;YDRAULIC DESIGM We have reviewed the Cycle 2 reload to confirm that the thermal and

hydraulic design cf the reload core has been accomplished using accept-able analytical methods and provides acceptable margin-of safety from conditions which would lead to fuel damage during nonnal reactor operation and anticipated operational transients.

ANO-2l Cycle 2 consists oi presently operating Batch A, B, and C assemblies, along with fresh Batch D assemblies. 'The Cycle 1 termination burnup has been assumed to be approximately 12,500 MWD /t. 0ur review consisted of?

the following: -(a) statistical combination of uncertainties to calculate minimum DNBR; (b) CETOP-D and CETOP-2 thermal hydraulic (T-H) computer codes used for DNBR analysis; (c) CE-1 correlation used for DNBR analysis;

.(d) effects of fuel rod bow on DNDR. margin.(see Section 2.1); (e) compar-ison of the Cycle 2 thermal-hydraulic parameters at full power with those of Cycle 1; (f) CPC and CEAC software modifications; (g) addition of asym-

. metric steam generator transient protection trip in CPC based on instan-taneous closure of a single MSIV (ICSM) transient (see Section 2.4);

c (h) determination and evaluation of the most limiting event from the design basis Anticipated Operational Occurrences (A00s) for which a DNB trip occurs and the thermal margin LCOs is maintained; and (i) proposed Technical Specifications modifications.

2.3.1 STATISTICAL COMBINATION OF UNCERTAINTIES The criteria established in 10 CFR 50, Appendix A, imposes the require-ment of a high degree of assurance that neither the phenomenon known as DNB (Departure from Nucleate Boiling) nor melting at the fuel center-l line occurs. Calculational methods have evolved over the. years that-predict the conditions causing the phenomena. The results of the cal-culations are then entered into the reactor protection system method-logy to provide the necessary assurance neither occurs.

There is a degree of uncertainty in the knowledge of the exact value of each of the variables used. This uncertainty has been handled in tne past by assuming that each variable is at its extreme most adverse limit

(

of its uncertainity range. The assumption that all factors affecting DMB and fuel centerline temperatures are simaltaneously at their most adverse value is very unlikely and leads to conservative restrictions on reactor operation. The potential for greater operational flexibility has provided a strong incentive to reduce the degree of conservatism.

(

. The licenses has proposed use of a new methodology that reduces the conservatism L, atatistically combining the uncertair. ties. Thetreport:

CEN-139(A)-P. (Rei. 2.3-1)~ describes the methodology to calculate new

-MDNBR limits-for AMO-2.

It ensures with at least 95 parcent probability and 95 percent _ confidence level that:0NB will not occur.

CEN-139(A)-P: describes methods used to statistically combine uncertainties.

in.those variables which are not monitored while the reactor is in operation.

The methods are tnen used to develop ainew MDNBR. The variables so-considered are termed system variables and include such things as reactor geometry,-pin-by-pin power oistributions, and inlet and outlet flow boundary conditions.. The variables affecting DNB.whose uncertainties a're not con-sidered are those which are monitored during-reactor operation and are termed state variables.

Though it is not_specifically stated, the state variables are considered in the CPC and are described.in other documentation supporting operation.of ANO-2.

The licensee proposes that the difference in.the basic.DNBR limit value of 1.19 discussed in the section on the CE-1 correlation and the limit value of 1.24 is sufficient'to account for these uncertainties. Our~

. review of SCU has not progressed sufficiently.to enable us to make a

~

finding on the precise value of the thermal margin credit gained by

-inclusion of:SCU in the DNBR limit. We are currently reviewing the individual uncertainty components of-the system parameters to evaluate the l

SCU credit.

c L

'However, we have reviewed tn. basic-SCU methodology and find it' acceptable.

Upon'. completion of"our review, if it is required. we will require that anf reduction in the credit currently proposed for SCV by the' licensee be accounted for prior to authorizing full power operation.

If necessary l-this would most likely_be done through adjustment of the addressable constant s

on power uncertainity, BERR 1.

The' licensee has proposed an interim value

'of 1.112 for BERR 1 and we find this vaule should be adjusted upward by 5.6% to' account for rod bow compensation described in the technical spec-ifications. We, therefore, conclude that the BERR 1 value of 1.174 should beEused.for the interim period of operation ~of AN0-2 at reduced power level.

2.3.2 'CETOP-0 COMPUTER CODE

-The CETOP-D computer code is used as a core thermal margin design analysis tool for. the Cycle 2 reload. CETOP-D is an open lattice thermal hydraulic code which solves the same conservation equations and uses Se same con-

, stitutive equations as in the TORC code (CENPD-161-P). TORC, derivad from COBRA-III C is a multi-stage thermal margin code. Based on the cude of the changes in the CETOP-D code relative to its predecessors se under-took a complete review of the CETOP-D code as a design analysis tool, l.

i l

I a.,.

- - - -. -.. -.. -. -. ~ -. -. -. -.. - -. -

1

- Yo -

-Although our review of the.CETOP-D_ code is'near cc=plet kn the details-of our evaluation of the code' have not' been finali::ad. Our su :ary l

evaluation of the code.is-based on comparisons provided by the_ licensee

-of CETOP-D results to TORC results.over a wide spectrum F operating conditions'for ANO-2, Calvert Cliffs Units l_and 2,-and San Onofr.e Units-12.and 3.

In all= cases, referenced in the response to cuestions 492.7 and 492.68, the CETOP-D code predicts the minimum DNBR to be lower-l than does TORC. Since we have previously approved TORC for use in CE thermal-margin plant analyses we conclude, based on tne conservatism of

.CETOP-D relative to TORC, that the CETOP-D code is acceptable for ANO-2 thermal margin calculations, with the condition that tne'not assembly inlet flow factor with the value described in response to MRC cuestion

~

492.14, or a smaller value, be used.

- 2.3.3 ' CETOP-2 COMPUTER CODE =

The staff has_ reviewed the CETOP-2 functional specification and has performed an audit of the functional' tests of the integrated system to assure that CETOP-2 with the algorithm uncertainty factor is programmed properly and predicts minimum DNBR conservatively.

Th'e CETOP-2 functional description is provided in'the Appendix B of CEN-143. The following'is a. summary of-the results of our review:

'(a) Errors,have been discovered in the-Martinelli-Nelson void fraction correlation and the two-phase friction factor multiplier. However,.

.the errors have been identified as just typographical errors and

- are prograntned properly. Therefore, these errors are nonconsequen-tial.

(b), The single-phase friction factor calculation using the Blasius l

correlation, where the friction factor is a function of Reynolds l

number, has'been studied. Since ANO-2 fuel cladding surface rough-ness ranges from 14 to 21 micro' inches RMS, the calculated friction factor agrees with the Moody friction factor within three percent in the normal opgrating condition range where the Reynolds number is around 5 x 103 Therefore, the friction factor calculation using'the Blasius correlation is acceptable.

l (c)

In order to reduce the.CPC execution time, many friction factor and l

two-phase multiplier calculation algorithms have been converted i

from exponential functions to polynomial fits. The staff has l

examined'the accuracy of these conversions and found them acceptable.

(d) CETOP-2 uses lumped channel modeling wherein the core is divided into four modeling channels, i.e., core region channel, hot assembly l

channel, buffer channel, and hot channel. The hot channel is a pseudo-hot channel which models a corner guide tube subchannel.

The staff has raised questions (Ref. 3-2) as to how the hot' channel is selected; whether the selected hot channel always predicts the lowest ONBR; whether minimum DN8R always occurs in a guide tube

+---e-u-t m r-N g

--Mv3

-geyt i *tr

• g e-w tw ee-r%9r7-*-p*e--

e qPyg---y----Sumt gm+>9 er e+f-r=to T"-Pee

=w hk'e

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.ci 46661, a6d nhsth6r it is legitimata to usa a guida tuba channal i

to reuresenL oL.ner crioririte is writ:re - the risinliiiuus UGors iniu'it occur. - To r

l ans.;cr these questions, the licensee has stated that the modeling-l is indecendent of the actual location of the hot assembly and hot j

channe!_within the core. An inlet flow split. factor for the hot i

.assad ly is used to yield conservative DNBR predictions relative-to the detailed-TORC code. The inlet flow-split factor is obtained l

.from the reactor model flow test experiment. During ooerating.

transients, the flow split may change significantly. However, the most adverse of the flow splits has been used in the CETOP-2. The l

= inlet flow split factor is described in Table B-2 of CEN-143, plant--

l specific connants for ANO-2. As for the legitimacy of_using a guide L

. tube subchannel,- the licensee has stated that-the present fuel l

management schemes result in power distributions which produce the l

largest pin peaks near guide tube water holes throughout the core l

life..The cold wall correction factor'in the CE-l CHF correlation l'

is also used to reduce the predicted DNBR in the guide tube channels, l'

As a result,-the minimum DNBR will always be predicted to occur L.

in a corner guide tube channel. The staff concludes that.the pseudo-hot channel modeling is acceptable, provided that the fuel management scheme ensures that the calculated minimum DNBR always occurs at a guide tube subchannel.

(e)

In.the lumped channel modeling, transport coerr'icients are used to account for the fact that the coolant properties associated with t

turbulent mixing and diversion crossflow between adjacent-channels

~

are not the lumped channel average values. Constant values of transport coefficients are used in-the~CETOP-2. ~In response to the staff question 492.3,'the licensee has provided a sensitivity l

study of the DNBR with respect to the transport coefficients. The l

DNBR has-been shown to be insensitive to the pressure transport P

coefficient. However, the enthalpy transport coefficient has been shown to_have a significant effect on the hot channel enthalpy.

In CETOP-0, an enthalpy transport coefficient is calculated for each axial level. The value chosen for the CETOP-2 is such'that the CETOP-2 results match the CETOP-0 results for a typical axial power distribution and nominal operating conditions. Any errors resulting f

.from this simplification are covered by an algorithm penalty factor on core power.

(f) The algorithm uncertainty factor is a compensation applied to the core power in CPCs to ensure that the DNBR results from CETOP-2

-are conservative relative to CETOP-D. The licensee has had 6400 cases run of comparisons between CETOP-2 and CETOP-0; and a compensation factor has been-derived so-that application of the compensation factor to the core power results in a 95/95 probability / confidence level that CETOP-2 is more conservative than CETOP-0. These cases are run using the Value of BERR 1 equal to 1.0.

Using the algorithm uncertainty power compensation factor or.a larger value as a core power multiplier will result in a conservative DNBR prediction from l

CETOP-2.

(g) Based on the compensation factor being built into the CETOP-2 software E

as a plant specific constant and our other findings as reported in

-(a) through (f) above, the staff concludes that the CETOP-2 code design as applied to ANO-2 is acceptable.

. ~.-.

1 u.

~2.3.4 rF I 'CURHtELAIlON k

. For_ANO-2 Cycle 2.the critical heat-flux correlation (CHF)'has been changed from tne W-3 correlation to theLCE-1: correlation. 'The CE-1-ecrrelation-

' has'previously been approved by the staff for interim plant specific

- application witn a minimum DNBR. limit of-1.19. ";asad cn the results of our_ review of:ANO-2 Cycle 2' operation we conicude that the value of 1.19 is consistent with the submitted data base. :Therefore, the_DNBR

. limit for tha.CE-1~ correlation is 1.19 before any other! uncertainties are

- accountec -: r.

i This value of 1.19 is consistent with.the licensee's proposal and~is acceptable.. Tne accounting of other uncertainties,Lsuch as SCU and rod bow-and the final value ofEthe limit for ANO-2 Cycle 2 (1.24) are discussed-

-in other sections'of this report.

U 2;3.5 CPC/CEACl SOFTWARE MODIFICATIONS AND PHASE II TEST RESULTS:

o The Core Protection Calculators (CPC) and Control Element-Assembly

- Calculators -(CEAC) of the ANO-2 Cycle 2 are ' basically identical hard-ware with a modified version of the software from that of Cycle 1.

The major software modification includes (i) the use _of-CETOP-2 in! place of-

~ ~

~

CPCTH for core thermal hydraulic?chlculations; (ii)~ replacing W-3 with ~

i CE-1 correlation 'for CHF calculations; (iii) the use_ of a statistical-

.' combination of uncertainties ~(SCU)itted a summary of. the CPC/CEA.. soft-of system parame DNBR limit; The licensee has subm ware modifications-over that of Cycle 1 (CEN-143(A)-P.

Since the' Cycle 1 CPC/CEAC had been reviewed extensively and approved, L

the staff's review effort of the Cycle 2 CPC/CEAC has been concentrated on the software modification.

1 The implementation of.the Cycle' 2 Reload modifications into the CPC p

system has been examined through the utilization of Phase II testing.

L The. primary objective of tne Phase II testing -is to verify that the CPC and.CEAC-' software modifications have been properly integrated with the CPC and CEAC software and the system hardware. The testing also provides j

- confirmation that the static and dynamic operation' of the integrated system as modified is consistent with that predicted by design analysis.

l The' objectives are achieved by comparing the response of the integrated ta l system to the response predicted by the CPC FORTRAN' simulation code.

The applicaat has submitted the CPC F.e.se II test report.

In the Dynamic Software Verification Test (DSVT), 40 transient cases, ranging from four-l

- pump loss of flow to CEA withdrawal and primary system depressurization

. transient, have been run on both the FORTRAN Simulation and the CPC soft-ware'on sinole channel test facility.

r l

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g ywT-g wmy 90-'ywwwg y + g a-qi,.g 9 4 y ywe,ywy g4-v.g--

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• p grTy W-vd Marte s= =wv e 4W G -W""S"e r eweww + e afir'agi*A-m au--W

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- The resulting initial 1DNBR, initial LPD and the trip times from the-single channel test; fall well within the accept:n:: criteria for each case established from the FORTRAN simulation run:;. ~ For a few cases where the trip time fails to stay within the acceptance criteria, the

~

cause has been identified to be the differences in-interpolation 6f time dependent parameters between the single ^hannel and FORTRAN simulation.

The staff-tas made uan audit on the. Phase II test and confirmed the accuracy of.the report. The agreement of the Phase II testing has l

shown the adequacy of the implementation of the functional specification.

Therefore, the staff concludes that the software modification implementation is acceptable.

2.3.6 THE MOST LIMITING EVENT'FOR WHICH A DNB TRIP OCCURS AND THERMAL MARGIN MAINTAINED:

l' Results of the analyses performed by the licensee (Ref. 2) indicated l

most limiting A00's on the basis.of DNBR and CTM limits are (1) con-trol element assembly withdrawal and (ii) loss of coolant flow.

CESECLII version was used to simulate the primary system response, and CETOP/CE-l_was used in DNBR analyses.

The staff has reviewed _the initial conditions used in the analyses of. the above transients. With the initial power level assumed to be 103% of the rated power the. final transient results-show that DNBR> 1.24 and PLHR <21.0 kw/ft and the staff finds these results acceptabin.-

r 2.3.7 COMPARISON OF CYCLE 1 TO CYCLE 2 Comparison of thermal hydraulic design conditions for ANO-2 Cycle.1 and Cycle 2 is provided in Table 2.

It can be seen that the difference i.

in Cycle 1 and Cycle 2 design parameters is in calculational factors.

L This is due to application of SCU and new methodology CETOP/CE-1.

i L

I l

- 20 '

TABLE Arkansas ::ccisar Cnc '.'.-i: 2 inermal hydraulic' Farameters Quil Nec Reference Gercral Ch: Et@4tice Unit Cycle 1 Cycle 2 2315' 2915 Total Heat Output-(Coretonly)

MWt 9608 9608 106 Stu/hr 0.975 0.975 Fraction o Weat Generated in Fuel Rod Primary System Pressure Nominal.

psia 2250 2250

, Minimum in steady state psie.

2203 2203 Maximum in steady state psia 2297 2297 Inlet Temperature (MaximumIndicated)

• F 554~. 7 554.7 Total Reactor Coolant Flow (Minimum Steady State)gpm 322,000 322,000 106 lb/hr 120.4 120.4 Coolant Flow Through Core (Minimum)~

106 lb/hr_

116.2 116.2

Hydraulic' Diameter (Nominal channel) ft 0.039 0.039

. Average Mass Velocity

106 lb/hr-ft2 '2.60 2.60 Pressure Drop Across Core (Minimum steady psi 15.6 15.5 state flow. irreversible' ap over entire fuel assembly)

(

Total-Pressure Orop Across Vessel (Based on psi 35.9 35.9 nominal dimensions and minimum steady state _ flow)

STU/hr-ft2 184,720 184,720' l

Core Average Heat = Flux (Accounts for fraction of heat generated in fuel rod and axial densification factor). _.

I Total' Heat Transfer Area. ( Accounts for ft2 50,707 50,707*

axial densification factor) 2 Film Coefficient at Average Conditions STU/hr-ft *F 6200 6200 i.

Average Film' Temperature Difference

'F 30.6 30.d Average Linear Heat Rate of Undensified Fuel kw/ft 5.40 5.40*

l.

Rod-(Accounts for fraction of heat

. generated in~ fuel rod) j

-Average Core Enthalpy Rise 3U/lb 82.7 82.7 Maximum Clad Surface Temoerature

• F 656.5 656.5 i

e i

h

-n-If? J*nuec)

~~ 3I E 2

~

Reference cgla 1 Cycle 2

,..... 1..o. r... 21 rx.

..r 3

.l.03

1.025++

Engin'eerin.. sit Flux Factor 1.03

1. 008+, ++

Engineering F:ctor.cn Hot. Channel Heat Input 1.05 1.05.&+

Rod Pitch,- Bowing and' Clad Diameter Factor l~.002 1.002

. Fuel Censification Factor (Axial) 14.5 14.5 Peak Linear Heat Generation Rate (kw/ft) 1.30 1.24++-

Minimum DNBR NOTES:

.Sased on.il28 shims.

Based on "as-b'uilt" information.

+

++ These factors have be an ccmbined statistically with other uncertainty factors-at 95/95 confidence / probability level to define a new design limit on CE-1 minimum DNBR), hen iterating on power as discussed in Reference 6-7.

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2.4 ACCIDENTS AND' TRANSIENTS-2.4.1 ' ANTICIPATED.0PERATIONAL OCCURRENCES 2.4.1.1 CONTROL ELEMENT ASSEMBLY (CEA) WITHDRAWAL CEA withdrawal was reanalyzed for the conditions of Cycle 2 to demonstrate l~

that the initial margins were maintained by the applicable ~ values of the l

-Technical Specification Limiting Conditions for Operation (LCOs) and the l

reactor-coolant system design pressure limits. Trip signal calculations are performed within the Core Protection Calculator (CPC) where the algorithm uses core power, heat flux and reactor constants to provide a conservative estimate of the trip signals in such a way as to prevent exceeding j

MDNBR,. maximum local power density, or RCS. pressure. The CEA transient has been calculated for withdrawal frcm subcritical, one percent power and 1

l full l power conditions.~ Including feedback effects and control rod position i

at critical, the most critical parameter for the subcritical case is' reactivity. addition. rate. ; The input values selected maximized the-power increase and the margin degradation. No safety limits were exceeded.

The CEA withdrawal from one percent power was similarly analyzed. The.

l transient is terminated by a high pressurizer pressure trip.. The resulting-maximum RCS pressure.is 2662 psia and occurs before the high1LPD or the MDNBR trip would _be activated. No DNBR or LPD limits are exceeded.

The review of the CEA withdrawal. indicated that none of the LCOs will be exceeded..hence, the results are acceptable.

l t

I

t 23'-

r

' 2. 4 ~.1. 2 FULL AND PART LENGTH CEA DROP l:

Tne.most-important factor in such a transient is the possible rate h

of reactivity insertion. 1The CPC shall initiate a trip infa manner F

'that:thetinitial margins be maintained by the LCO to prevent violation' l-of the DNB, CTM or LPD des.gn limits. Two cases were~ analyzed i.e., full-I slength and part length CEA drop. The CPC' constants include CEA penalty factors which account for any CEA misalignment including a drop. The

(

' penalty factors assure-a conservative estimate of the transient-MDNBR and. maximum LPD.

l.

The single full length control rod drop can Lcause an increase of the l

peaking factors by 17 pcrcent over predrop values. However, the CEA penalty factors 11n the CPC will prevent power distributions which could violate MDNBR limits.

The.part length;CEA drop not only can cause severe axial and radial flux distortions but-it can' insert positive reactivity. However, the CPC-. initiated MDNBR.or maximum LPD will prevent the respective limits from being exceeded.-

The methods used in~the analysis are consistent with those used in the FSAR which we have.previously reviewed land approved.

The review o'f the CEA misoperation indicates that CPC originated trips l

will. prevent violation of MDNBR, CTM or LPD limits, hence it is acceptable.

l 2.4.1.3 FUEL MISLOADING The original submittal did not deal with the potential consequences of l.

fuel misloading on the assumption tha?. such consequences would be no-more severe than'those analysed.for the first cycle.. At our request the licensee submitted additional'information for the ANO-2 Cycle 2 misloading analysis. -The' analysis was based in part on the analysis aerformed for Calvert Cliffs Unit 2 Cycle 4.

The analysis demonstrates that differences-in power peaking and power distribution for fuel assemblies irradiated for one orrtwo cycles will-be detectable by symmetry checks. The misloading considered includes fuel assembly interchange and assembly misrotation.

When'the assemblies are such that their-reactivity differences are not detectable with the CEA symmetry checks, the-increase in power peiking will'be'small and will not reduce significantly the available margins.

i L-

4

.y In add!. tion the licensee' stated tut tu wom =hinMing event which rc:n be postulated for AHO-2 Cycle i er Calvert Cliffs 2 Cycle 4 cannot occur'in.AN0-2 Cycle 2, hence, the latter is bracketed by the existing analyses. 'Hence, the _ consequences of undetectable misloadings for ANO-2 Cjcle 2 will be less severe than the:c evaluated and approved for ANO-2

Cycle 1 and Ca.1 vert Cliffs Unit 2 Cvele 4._

We find those arguments reasonableiard the misloading case. acceptable.

~ 2.4.l.'4' CLOSURE OF ONE MSIV

-The transients resultin'g from the instantaneous closure of a single Main Steam Isolation Valve (MSIV) were analyzed for Cycle 2.to determine the Core Protection Calculator (CPC) Asymmetric Steam Generator Transient Protection

(;SGTP)' trip setpoint. This setpoint is determined in conjunction with the initial margins maintained by the LCOs so that the DNBR and fuel center-line-to-melt (CTM) ' design limits-are not exceeded.

CESEC'II version was used to simulate the primary system response, and CETOP/CE-1 was used in MDNBR analysis.. Although CESEC II version

-has not been approved, the' staff finds nit acceptable for this -

application.

The four evente which'af'fect'a. single steam generator are:

(a)' _ loss:

~

of load to1oneesteam-generator; (b) excess ~ load to one steam generator;_

i L

,(c)' loss'of feedwater to one steam generator; and (d) excess feedwater

-to one steam' generator.

The licensee has justified, by the detailed studies documented in reference 5, that the loss of load to_one steam. generator (LL/lSG) i-event produces by far-the largest margin degradation and thus is l-the.most limiting asymmetric event. =Since this event is most limiting l

itlwas the only-asymetric event analyzed for Cycle 2 to establish the i

ASGTP set points. JThe staff'has" reviewed the referenced studies and l

-finds this approach acceptable.

L I'

This event was analyzed for Cycle 1 operation prior to the installation

(

'of the ! Asymmetric Steam Generator. Transient Protection related CPC trip in AN0-2 This' Cycle-2 : analysis establishes the reference analysis

.for future cycles-in which the ASGTP trip is operational,

~

The event.is initiated by the inadvertent closure of a single main steam l

isolation valve _. causing a loss of load to one steam generator. Upon loss sof load,-pressure.(and temperature) in the affected steam generator will increase to the' opening pressure of the main steam safety valves. The intact l

steam generator picks up the load loss, which causes its temperature and b,

pressure 5to decrease. The cold leg temperature a ymmetry leads to a reactor inlet temperature tilt which produces an azimuthal core power tilt.

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/ Conservative assumptions were used in the anal / sis to account for the maximum power tilt and hot channel recul peaking factor increase (the

=ssumption;used in this caseLi

~ coefficient of -3.5 X 10' ap/g the mnst negative.mooerator temoerature F.sinca this maximizes the power tilt and the increase in the hot. channel radial peaking factor)'. With.this

,conserva+.ive assumption, the greatest asymmetry in core inlet temperature

-distribution, the greatest increase in hot' channel radial peaking factors and-the most' limiting DNBR of 1.24 will result.

'A reactor trip is generated.by the CPC low DNBR trip on high differential; temperature between the cold-legs' associated witn tne two steam _ generators. The

-ASGTP tr,ip;setpoint within the CPC. ensures that the acceptable DNBR limit will'not be: exceeded at'any time during the' event.

.A maximum allowable initial linear heat generation rate of 16.5 KW/ft.could exist as an initial condition' ~without exceeding the maximum linear heat generation rate-of 21.0 KW/ft.above which' fuel centerline melting is predicted to occur.during this transient.

This amount of margin is assured-by setting the Linear Heat Rate LC0 based on the more limiting'of the allowable 71inear heat rate for LOCA-14.5 KW/ft.and other transients.

Initiating the event from the extremes of the LC0 in conjunction with the CPC (ASGT

protective) trip will prevent DNBR or' centerline fuel temperatures from exceedi.ng the design. limits and.the maximum pressure within the RCS and main ' steam systems from exceeding:110% of the design pressures'.

'The analysis:resul'ts of this transient meet the acceptance criteria of f

~~SRP.section L15.2.3.3 and are acceptable.

2.4.1.5 BORON DILUTION An-inadvertent boron dilution event adds positive reactivity by reducing the baron crocentration in the primary coolant. This produces power and temperature. increases in the core and'may' cause an approach to both the'-

DNBR.and the fuel centealine-temperature-to-melt (CTM) limits.

'The boron dilution event was reanalyzed for Cycle 2 to demonstrate that (1) sufficient. time is available for the operator to identify the cause

~

of and to. terminate an ap? roach to criticality for all subcritical modes of operation and (2) to-demonstrate.that sufficient scram reactivity is available in all operating modes.

In a baron dilution event during power operation (Modes 1 & 2) the core protection calculator system's DNBR trip,.or, for more rapid-power excursions,:the high logarithmic power level trip will occur prior to reaching the DNBR o.r CTM limits. The high pressurizer pressure trip will 1

prevent the RCS pressure from reaching the RCS pressure limit. These

~ trips'will provide a positive means of alerting the operator to a boron 4

-dilution event in-progress and will provide adequate time to terminate the boron dilution event. We. find these results acceptable.

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-For.a coron'aiiution event ouring the suocr1tical mooes.(Mooes 3,: 4, 5 ano

6) tne colo shutdown mode (Mode 5) with the Vessel Water level drained'

'down to. the lip of the outlet nozzle and the refueling mooe provide the most limiting times fro'm the initiatici of the event until the five percent shutdown margin,.is exhausted and the reactor. returns to critical.

vunsidering this assumption for Mode '5 and 6, times of'35 and 40 mi,utes respectively are. calculated to elapse between initiation of the t' lution anc loss of the five percent shutdown margin. ~However, the time of importance, k meet the acceptance criteria of the Standard Review Plan, is the time betwee.n the' provision of a positive indication-to the operator and a return to criticality. This time should be at least' 15 minutes for ibde 5 ar.d 30 minutes for Mode 6.

Therefore it is the staff's position that a positive means for. alerting the operator to a baron dilution event in progress be installed as soon as' practical.

In order to be able to take credit in-the analysis for this alarm it must meet the. single failure criteria

.per section II.2.C of SRP section 15.4.6.

2.4.1.6 LOSS OF LOAD / LOSS OF CONDENSER VACUUM / TURBINE TRIP The loss of load (LOL). loss.of condenser vacuum.(LOCV), and turbine trip events are analyzed to demonstrate that the RCS and main ~ steam system pressures do

. not exceed'110% of design values (i.e.. 2750 psja and 1210 csia, resoectively)

- for Cycle 2 operation. These three events were presented in the FSAR

.as' separate events.-.'For Cycle 2 an analysis-was performed of a single event which bounds all three FSAR events.

.The bounding event considered is-a Loss of Load event initiated by a turbine l

trip without a simultaneous reactor trip,'and assuming the Steam Dump and Bypass system is inoperable.

If the turbine trip was caused by a Loss'of Condenser Vacuum, the main feedwater pump. steam turbines would trip _at the same. time. Therefdre, to cover these events a LOL concurrent with loss of feedwater was analyzed. The loss of load causes steam generator pressure to' increase to the opening pressure of the main steam safety. valves.. The reduced secondary heat sink leads-to a heatup of the RCS ar.d, in the presence of the assumed positive MTC, an increase in core power. The transient ~is terminated by a reactor trip on high pressurizer pressure.

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. Conservative assumptions were usad in the transient analysis to account for J

(a) the steam dump and bypass valves which were assumed to remain closed,,

(b) positive MTC of 0.5 x 105 -e/*F, and a least negative Doppler coeffi-cient with a: fuel tencerature coefficient multiplier of 0.85, and (c) a bottom peaked axial shape which minimizes'the negative reactivity insertion during the initial portion of the scram following a reactor trip and maximizes the time required to mitigate the pressure and heat flux increase.

j The Loss'of Load' transient analysis resulted in a peak reactor coolant pressure of.2671 psia. The increase in seconaary. pressure.is limited by the opening of the main steam safety valves. The secondary pressure peak value of 1144 psia was reached at 13.3 seconds after initiation of the event.

The results 'of the analysis demonstrate that the Loss of Load type event will not result-in peak RCS pressure or. peak ' main steam pressure in excess of their: respective upset pressure limits and that the mininum DNBR did not

-decrease below 1.24..

The analysis results for this transient meet the acceptance criteria of SRP

.Section 15.2.1 and are acceptable.

l 2.4.1.7 10SS OF-COOLANT FLOW The loss of Coolant Flow event was reanalyzed for Cycle 2 to determine the j.

L minimum' initial DNBR margin that must be maintained by the Limiting Conditions i

for 0perations -(LCOs) and the margin degradation rate which must be projected l

by the Core Protection Calculators (CPCs) such that a low DNBR trip will

. be initiated before the DNBR limit is exceeded.

.The methods used to analyze this event are consistent with those discussed l

'in the FSAR-with the exception that the design thermal margin model CETOP

~

l-was used for all DNBR calculations. The acceptability of the changes in the i

analytical models are discussed in Section 2.3 of this report.

> The 4-Pump Loss of Coolant Flow (LOCF) produces an approach to the DNBR l.

limit due to the decrease in the core coolant flow. Aside from the basic-1.

assumption of 4-pump LOCF without a simultaneous reactor trip, other conser-(

vative assumptions were used in the LOCF transient analysis to reflect the j

..following initial conditions:

(a) the Technical Specification LCOs, and

(b) an axial shape with a negative shape index of

.18.

The analysis for this transient'showed that-the minimum DNBR did not decrease-below the 1.24 limit.

The CPC low DNBR trip assures that loss of l

coolant flow events initiated from within the Technical Specification LC0's

. will not. result in a violation of the DNBR design limit. The maximum pressure within the reactor coolant and main steam systems did not exceed 110% of the l

. design pressures.

The analysis results, for this. transient meet the acceptance criteria of SRP Section 15.3.1.and are acceptable.

Li'

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2.4.2 ' POSTULATED-ACCIDENTS 2.4.2.1 -CONTROL ELEMENT ASSEP. SLY-EJECTION A zero and fu11' power CEA' ejection accioents-nave been analyzed..The analy-Ltical method, detailed in topical-report CENPD-190-A, has been approved.

t The calculational procedure computes the radial = average' and centerline fuel enthalpies.to determine the fraction of the pins that exceed the criteria for clad damage. To assure that the calculated values bound-the most adverse

- conditions the following assumptions were made:

-(a) the beginning of cycle 1(BOC) Doppier coefficient was assumed which yields the least negative reactivity ' feedback;-

(b) the_ BOC moderator temperature coefficient of.5 x 10

• ap/'F was assumed which results in a positive reactivity-feedback; and

_ (c) an end:of cycle (EOC) delayed neutron fraction was used which results in the highest power rise during the transient.

'Very low (zero) 'and full power conditions were analyzed with both terminating from a high linear power level trip.

The results.of the analysis indicate that a small fraction (<.005) of the fuel reaches the incipient centerline melt threshold. The total energy deposited dur#"1 the transient is less than 200 cal /gm criterion for clad damage. The resa... for Cycle 2 compare well with the corresponding results for Cycle 1.

The methodology and the results' of the,CEA ejection accident are acceptable.

2.4.2.2 SEIZED RCP SHAFT The ' seized shaft event was reanalyzed for Cycle 2 to demonstrate that the RCS~ pressure-limit of 2750 psia will not be exceeded and only a small

-fraction of fuel pins are predicted to fail during this event which will i-not cause the 10 CFR 100 site boundary dose guidelines to be exceeded.

1 Thel single reactor coolant pump shaft seizure is postulated to occur as a j

consequence of a mechanical failure.

In this hypothetical event, the RCS fimv f rapidly decreases to 'the three-pump value. A reactor trip is initiated by a low coolant flow rate which'results in.a rapid reduction in the margin to DNB, so that a CPC low DNBR trip occurs to terminate the transient.

The analysis for this_ transient used an axial shape index value of

.18.

This case is selected to be consistent with the Loss of Flow case.

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The analysis, however, did not take into consideration the single failure cri terion.

It is our position that the licensee should provide a confirmatory analysis diat shows that the consequences of this accident, including the worst single failure, will still meet the primary system pressure limit anc tne 10 CFR 100 dose limit for this transient.

2.4.2.3 STEAM LINE BREAK The steam line break (SLB) transient is an. overcooling event. The full power SLB events were. reevaluated for Cycle 2 to account for the reduced shutdown margin from -8.6 to -7.9% 40, increased Doppler feedback, and decreased reactivity insertion during moderator cooldown. The steam generator blowdown and associated reactor coolant system (RCS) cooldown were not recalculated for Cycle 2 since the net effect of changes in the above parameters on the blowdown will be small. The Cycle 1 cases _are based on cooldown curve associated with an initial allowable MTC of -3.5 x 10

• ao/*F, while the Cycle 2 cases are based on the cooldown curve associated with an initial allowable MTC of

-2.8 x 10 6 ap/*F. Comparison of the Cycle 1 and Cycle 2 results from these curves shows that the positive reactivity insertion due to cooldown of the moderator is less for Cycle 2 by 1.1% ao at the time of minimum negative reactivity. 'This improvement in the moderator cooldown behavior is sufficient to completely offset the 0.7% ap decrease in available shutdot9 worth and

.2% ao increase in positive reactivity insertion due to Doppler feedback.

The results of the analysis of the spectrum of steam line break accidents demonstrated that the peak reactivity experienced during the transient for Cycle 2 is bounded by the FSAR results. On this basis, the licensee concluded that the FSAR results are conservative and that the conclusions presented in the FSAR remain valid for Cycle 2.

The licensee's analyses showed that based on DNBR criteria, no fuel damage will result. Without fuel damage a detailed dosage reassessment is not required.

l Based on the above, the staff concludes that the analysis results for this transient meet the acceptance criteria of SRP Section 15.1.5 and are acceptable.

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=a w 2.4.2.4 FEEDWATER LINE BREAK The acceptance criteria for _this event as stated in SRP Section 15.2.8 are that the RCS pressure sheld r.ot exceed 110", of design pressure and any fuel' damage predicted to occur should be sufficiently limited such that core coolability is maintained. The feedwater line break accident was reanalyzed for Cycle 2 to determine that the RCS pressure upset limit 2750_ psia is not i

exceeded.during.the transient, and that any fuel damage predicted to occur is limitte ;.

The feedwater line break analyzed was assumed to occur during full power operation and with concurrent loss of non-emergency A-C power at-time of.

trip. This results in the maximum initial stored energy and minimum steam l

generator inventory.

In addition, in response to loss of non-emergency AC power upon trip, the following will' occur to maximize RCS pressure increase:

to coastdown; (3) pressurizer control systems are lost; and (4) pumps begi (1) turbine trip valves close immediately; (2) reactor coolant 112.4 seconds L

rather than 97.4 seconds are required for the automatic initiation of emergency feedwater to the _ unaffected steam generator. This combination of parameters maximizes the calculated RCS peak pressure.- The limiting break size was determined by a parametric study performed with the methodology previously reported in the FSAR.

The results of the Cycle 2 reanalysis predicted that the RCS pressure would L

L

. increase to 2705 psia. Following reactor trip on either hign pressurize l

pressure or low steam generato' water level, the decay in core power and the action of the primary and secondary safety valves result in a reduction of I

RCS pressure. Subsequently, the effects of system flow coastdown due to loss of AC upon trip, continued blowdown of steam from the intact steam generator through the break and the entering of emergency feedwater to the intact steam generator cause the RCS first to go through a mild pressure increase and then a steady decrease. The decrease is reversed when the low steam generator pressure initiates the closure of Main Steam Isolation Yalves l

(MSIV).-=The MSIV closare terminates the blowdown of steam through the break, thus causing the RCS to hu t up once mora. Eventually, the heatup is. terminated by the opening of secondary safety valves.

The results of this analysis demonstrate that the Feedwater Line Break-l Event will not result in a peak RCS pressure which exceeds _ the upset l

pressure limit of 110% of the design pressures. The licensee's analyses showed that based on DNBR criteria no fuel damage will result. Without fuel damage a detailed dosage reassessment is not required.

. Based on the above, the staff concludes that the analysis results for this ttransient meet the acceptance criteria of SRP Section 15.2.8 and are I

acceptable, r

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. 2.4.3 LO5S-GF-CGGLAiii ACCIDE*iT Much of the analysis for Cycle 1 ooeration was used as the basis fer the Cycle 2 evaluatten. Only th: fuel pin thermal analysis using STR!rm-II and the PARCH codes was per crmea for the Cycle 2 worst break.

It. was not necessary to repeat the biv down and reflood hydraulic analyses since the analyses performed for Cycle 1 are applicable to Cycle 2.

The table below compares the results of the Cycle 7 analysis with the Cycle 1 analysis. As shown, the perfor.T.ance requirements of 10 CFR 50.46 are not exceeded. We, therefore find the LOCA analyses acceptable.

ANO-2 LIMITING BREAK (1.0 DEG/PD) RESULTS Peak Cladding Peak Local Core Wide Case Temperature (*F) 0xidation (%)

0xidation (%)

Cycle 1 2078 11.82

< 0.617 Cycle 2 2041 11.80

< 0.621 10 CFR 50.46 2200 17.0 1.0 2.4.4 RADIOLOGICAL CONSEQUENCES OF POSTULATED ACCIDENT The licensee provided revised evaluations where previous Cycle 1 values did not bound Cycle 2 values. As shown below, the Cycle 2 radiological consequences are acceptably small fractions of the 10 CFR 100 limits for accidents.

2.4.4.1 SEIZED RCP SHAFT ~

Because of numerous changes in parameters and methodology tne number of rods

)

calculated to have DNBR values below the limit is lower in Cycle 2 than in Cycle 1.

Therefore, the radiological consequences for the seized reactor coolant pump shaft event are no greater that previously approved values.

30 -

2.4.4.2 CONTROL ELEMENT EMPT!M The licensee reevaluated tne control element assembly ejection accMc.;t for Cycle 2 and concluded, usina methods oreviously approved in CENFD-130, that the frac';1on of rods predicted to suffer clad damage for the limiting case is less than the fraction predicted for Cycle 1.

Therefore, the radiological consequences for CEA ejection accident are no greater than previously approved values.

i 2.5 REACTOR PROTECTION SYSTEM 2.5.1 VERIFICATION AND VALIDATION OF CPC SOFTWARE MODIFICATIONS The licensee proposed a number of modifications to the Reactor Protection System's Core Protection Calculawr System sof tware. The principal purpose of the changes was to implement new thermal hydraulic and physics algorithms.

The. acceptability of the new algorithms is discussed in Section 2.3 of this report.' We have reviewed the verification and validation procedures for changes to CPCS software which were followed to assure that the new algoritnms l

have been incorporated into the CPCS software as intended. The verifica-t tion and validation procedures were originally reviewed and approved by the NRC staff during the operating licensing review for ANO-2. These procedures include provisions for confirmatory testing of a modified proto-type but single channel CPCS. Based on an audit of the licensee's procedures and tett program, we conclude that the procedures and tests previously accepted by the staff have been followed during the current rouno of CPCS modi fications.

On the basis that acceptable procedures and test programs have been followed l '

in modifying the CPCS sof tware, we consider the implementation of the new sof tware to be acceptable.

2.5.2 TECHNICAL SPECIFICATIONS TO CONTROL MODIFICATIONS TO CPC ADDRESSABLE i

CONSTANTS The licensee has proposed an increase in the number of CPC " addressable l

constants". Addressable cons *. ants are variables which may be modified L

between cycles or even during reactor operation. Because the CPC is a part of the ANO-2 protection systera, we believe that appropriate measures should i

be taken to assure that such nodifications are done correctly and that the new values of the constants inserted do not decrease safety margins. Consequently, we asked that the ANO-2 Technical Specifications be amended to include provisions to control modifications to addressable constants. This request was also

. prompted by the proposal by the licensee to pennanently connect the data links which allow transfer of information from the CPC and CEAC systems to the plant computer, an issue adressed more fully in Section 2.5.3 of this I

report.

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Specifically, the lic-eties has propmM in a letter dated May 19. 1991 (Ref. 5) the f.Micwing controls:

(1) All CPC addressable constants are to be identified in the Technical c

Specifications.

(2) The bases to the Technical Specifications (Section 2.2.2) shall reference a document which explains the methodology and procedures for obtaining modified values of addressable constants.

(3) Those addressable constants expected to be modified frequently and from the operator's console will be restricted to specified ranges unless ~ approved by the Plant Safety Committee.

(4) Tnose addressable constants expected to be modified less frequently by loading from a disc storage unit shall be approved by the Plant Safety Committee unless the modification is based on a technical specification or core physics test requirement.

(5) An independent verification shall be conducted to confirm that the desired value of each constant to be modified has actually been entered.

(6) Modifications to the CPC addressable constants based on information obtained through the plant computer (CPC data links) shall not be made without prior approval of the Plant Safety Committee.

Although a complete and approved document to meet provision (2) will not be available for Cycle 2 startup, by letter dated May 26,1981 (Ref. 6), an interim document has been provided with a commitment to provide a final document by August 17, 1981.

We consider the propoua Technical Specification controls on addressable constant modifications to be acceptable, including the delays in providing a final document specifying the methodology for modifying constants.

2.5.3 DATA LINKS BETWEEN THE CPC/CEAC AND THE PLANT COMPUTER The licensee has proposed to permanently connect the plant computer, a non-safety grade computer, to the core protection calculators (CPC), and control element assembly calculators (CEAC), part of the safety grade protection system. A similar proposal was made during.the operating license (OL) review but was rejected by the staff because of concerns that the connection added unnecessary complexity to the CPC/CEAC design, and that there'might be an adverse indirect effect on the protection system if data from the plant computer were used in calibration of the CPC addressable constants. The issue was discussed in our safety evaluation report for ANO-2 OL, NUREG-0308, and in particular in relation to Position 20 of Table 7.1 of that document and its supplements.

4

-- 3a.

The concern that data from tne.nlant computer mignt De useo to modify CPC addressable constants and thereby adversely-affect the CPCs nas Deen adaresseo by establishing contrel: On the modifications of CPC addressable constants:in a

- the Tecnnical Specificcticns (Section 2.2.2).

As discussed elsewnere in this

. report changes to adorestanie' constants based on data from tne olant computer may be made only upon approval of the Plant Safety Committee. We consider

. this to be an acceptable resolution of;this-concern.

. The staff concern about the unnecessary complexity associated with the. data link design at the time of OL review was a general concern rather than one based 'cn1 a potential deficiency in the measures taken to physically. isolate the plant computer from tne.CPCs and CEACs.. The use of qualified optical isolation devices at both ends of the digital data links and use of qdalified.

current-to-current isolation devices for the analog data links to.the plant computer preclude the possibility of a fault in the plant computer being propagated to the CPCs or CEACs. Furthermore, 'the watch-dog timers are used to prevent delay in a needed CPC trip should an inordinate time be

- spent in processing data througn the data links to the plant computer.

Although the existence of the data links' adds some complexity to the CPC/CEAC design as stated in the OL SER (NUREG-0308), we have reconsidered the issue and bel.ieve that the possibility of an adverse impact on safety is remote.

Also, the new controls'on CPC addressable constant modifications will

- prevent an Jn8CCeptable impact on the CPCs from recalibration using plant computer data. We, therefore, conclude that-the permanent connection

- of the data links between the CPC/CEAC'and the plant computer is-acceptable.

- 2.5.4 - MONITORING OF CPC ROOM TEMPERATURES

' During Cycle 1 operation the licensee reported instances where sequences of CPC ~ auto restarts were attributed to high CPC room temperatures. To assure that high room temperatures.do not affect CPC reliabiltiy, we requested that

- the licensee address this issue in the Technical Specifications..The licensee has' proposed Technical Specification 4.3.1.1.6 to require a CPC channel functional test if a CPC room high temperature alarm is received. We consider this acceptable.

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. 2.6 LICENSE CONDITI6W3 The licensee has satisf actorily completed the requirements 'of na following four license conditions. Accordingly, these conditions are 5:r:by deleted from the license. The fifth license condition in the list below nas also been modified as stated.

2.6.l ~ FUEL PERFORMANCE License Condition 2.C.3.a on fuel performance required that, prior to startup for that fuel cycle in which burnups greater than 20,000 mwd /Mt, the Commission be provided with fission gas release calculations and analyses using calcula-tional methodology approved for burnups greater than 20,000 mwd /Mt. This matter has been acceptably resolved by the licensee as is oiscussed in more detail.in Section 2.1.1 of this report.

2.6.2 INSTRUMENT TRIP SETPOINT DRIFT ALLOWANCE License Condition 2.C.3.d required the licensee to submit values of (1) instrument drift (2) cumulative instrument bias and (3) the margin between the trip setpoint and the assumed accident analysis trip value for inclusion

.in the Technical Specifications.--

The staff's initial request on this subject was transmitted to the licensee by letter dated March 22, 1977. The licensee responded to that letter and to the license condition by letters dated February 28, 1979 and November 27, 1979. The staff assigned review of the licensee's submittals to the Lawrence Livermore National Laboratory (LLL) under the staff's technical assistance

. program.

The licensee's submittals included specific values of the reactor protection system and engineered safety feature actuation system trip setpoints for inclusion in Technical Specification Tables 2.2-1 and 2.2-2, respective;v.

- The staff's consultant (LLL) concluded that the proposed changes to the s6t=

point values are acceptable. Further modification of some of these values has 'been proposed by the licensee to accommodate the Cycle 2 Reload Report analyses requirements. We have evaluated these differences in setpoint values and on the bases that the differences are relatively small and the effects of the differences are reflected in the Cycle 2 Reload Report analyses results, we conclude that the proposed Cycle 2 values are acceptable. A related change to the pressurizer high pressure trip setpoint is discussed in Section 2.7.1 of this report.

i

30 The:11censee's ::6 mitt:Is also included a recort describing the setpoint methodology that Cc=bustion Engineering, Inc. used to detemine the ANO-2 setpoints. The LLL review of the methodology concluded that the sthod used for determining the total equipment error is a reasonable method for detemining the RPS and ESFAS trip setpoints and allowable values.

Further details of the LLL review are contained in a copy of their Technical Evaluation Report which is attached to this report. Based on our review l

.of the LLL report and the Reload Report as described above, we have concluded that the RPS and ESF Technical Specification setpoint values as identified in the Reload Report are correct and acceptable.

2.6.3 RCS OVERPRESSURE MITIGATING SYSTEM License condition 2.C.3.f required that the licensee achieve full implemen-tatirn of the proposed reactor coolant system overpressure mitigation systems described in the licensee's letter dated October 11, 1977 prior to Cycle 2 acartup. The staff has previously reviewed the proposed design as reported in Supplement No.1 to the SER dated June 1978. The design was approved subject to the stipulation that the design be demonstrated to meet the following specific criteria:

(1) Provide an interlock or alarm on the isolation valves which meets the applicable IEEE Standard 279 criteria and seismic Category I criteria l

for valves numbered 2CV-4730-1, 2CV-4731-2, 2CV-4720-2 and 2CV-4741-1, such that if the reactor coolant system temperature drops below the proposed temperature, and all the isolation valves are not fully open, an alarm sounds in the control room or the isolation valves open automatically.

(2) The electrical portion of the pemanent fix conforms to safety-grade criteria.

We requested additional information and the licensee responded with references 2 and 3.

Based on our review we conclude that the design meets the above criteria and is acceptable.

l I

2.6.4 CEA GUIDE TUBE SURVEILLANCE PROGRAM License Condition 2.C.3.1 required that prior to Cycle 2 startup the licensee submit the results of a surveillance program conducted on the design modifica-tions to the CEA guide tubes. The program was to be directed toward determining g

whether unac:eptable degradation of the guide tuoe components had occurred.

l l

Our review of this matter is presented in Section 2.1.5 of this report wherein we conclude that the issue of guide tube wear is resolved for ANO-2. Therefore,

.on the basis of otir findings and conclusions as presented in Section 2.1.5 l

of this report if cense condition 2.C.3.1 is satisfied and is hereby deles 1 l

from the license.

~

, 2.6.5 MAXIMtM PO'.!CR LEVE' l

License Condition 2.C.1, Maximum _ Power Level, presently centains the requirement for the licensee to complete certain preoperational tests, startup tests and other items identified in Attachment 2 to the license. contained items for which completion was reouired prior to attaining full power following the initial licensing of tne plant. ~ These items were the remainder of the total list of-tests at the time the license was issued. Accordingly, once the plant's preoperational and first startup testing program has been acceptably completed no need exists for Attachment 2.

Therefore, license Condition 2.C.1 is modified to delete reference to Attachment 2 and Attachment 2 is deleted from the license.

License Condition 2.C.1 is also modified to limit plant operation to seventy p.. cent of..the licensed power level of 2815 MWt pending completion of the final details of the staff's review of the Core Protection Calcu-lator system sof tware changes for Cycle 2 operation as. discussed in Section 2.3 of this report.

e 4

J

L -,

c.

a

- 3e -

2.7 OTHER NATTERS

-2;7.1 CONTAINuEHT PRESSURE, TEMPERATURE, AND HUMiniTV i

.On November: 19,:1979,'the licensee submitted a propocad change to Technical

~ Specification 3.6.1.4, on containment atmosphere conditiens. The proposed ~

change would reduce the allowable containment temperature over a range of'

~

pressures. The changetis proposed to make the TS-consistent with assumptions contained in the FSAR on initial containment pressure, temoerature and relative humidity such that the maximum-differential oressure across the

~

containment would be 5.0 psi in the event of an inadvertent containment spray actuation. This will assure that the containment-pressure, as a

-result of inadvertent spray operation, will not be lower than the contain-ment's external design pressure of -5.0 psig. The current specification does not provide this assurance.

I

.TheltaffhasreviewedtheproposedchangetotheTechnicalSpecifications and-on-the basis of-hudit calculations finds that the proposed change will l assure that the external design pressure will not'be exceeded for inadvertent

-spray operation. We conclude that the proposed change is' acceptable and should be' implemented.

2.7.2 -PRESSURIZER HIGH PRESSURE TRIP SETPOINT h

Ths licensee,'in its submittal of November 27, 1979, requested a change to Unit 2. Technical Specifications (TS). The high pressurizer pressure trip L

setpoint, Item 4 of Table 2.2-1 of the Technical Specifications, is presently b

1 2345 psia..It is proposed _to increase the trip setpoint by 17 psi to l-

< 2362 psia.

l The increase in the high pressurizer pressure trip setpoint of 17 psi is

-to eliminate a dynamic. allowance imposed. prior to. operation. The. test data collected at startup of Cycle 1 has demonstrated an instrument channel

~

response time less than assumed in'the safety analysis. Therefore, the

' dynamic allowance factor is no longer required.

The existing narrow range pressurizer pressure ~ instrument used for this j

trip has a. range of 1500 to 2500 psi. The present trip setpoint is < 2345 j

psia with an allowable drift of 8.887 psi (allowable value.of < 23537887 psia). The proposed trip settsint is-< 2362 with an-allowable drift of 8.887 psi (allowable value of 2370.887'~ psia)..The new trip setpoint is well within the range of the narrow range pressurizer pressure instrument and the allowable. drift (8.887 psi) for the-proposed setpoint is identical to that f0F'the'present setpoint. Therefore, it-is concluded that the proposed r

. trip _setpoint is acceptable.

Based on our review of the licensee's submittal, we conclude that the proposed change to the technical specification - high pressurizer pressure trip. set-

[

point _of < 2362 psia and allowable value of < 2370.887 psia is acceptable.

I

. -....-..,_-.-_,._ _-._.~...-..,_.-_._......-......_._ ~ _ ___ _.-

. ~

3.0 ' TECHiiiCAi. 5 ECIFICATIGN CHANGE 3 l.

i 3.1 THER;*al vaRGIN LI:4ITS The Technical Specifications a're modified to reflect the changeover from the W-3 DNBR correlation to the CE-1 DNBR correlation in conjunction with the statistical combination of uncertainties (SCU) methodology. The basic j

DNBR limit is changed from 1.30 h!-3) tu 1.24 (CE-1/SCU).

The pages affected for the DNBR limit change from 1.30 to 1.24 are:

2-1, 2-6, B 2-1, B 2-2, B 2-6, B 3/4 2-3, B 3/4 4-1.

~

' A change related to the change in DNBR correlations discussed above is the inclusion of limiting values on the addressable constant BEkR 1, the power uncertainty factor used in the DNBR calculation, in the TS. The page affected is 2-6.

The new DNBR limit and BERR 1 values have been found acceptable in l

Section 2.3 of this report for the conditions of.operatior authorized by the related license amendment. The staff's evaluation of the.DNBR limit and BERR 1-value for full. power operation will be addressed further in a forthcoming Safety Evaluation accompanying an amendment authorizing full power operation.

3.2 PEAKING FACTOR DEFINITIONS A definition of the planar radial peaking factor, F has been added to standardize.the ANO-2 definition and symbol with otEer, CE plants. The

- value of 1.053 for F is documented in a report which has been reviewed and approved by the fdC staff. The acceptability of these changes is dis-l cussed in Section 2.2.2 of the SE. Further details. may be found in the.

L

. licensee's letter dated May 11, 1981 response to question four. Since the I

5.3 percent uncertainty value has been reviewed and found acceptable these changes should be made. The pages affected are:

1-6, 3/4 2-4, B 3/4 2-1, l

B 3/4:2-2, and B 3/4 2-3.

l 3.3 DNB RELATED PARAMETER LIMITS The previous TS 3.2.6 on core average temperature is consistent with Standard Technical Specification requirements for non-CPCS CE plants.

However, for the ANO-2 DNB related safety analyses, core' average temperature is not an input parameter. The relevant parameters for the ANO-2 analyses are reactor coolant cold leg temperature, axial shape index and pressurizer pressure. Accordingly, the licensee has proposed TS limits on these values consistent with the values assumed in the ANO-2 Cycle 2 safety analyses.

Since, as reported in Section 2.0 of this report these safety analyses have been reviewed and approved for Cycle.2, the proposed TS changes should be made. The affected pages are:

3/4 2-12, 3/4 2-13, 3/4 10-2, 3/4 2-14 and B 3/4 2-4.

e

?

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- L-3.4 FAF.TIR :"2: nornaTION Partial--pump operation in MODES 1 and 2 has not boca allowed.in Cycle 1 and is not allowed in Cycle 2 since the licensee has not submitted for the

' Commi ssi cr.': rcvie.: and approval the ' safety analyses supporting such operations. % wever the licensee proposes certain changes to the TS.

l-to clarify this situation. These changes include the change of the term

(

ECCS" to " Safety" in-various footnotes to reflect that analyses must be submitted not only for ECCS-perfomance but for transients' and other L

accidents as well. :In addition clarifying language is added to other TS.

l 1These changes are e'ditorial in nature, do not affect safety, and are acceptable. The affected pages are:

2-5, 3/4 4-1,.3/4 ' 4-2. and 3/4 7-3.

In addition TS -3.4.1 (page 3/4 4-2) has been subdivided to provide a separate

' ACTION for MODE 3 to ensure that reactor coolant pump operation in MODE 3 is consistent witn the assumptions made in the main steamline break analysis.

l' This. change provides consistency.between the TS and the MSLB safety analysis and is, therefore, acceptable.

i 3.5 AVAILABILITY OF BORATED WATER FROM RWT MODES 1, 2, 3 AND 4 Because of the higher core average enrichment and an increase in the available l.

shutdownmargin requirements the licensee proposes to increase the refueling water tank-(RWT) required volume from 40,200 gallons at 1731 ppm to 56,455 l

gallons of 1731 ppm borated water. The proposed value was considered in the-Cycle 2 safety. analyses. Since as stated in Sc-tion 2.0 of this report, these safety analyses:have been reviewed and found acceptable the proposed TS change should be made. The page affected is B 3/4 1-2.

? MODES 5 AND 6 Because of the increased shutdown margin requirements the licensee proposes to increase the RWT required volume from 4,700 gallons at,1731 pr to 8185 gallons of 1731 ppm borated water. The proposed value was consioe d in 1

L the Cycle 2 safety analyses. Since, as stated in Section 2.0 of this report,

~

these safety analyses have been reviewed and found acceptable the proposed

'TS changes should be made.

f LA typographical error is also corrected to make the BASES consistent with 1

TS 3.1.2.7.

The page affected -is 4 3/4 1-3.

3.6 SHUTDOWN MARGIN FOR MODE 5 l

The shutdown margin was evaluated for a boron dilution event during the cold shutdown condition. It was determined by the licensee that a Si, AK/K shutdown margin would be required so that at least 15 minutes would be available to the j

-operator in order to terminate-the deboration transient. We find this l-l l

c I

4 proposed TS cnange acceptable..The pages affected are: 3/4 1-3, 3/4 1-8,.

3/4 1-10, 3/4 1-12,,3/4.1-15, B 3/4 1-1, B 3/4 1-2 and B 3/4-1-3.

3.7 'RPS AND ESFAS TRIP SETPOINTS

. a). The licensee proposes-to change the value of certain RPS and ESFAS

(

trip setpoints. The specific parameters are Linear Power Level-High, Pressurizer Pressure-Low, Steam Generator Pressure Low, Steam Generator Level-Low, Steam Generator Level-High, and Steam Generator P-High.

The acceptability of these changes is addressed in Section 2.6.2 of, this. report. The pages affected are:

2-5, 2-6, 3/4 3-16, 3/4 3-17

~

3/4 3-18.

~

(b). The' licensee also proposes to correct a typographical error in the refueling water tank' level minimum allowable value from 5.300% indicated level to 5.111% of indicated level. The acceptability of this change is addressed in the LLL report referenced in Section 2.6.2 of this report. The page affected.is 3/4 317.

(c). The licensee proposes to change the Pressurizer Pressure-High setpoint from 2345 psi to 2362Lpsi. The acceptability of this change is addressed in Section 2.72 of this report. The page affected is 2-5.

l l

The changes discussed above in.a. b and c are considerad in the Cycle 2 safety 1

- analyses. Since we have reviewed and approved these safety analyses the i

proposed TS changes should be made.

j l

3.8 'CPCS ADDRESSABLE CONSTANTS l

The licensee has proposed TS to control modifications to addressable constants.

The acceptability of these changes-is-addressed in Section 2.5.2 of this report.

l The pages affected are:

2-4, 2-7, 2-8, 2-9, B 2-7, 3/4 3-8, 3/4 3-9, and 6-13.

i i

3.9 CPCS ROOM HIGH TEMPERATURE L

. The licensee has proposed TS to-verify the OPERABILITY of the CPCS in event L

.of a valid high CPCS room temperature alarm. The acceptability of this TS l

is. discussed in Section 2.5.4 of this report. The page affected is 3/4 3-la.

l l-1:

I' p

i L

w

- I.iG R?3 AND'ESFh3 TRIF LIMIT TABLE FGCTNCT;3 5

-In the 'foote)tes to.these tabl'es.the licensee proposes to change the

'----."*- - *ssuri:er pressure is. reduced" and "as steam generator pressere ts. reduced" Lto "during a planned reduction in pressurizer pressure" and "doring a planned reduction in steam. generator pressure" respectively.

(This change properly. limits the manual reduction by the operator of the setpoint.to-occasions of controlled and planned reductions in pressure and is acceptable.1The'affected pages are: 2-6, 3/4 3-18.

3.11 MODERATION TEMPERATURE COEFFICIENT (MTC)

The acceptaoility of the MTC in TS 3.1.1.4 is supported by the discussion in Section 2.2/1 of this report. The affected page.is 3/4 1-5.

~

3.12 STEAM GENERATOR LEVEL-LOW TRIP The licensee proposes to change the BASES wording to reflect the fact that the AN0-2 emergency feedwater system. is actuated and supplies water to the steam generators automatically upon receipt of an ESFAS versus being required

- to be manually / actuated within a ten minute period. Since actuation and

_ feed.is automatic upon demand there is no basis.for the ten minute period in the ANO-2 safety analyses. The change is acceptable. The affected page is B 2-5.

3.13 CEA INSERTION LIMITS TS Figure-3.1-2 has been changed to reflect the provision of adequate-shutdown margin for MODES 1 and 2.

The acceptability of the shutdown

. margin is discussed in Section 2.2.1 of this report..The affected page Jis 3/4 1-27.

3.14 LINEAR HEAT RATE MARGIN L.

The licensee has proposed an additional TS Figure to provide-further definition of-the acceptable' operating limits.for the conditions of COLSS in-service and COLSS out-of-service. The previous TS did not include a figure defining the limits for these two conditions. : Based on the licensee's response to items 25 and 26'in their May 6, 1981 letter the

' change provides clarification and does not change safety margins.. There-fore the change is acceptable. The pages affected are: 3/4 2-1, 3/4 2-2,

(

_and 3/4 2-3.

'3.15 DNBR OPERATING LIMIT

.The licensee proposes to change Figure 3.2-4 due to Cycle 2 reanalysis of COLSS out-of-service ONBR margin requirements. These limits are reflected

'in the determination of the initial. conditions for Cycle 2 anticipated operational occurrences which we have evaluated and found acceptable E

in Section 2.0 of this report. Therefore the proposed change is acceptable.

The affected pages are: 3/4 2-7, 3/4 2-9, 3/4 2-10 and B 3/4 2-1

.... -. -. - -~

4 -

?. N R 3 an0 ESFAS ADDLICABLE *' ODE ann arTinn NnTE".

i a 'icensee proposes to delete MODE 1 from the Table 3.3-1, page 3/4 3-2, unctional Unit 3a APPLICABLE MODES column. This trio which would occur at C.75". pcier is bypassed'before reaching MODE-1 and is not applicable to Mnne-1 operations.

Therefore, its deletion is acceptable.

The: licensee proposes to add _ additional modes of applicability for pres-surizer oressure-low and steam generator pressure-low trip setpoints to ensure acceptability of the main steamline break analyses. We have found acceotaoie the MSLB analyses as stated in Section 2.0 of this' report.

Therefore these changes should be made.

The affected pages are:

3/4 3-2, and-3/" 3-7.

The licensee also proposes to add the provision that TS 3.04 is not applicable to the CEAC's.in Table 3.3-1..This has previously been the case for the CPC's., With the proposed change the requirements for the CEAC's are consistent with presiously approved requirements for the CPC's.

We find this acceptable. The affected page is 3/4 3-3.

3.17 MARGINS WITH CEAC'S INOPERABLE The licensee proposes to increase the required margins of TS 3.2.1 and 3.2,4 with COLSS out-of-service from greater than or equal to 8 percent to gieater than or equal to 11 percent based on the Cycle 2 reanalysis of th) CEAC; inoperable margin requirements.. This increase in the margin provided is acceptable.

The affected page is 3/4 3-Sa.

3.18 CONTAINMENT PRESSURE, TEMPERATURE AND HUMIDITY i

By letter dated November 19, 1979 proposed additional limits on the acceptable ~

combinations of containment pressure and temperature to be included in TS i

Figure'3.6-1. The acceptability of this change is addressed in Section 2.7.1 of this change it, addressed -in Section 2.7.1 of this report. The affected page is 3/4 6-7.

3.19 REFUELING MACHINE The licensee proposes.a. change.to.TS 3.9.6.to delete CEAs from items to be moved by the refueling machine..The licensee states tnat the-l refueling machine does not include provisions for moving CEAs and notes that

'CEAs are instead moved with a mantral tool or with a CEA change mechanism located over the fuel transfer me' vnism upender. Therefore this change

^

in the TS is required to reflect the actual design capabilities of the refueling machine and the handling practices of the licensee and is acceptable.

The affected pages are:

3/4 9-7 and B 3/4 9-2.

3.20; PRE 55URIZE2 SAFETY VALVE CAPABT!!TY ide-iicensee proposes a TS change to the CASES for pressurizer safety valve testing to amend the designated relief capability of 395,000 lb m/hr.to 420,000 lb m/hr. This change i: made to update-the preliminary

. design valve (395,000) to the actual rated valve (420,000) and is acceptable.

, The-page affected is 8 3/4 4-1.

3.21 MISCELLANE0US CHANGES Various TS pages are changed due to a change in the page number due to other new pages being added, due to correction of typographical errors and other changes of an administrative non-safety related nature. These changes are acceptable. The affected pages are:

3/4 2-5, 3/4 2-6, 3/4 2-7, 3/4 2-8, 3/4 2-9, 3/4 2-11, 3/4 6-18 and 5-5.

4

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4.01 PHYSICS IESTING

'The startup physics test program.as outlineo by tne licensee was reviewed.

..The precritical-tests include control element assembly trio. tests and reacter coolant flow coastoown tests..The low power tests-include critical bcron concentration, CEA: symmetry, and temperature reactivity worth tests. Power escalation tests include core power distribution tests at 50 percent and 1100 percent power.and. isothermal temperature coefficient ano power coefficient

. tests.at 50 and 100 percent power. The acceptance criteria supplied for each

. test was reviewed as well as the procedures.if acceptance criteria were not met. We find this physics startup test program-acceptable.'

5.0 ENVIRONMENTAL CONSIDERATION

We have determined that this amendment does not authorize a change in effluent typestor total amounts nor an-increase in power level and.will not result in any significant environmental. impact..Having made this determination,'we.have further concluded that the amendment involves an

~ action which.is insignificant from the standpoint of environmental; impact' and pursuant to 10 CFR'51.5(d)(4) that an environmental impact statement,.

or negative declaration and environmental impact appraisal need not be

prepared iniconnection with the issuance of this amendment.

6.0 CONCLUSION

We have concluded, based on the considerations discussed above, that:

(1) because the amendmentzdoes not involve a significant increase in the prob-ability or consequences.of accidents previously-considered and does not involve a significant decrease in a safety margin, the amendment does not

. involve a significant hazards consideration, (2) there is reasonable assurance.that the health and safety of the public will not be endangered.

by operation in the proposed manner, and (3) such activities will be con-ducted in compliance with.the. Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security l

or to the health and safety of the public.

19, 1981 l

Date: June l

{

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4 7.0'. REFERENCES 1.0 -Introduction There are two basic references which comprise the ANO-2 Cycle 2 Reload Re: Ort these being the submittals from the licensee dated February 20 and.

March 5, 1981. However numerous other documents were generated by the staff and by the licensee in support of the reload review and other issues addressed in this Safety Evaluation. Therefore,-for convenience's sake on the following pages the references are listed under the SE section to which they apply.

l l'

l c

l l

-... _, _.., _, _ ~. -...

. ~...

, 2.1 Fuel Design L

1.

Letter from D. Trimble (AP&LCol to R. A.. Clark (NRC),

Subject:

Cycle 2 Reload Report, dated February 20, 1931.

2.

Letter from W. Cavanaugh,.III, (AP&LCo) to R. A. Clark (NRC),

Subject:

Cycle 2 Reload Report, dated March 5. 1981.

3.

Letter from D. Trimble (AP&LCo) to.R. A. Clark (NRC),

Subject:

Information Regarding AN0-2 Reload Report,-dated Acril 30, 1981.

4.

"The Evaluation and Demonstration of Methods for Improved Nuclear Fuel Utilization First Semi-Annual Progress Report: -Inception to June 30, 1980,"

C-E draft report CENPD-384, October 1980.

5.

" Test Fuel Rod Irradiation:

16X16 Nuclear Reactor," C-E report CENPD-256-P-A, August 1977.

6.

" Fuel Evaluation Model," C-E report CENPD-139-A, July 1974.

7.

Letter from 0. D. Parr (NRC) to F. M. Stern (C-E), dated December 4, 1974.

8.

Letter from D.1F. Ross, Jr., (NRC) to A. E. Scherer (C-E), dated November 23, 1976.

9.

Letter from R. S. Boyd (NRC) to W. Cavanaugh, III, (AP&LCo),

Subject:

Issuance of Amendment No. 1 to Facility Operating License NPF-6 (AN0-2), dated September 1, 1978.

10. Letter from D. Trimble (AP&LCo) to R. A. Clark (NRC),

Subject:

Responses to NRC Questions on ANO-2 Reload, dated Ma, 6, 1981.

l 11.

"CEPAN Method of Analyzing Creep Collapse of Oval Cladding," C-E report CENPD-187-A, March 1976.

12.

" Fuel and Poison Rod Bowing," C-E report CENPD-225, Supplement 3-P, June 1979.

13. Memorandum from D. F. Ross, Jr., and D. G. Eisenhut (NRC) to D. B. Vassallo and K. R. Goller, " Interim Safety Evaluation Report on the Effects of Fuel Rod Bowing in Thermal Margin Calculations for Light Water Reactors," dated December 8, 1976.

14. Memorandum from D. F. Ross, Jr., and D. G. Eisenhut (NRC) to D. B. Vassallo and K. R. Goller, " Revised Interim Safety Evaluation Report on'the Effects of Fuel Rod Bowing in Thermal Margin Calculations for Light Water Reactors,"

(

dated February 16, 1977.

l I-

. ~.

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15.. Letter from R. A. Clark (NRC) to x. Cavanaugh, III. (AP&LCo), dated April.10, 1981.

16.

" Safety Evsluation Report related tc the operation of Arkansas Nuclear r

One, Unit 2," Section 4.4, NRC report NUREG-0308, Supplement.1, June 1978.

17.

"Zircaloy Growth In-Reactor Dimensional Changes in Zircaloy-4 Fuel Assenhlies,"'

C-E report CENPD-198, December 1975.

-18.

"Zircaloy Growth Application of Zircaloy Irradiation Growth Correlations fer

.the Calculation of Fuel Assembly and Fuel Rod Growth Allowances," C-E report CENPD-198, Supplement-1, Decmeber 1977.

-19

" Response to' Request for Additional Information on CENPD-198-P, Supplement I,"

C-E report CENPD-198, Supplement'2-P, November 1, 1978.

20. Letter from R. L. Baer (FRC) to A. E. 5:herer (C-E), dated August 21, 1979.

21. Letter from K. Kniel(NRC) t1 A. E. Scherer (C-E), dated June 22, 1976.

22.. PNO-77-221, preliminary notification of event on unusual occurrence of guide

. tube wear, December 14, 1977.

23.. Letter from A. E.-Scherer (C-E).to-V. Stello (NRC), dated December 23, 1977.

24.- Letter from W. Johnson (MYAPCo) to V. Stello (NRC), dated February 14, 1978.

25. Letter from A. E. Lundvall, Jr., (BG&ECo) to V. Stello (NRC), dated February 17,.1978.

26.. " Safety-Evaluation Report related to the operation of Arkansas-Nuclear Or.e, Unit 2," Section 4.2, Supplement 2, September 1978.

27. Letter from D. Trimble (AP&LCo) to R. A. Clark (NRC),

Subject:

E0C-1 CEA Guide Tube Surveillance Program, dated March 30, 1981.

28. Letter from D. Trimble (AP&LCo) to R. A. Clark (NRC),

Subject:

Preliminary Results of ANO-2 Fuel Inspection, dated May 22, 1981.

29 Letter from D._Trimble (AP&LCo) to R. A. Clark (NRC),

Subject:

NRC Request for Information on Fuel Assembly Spacer Grid Damage, dated June 4, 1981.

I-l

4 4

49 -

2,2 Nuclear Analysis 1.- Letter from R. A. Clark (NP.C) to W. Cavanaugh, III, (AP&LCo) dated March 23, 1991 transmitting six pnysics uuestions.

2.

Letter from D. C. Trimble (AP&LCo) to Director, NRR dated April-14,1981 transmitting responses to staff's March 23, 1981 questions.

3. ' Letter.from R. A. Clark (NRC) to W. Cavanaugh, III, (AP&LCo) dated April 29, 1981 transmitting five additional physics questions.

4.

Letter-from O. C. Trimble (AP&LCo) to Director, NRR dated May 11, 1981 transmitting responses.to the staff's April 29, 1981 questions.

5.

Letter from'O. C. Trimble_(AP&LCo)'to Director, NRR dated May 27, 1981' transmitting information on fuel assembly misloading analyses.

i l

l l

u I

L

i

~

~

Gl 2.5-Reactor Protection Syste:

1.

Letter from D. C. Trimole (AP&LCol to Director, NRR dated Seotenber 3,1980 CPC/CEAC - Plant Computer C:talink.

2.

Letter from D. C. Trimble (AP&LCo) to Director, NRR dated Feb'ruary 28, 1981,-CPC/CEAC - Plant Comouter Datalink.

3.

Letter from R. A. Clark (NRC) to W. Cavanaugh, III, (AP&LCo) dated April 10, 1981, Part II - Instrumentation and Controls System.

4.

Letter from R. A. Clark (NRC) to W. Cavanaugh, III, (AP&LCo) dated May 5',

~

1981 requesting documentation of addressable constants modification procedures.

- 5.

Letter from W. Cavanaugh, III,-(AP&LCo) to Director, NRR dated May 19, 1981 responding to Part II of staff's April 10, 1981 letter.

- 6.

Letter from 0. C. Trimble (AP&LCO) to Director, NRR dated May 26, 1981 providing a document in response to staff's May 5,1981 letter.

l l

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,-t.y----

t-w

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2.6 License Conditions 1.

Letter from D. A. Reuter (AF&LCo) tcfJ. F. Stolz (NRC) dated 0:t:t:r ll,

-1977 proposing a design for long term overpressure protection eauioment.

2.

Letter from D. C. Trimble (AP&LCo) to Director, NRR dated Decemoer 1, 1980 responding to the license condition and the SER open items.

3.. Letter from D. C. Trimble (AP&LCo) to Director, NRR dated Marcs 30, 1981 responding to staff questions.

2.7 Other Matters 1.

Letter from W. Cavanaugh, III, (AP&LCo) to Director, NRR, dated November 19, 1979 requesting change to TS Figure 3.6-1 on containment pressure, terperature and relative humidity.

2.

Letter from W.-Cavanaugh -III, (AP&LCo) to Director, NRR dated November 27, 1979 requesting change to TS Table 2.2-1 on high, pressurizer trip setpoints.

t r

q s

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w

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,,--e w

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