ML19345G321

Safety Evaluation Supporting Amend 52 to License DPR-51

ML19345G321
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Site:	Arkansas Nuclear
Issue date:	03/09/1981
From:	Office of Nuclear Reactor Regulation
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ML19345G314	List: ML19345G313 ML19345G321 ML19345G326
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NUDOCS 8103180171
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>3 MC UNITED STATES

,og NUCLEAR REGULATORY COMMISSION

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WASHINGTON, D. C. 20555

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATI0ft SUPPORTING AfiENDMENT NO. 52 TO FACILITY OPERATING LICENSE NO. DPR-51 ARKANSAS POWER AND LIGHT COMPANY ARKANSAS NUCLEAR ONE, UNIT NO. 1 DOCKET N0. 50-313 1.0 Introduction By : letter dated -January 30,1981(1), as supplemented February 12(2), and 26(3),1981, Arkansas Power and Light Company (AP&L or the licensee) requested amendment to the Appendix A Technical Specifications (TSs) of the Arkansas Nuclear One, Unit No. I (ANG-1) Liceme No. DPR-51 which would permit ' power operation during Cycle 5.

Cycle 4 was tenninated after 329 effective full power days (EFPD) and Cycle-5 has a design length of 435 EFPD.

2.0 Evaluation 2.l' Fuel Assembly Mechanical Design The sixty-eight Babcock and Wilcox (B&W) Mark B-4 ISx15 fuel assemblies loaded as Batch _7 at the end of Cycle 4 (E0C 4) are mechanically inter-changeable with Satches 5B and 6 fuel assemblies previously loaded at ANO-1. Fuel assemblies _ of the Mark B-4 design were.used in the previous refuel at ANO-1 and the design 'is used in other B&W nuclear steam supply systems. Two' assemblies will contain regenerative neutron sources, and

retainers will be used to contain-the sources. Justification for the

~

~ design and use of the neutron source retainer is described in the " Burnable Poison Rod Assembly Retainer' Design Report" (4). A discussion of the burnable poison rods themselves 'is presented in Section 2.1.1 of this evaluation.

' 2.1.1 ' Reactivity Control System In addition to the permanent reactivity control system (soluble boron and control rods), 64 burnable poison rod assemblies (BPRAs) will be used to control reactivity changes due to fuel burnup and fission product buildup.

The BPRAs are normally removed from the reactor at the end of first' cycle' and reinserted only for-extended cycle operation, such as the previous-Cycle 14 and the proposed Cycle 5 operation at ANO-1.

In April

.1978,. two BPRAs were accidentally ejected from the core of another B&W-

designed. reactor lat. Crystal River (5). The ejected BPRAs were carried-out of the reactor _ vessel by the coolant flow to the steam generator,

where'significant damage to the steam generator tube ends resulted. _B&W determined, that the ejection of the BPRAs from the core resulted' from B103180 Q \\

. fretting wear in the holddown latching mechanism.

In order to avoid similar problems at other plants, B&W redesigned and replaced the BPRA holddown mechanism on all operating B&W cores.

The NRC staff recently approved (6) the new design. We threfore conclude that changes to the core reactivity control system have been adequately considered for Cycle 5 operation.

2.1.2 Fuel Rod Design Although all batches in ANO-1 Cycle 5 utilize the same Mark B-4 fuel, the Batch 7 assemblies incorporate a slightly higher initial fuel density, shorter active fuel length, and larger initial gap size.

The change in fuel initial density, from 94 to 95 percent of theoretical density, is a consequence of using a modified fuel fabrication process. The stability (densification resistance) of both fuel types is similar. As a conse-quence, the densified fuel stack height.and gap size are virtually unchanged for the -Batch 7 assemblies.

2.2.1 Cladding Collapse The licensee has stated that the cladding collapse analysis in the Cycle 5 Reload Report is bounded by conditions previously analyzed in the ANO-1 FSAR or analyzed specifically for Cycle 5 conditions using methods and limits previously reviewed and approved by the NRC. We conclude that additional staff review of the cladding collapse analysis is unnecessary for Cycle 5 cperation.

2.2.2 Cladding Stress The licensee has stated that the cladding stress analysis described in the Cycle 5 Reload Report is bounded by conditions previously analyzed in the ANO-1 FSAR or analyzed specifically for Cycle 5 conditions using methods and limits previously reviewed and approved by the NRC. We conclude that additional staff review of the cladding stress analysis is unnecessary for Cycle 5 operation.

2.2.3 Cladding Strain The licensee has stated that the cladding strain analysis described in the CyJe 5 Reload Report is bounded by conditions previously analyzed in the ANO-l FSAR or analyzed specifically for Cycle 5 conditions using methods and limits previously reviewed and approved by the NRC. We con-clude that additional staff review of the cladding strain analysis is unnecessary for Cycle 5 operation.

2.2.4 Rod Internal Pressure

ection 4.2 of the Standard Review Plan (SRP)(7) addresses a number of acceptance criteria used to establish the design bases and evaluation of the fuel system. Among those parameters which may affect the operation of the fuel rod is the internal pressure limit. The acceptance criterion for this (SRP 4.2,Section II.A.l(f)) is that fuel rod internal gas

-pressure should remain below normal system pressure during normal opera-tion unless othenvise justified.

. The licensee has stated (l) that fuel rod internal pressure will not exceed nominal system pressure during normal operation for Cycle 5.

This analysis is based on the use of the B&W TAFY code (8) rather than a newer B&W code called TAC 0(9). Although both of these codes have

-been approved for use in safety analysis, we believe(10) that only the newer TAC 0 code is capable of correctly calculating fission gas release (and therefore rod pressure) at very high burnups.

B&W has responded (11) to this concert, with an analytical comparison between both codes.

~In that response, they-have stated that the internal fuel rod pressure credicted by TAC 0 is lower than that predicted by TAFY for fuel rod exposures of up to 42,000 PWd/Mtu. Although we have not examined the comparison, we note that the analyses exceed the expected exposure

~ (35,309 mwd /MtU) in ANO-1 at the end of Cycle 5.

We conclude that the rod internal pressure limit has been adequately considered.

2.3 Fuel Thennal Design The average fuel temperature as a function of linear heat rate and life-J-

time pin pressure data used in the LOCA analysis (Section 7.2 of the Reload submittal) are also calculated with the TAFY code (8). B&W has stated (l) that the fuel temperature and pin pressure data used in the generic LOCA analysis (12) are conservative compared with those calculated for Cycle 5 at ANO-1.

. As previo~usly mentioned in Section 2.2.4.of this evaluation, B&W currently has; two_ fuel performance codes, TAFY(8) and TAC 0(9), which could be used

. to calculate-the LOCA initial conditions. The older code TAFY has been used for the Cycle 5 LOCA analysis. Recent infomation(13) indicates that the TAFY code predictions do not produce conservatively higher peak cladding temperatures than. TACO for all Cycle 5 monditions as suggested in Ref..ll.. The issue involves -calculated fuel rod internal gas pressures that are too low at beginning of life. The rod internal pres-sures: are used to detemine swelling and rupture behavior during LOCA.

B&W has1 proposed (14) a method of resolving this= issue which was accepted by the staff (15).. The method, which is applicable to ANO-1, involves

- the use.of: reduced LOCA kW/ft limits at low core elevations during the first 50 EFPD cf operation (se'e Table 7-1 of Ref.1). The licensee has incorporated these changes into the ANO-1 TSs to support Cycle 5 operation.

W have reviewed these changes' and find them acceptable. We the_refore conclude that the initial thermal conditions for LOCA analysis have been appropriately considered for Cycle 5 operation.

-2.4' Material Compatibility:

'*he chemical and material compatibility of_ possible fuel cladding and coolant: interactions is unchanged from the previous cycle of operation.

The impact-of this issue on the operational safety of ANO-1 need not be reconsidered for Cycle 5 operation.

2.5~

l Operating Experience B&W'has accumulated operating experience with the Mark B 15x15

. fuel assembly at all of the eight operating B&W 177-fuel assembly plants.

A sumnary of this operating experience as of July 31, 1980 is given on page 4-3 of Ref, l.

2.5.1 Fuel Failures Approximately 40 EFPDs into Cycle 4 operation at ANO-1, an increase in the reactor coolant system iodine activity level indicated that some fuel failures had occurred (16).

Continued Cycle 4 operation for ap-proximately 300 EFPDs at higher, but relatively stable, iodine levels indicated that no additional failures had occurred.

Fuel sipping was thus performed during the Cycle 4-5 refueling outage to locate the leaking assemblies.

All 177 assemblies in the core were sipped and a total of 24 assenblies were found to contain leaking fuel rods.

Based on the licensee's calcu-lations, it was estimated that an average of three rods per assembly were.

leaking. These were divided into nine Batch-4 assemblies, six Batch-5 assemblies and nine Batch-6 assemblies.

It is the licensee's intention to reinsert five leaking assemblies for continued operation during Cycle 5.

The reactor coolant system activity levels during Cycle 4 operation, after the fuel failures occurred, varied from 0.3 to 0.4 pCi/gm with a December 31, 1980,. level of 0.196 pCi/gm prior to shutdown for refueling. These values were well below the ANO-1 TS limit of 3.5 pCi/gm. The licensee estimates RCS activity levels in Cycle 5 will be less than 0.1 pCi/gm at startup.

Our interest-in this issue is based upon three fundamental concerns:

Fi rs t,

ihat coolant activity levels are kept as low as reasonably achievable and well within TS limits and safety analysis assumptions; second, that the cause of the failures be reduced or eliminated at the plant in question and also at. reactors of similar design; third, that NRC receive prompt notification in the event that further failures of this kind occur in Cycle 5.

In response to the first concern, the licensee has _ stated (16) that every reasonable effort has been made to lower the number of failed fuel assemblies and to mitigate their effects in order to satisfy ALARA criteria. The licensee further stated (17) that it would cost more than $7.5 million to remove and replace the five failed fuel assemblies at this time.

Based on this cc,t figure and estimates of the expected coolant activities in Cycle 5, we have determined that this situation is acceptable and does not warrant further review.

It should be noted, however, that the large cost quoted by; AP&L was the result of reinserting the five failed assemblies and reinstalling the reactor vessel head before making the cost estimate (and thereby creating a critical path. item),. and that our acquiescence in this _ instance should not be construed as generic approval to reuse damaged fuel.

6

. In response to our pecond concern, the licensee has stated that the fuel vendor, B&W, has reviewed the operating conditions at the time of the failed fuel occurrence and has been unable to find conditions that could have led to fuel failure.

B&W has also reviewed the manufacturing QA records of the identified failed fuel assemblies and has been unable to find an indication of a generic problem with respect to manufacturing defects.

For Cycle 5 operation, the licensee is reviewing the control rod maneuvering criteria to determine if any revisions are in order.

From the limited infomation currently available to the staff, it appears that the Cycle 4 fuel failures are not random. This assumotion leads us to believe that the cause of the failures stems either from (a) a possible fuel desi operational deficiency. gn deficiency or (b) a possible fuel-related Either cause would not only impact Cycle 5 operation at ANO-1, but also future cycles at this and other B&W-fueled reactors.

The licensee has connitted(17 and 18) to further investigate the problem and report the results to the NRC within six months.

In view of (a) the licensee's commitment to a thorough investigation that should reveal the cause of the failures within 6 months, (b) the apparent small effect that these failures are expected to have on normal operation during Cycle 5, and (c) the low probability of occurrence of a plant transient or accident whose course might be affected by this failure mechanism before such effects could be evaluated, we believe that our second concern has been adequately addressed.

In response to the third _ concern, the licensee has agreed to notify the NRC Resident Inspector by the 10th day of each month, in writing, as to the status of the equilibrium RCS activity levels from the previous month's operation.

This note will discuss significant variations in equilibrium RCS activity levels and the licensee's estimation of how this might relate

,to the amount of failed fuel present.

This written note will be supplied in this manner to ensure more direct NRC notification of changes in failed fuel levels as soon as practical and in a documented manner.

2.5.2 Guide Tube Wear Significant wear of Zircaloy control rod guide tubes has been observed in fecilities designed by Combustion Engineering. Similar wear has also been reported in facilities designed by Westinghouse.

In a letter dated June 13, 1978, we requested infomation from' B&W on the suscepti-bility of the. facilities desi.gned by B&W to guide tube wear.

The infor-lmation provided by B&W in a letter dated January 12, 1979 was insufficient

-for us to ' conclude that guide tube wear was not a significant problem in B&W plants.

This insufficiency was documented in our letter to B&W dated August 22, 1979.

J.

L Because guide tube wear could result in loss of control rod scram capability and also fuel assembly. structural integrity, we consider this wear phenomenon a potential safety concern. Therefore, we re-quested (19) additional ;information from the licensee on control rod guide tube wear.

In response to this request the licensee transmitted the B&W Control Rod Guide Tube Wear Generic Report, which is applicable to ANO-1. The report provides information on post-irradiation examin-L ations of guide tube wear in Oconee 1 and 3 and in Rancho Seco. The results of these measurements indicated that through-wall wear or ex-i cessive wall degradation will not likely occur during anticipated fuel residence time for rodded assemblies. Although we have not yet completed our review of that report, on the basis of our preliminary evaluation, 4

we conclude that guide tube wear has been adquately addressed for ANO-1

'during Cycle.5.

2.5.3 Holddown Speing Failures The upper end. fitting of the B&W Mark B-4 fuel assembly contains a hold-down spring to accommodate length chan.qes due to thermal expansion and irradiation growth while providing a positive holddown force for the assembly. On May 14, 1980, a failed holddown sprin

remote video inspection at. Davis Besse, Unit 1 (21)g was discovered by Further examination ultimately identified a-total of 19 failed springs in the Cycle 1 fuel

' assemblies. Subsequent' examination of spent fuel assemblies at other B&W reactors. revealed a small number of similar failures at Crystal

~ River 3 (22),- Oconee 1 (23), and at ANO-1. for the current refueling outage (24).

We have reviewed the BW hbiddown spring failures as a generic issue

^

-(25).. The predominant mode of failure appears to.have been fatigue

-initiated cracking followed-by stress corrosion crack propagation in springs with an improper metallurgical condition (grain size).

Based upon our review of information provided in a meeting with Toledo Edison'-(Davis-Besse) in June 1980 and responses to staff ques-tions issued to all B&W licensees in July.1980, we believe that there

is reasonable assurance that the holddown spring fail _ures will not irecur. on allarge scale, and that neither the potential. for loss of

- positive holddown force, loose parts, nor interference with normal control rod movement constitute a significant ' safety hazard.

Nevertheless,;because the recent holddown spring failure observed-at

~

ANO-1;does not appear: to be related to material of. improper metallur-

~

gical condition, and because 'some lateral and vertical motion of loose

as'semblies is possible under. certain. extreme conditions, we-haveiconcluded that further' surveillance (e.g., video examination)

L of the-assembly. - holddown ' springs'should be. carried out at the next refueling at ANO-l'. EThe. licensee has ' agreed '( 24) to perform a 100%

! inspection.of the core fuel assembly 'holddown springs during the next

' refueling outage and.will report the.results of the inspection-to the NRC '

i j_

a 1

O

• l On the basis-of B&W's analysis of the consequences of operating with l

failed holddown springs, the completion of our generic evaluation of the problem, and the results of the licensee's inspection of all Cycle-5 assemblies, repair of the damaged assembly, and commitment to inspect all holddown springs at the end of Cycle 5 operation, we r.onclude that there is reasonable assurance that the holddown spring issue has been correctly analyzed and that this issue does not pre -

sent a safety concern for Cycle-5 operation.

If further examinations l-are found to be necessary, the licensee will be notified.

?.6 Fuel Rod Bowing The licensee has stated that a fuel bowing penalty has been calcu-lated according to the procedure that was approved in Reference 27.

The burnup used in that calculation was the maximum fuel assembly burnup of the batch that contains the limiting fuel assembly.

For Cycle 5, this burnup is less than 35,309 tWd/MtU. The resultant bowing penalty was found to be a 4.8% reduction in DNBR.

To offset the' 4.8% penalty, the licensee has drawn upon both generic and plant-specific margins. The generic margin employed was a thermal credit equivalent to 1% DNBR. This credit is a result of the stand-ard flow-area-reduction factor included in all B&W hot-channel thermal-hydraulic analyses. The plant-specific margin employed was a 10% DNBR credit available because plant operating limits were set at conservative values that correspond to the original method (28) of calculating rod bowing penalties rather than the new procedure. We conclude that the DNBR reduction due to fuel rod bowing has been conservatively calculated for Cycle 5 operation.

In order to provide for a proper accounting of margins used to offset the DNBR penalty, we required -- as witn other operating reactors -- that the bases for. the Technical Specifi-cations for ANO-1 be amended to identify each generic and plant-speci-fic margin that was -used. The licensee provided such an amendment as part of the Reload Report (2) and we will include this amendment in the bases for the AN0-1 Technical Specifications.

2.7 Extended Burnup Lead Test Assemblies Four test assemblies of a revised design will be included in the ANO-1, LCycle-5 core loading. These extended burnup lead test assemblies (LTAs) are similar in design to the standard B&W 15x15 Mark B-4 fuel assemblies except for changes to -the fuel rod and fuel assembly structure to extend their burnup capability._ The. revised design, which is described in Refer-en'ce. 31 -is tertned the Mark BEB. Two of the four Mark BEB assemblies will contain. segmented fuel rods and will' have a specially designed end fitting for removal of these rods in the reactor fuel pool upon assembly discharge.

The segmented rod design is comprised of five individual fuel segments, three of which are representative of a full-length rod.

The base Mark BEB design employs fuel. rods of a solid pellet designi how-ever, four full-length rods in each assembly and several segmented rods contain annular pellets to gain incore high burnup experience with an

O

, annular fuel design. The annular pellet design was selected because of its lower operating fuel temperatures, which result in reduced fission gas release from the fuel pellet. Lower end-of-life fuel rod internal pressures result from the annular pellets' combination of lower operating temperatures and the increased void volume from the pellet central void.

The heat treatment for the guide tube and instrument tube material was changed from stress relieving to full annealing to reduce fuel assembly irradiation growth, which has been identified as a limiting condition for extended burnup operation of standard Mark B-4 fuel design, i

The licensee has stated (31) that, based on mechanical, nuclear, and thermal hydraulic analyses, the loading of four extended-burnup LTAs in the ANO-1, Cycle 5 core will not adversely affect the performance characteristics of the reactor and will be bounded by existing safety analyses. We have examined the LTA design report and find that it closely parallels the analyses presented for the standard fuel design.

There are, however, some exceptions. As an example, an unreviewed fuel performance code called TACO-2 (27) was used for the fuel thermal and mechar,1 cal performance analysis rather than the previously approved TAFY (8) or TAL3-1 (9) codes.

In response to our concern over the use of unapproved models in the LTA safety analysis, the licensee has stated (29) that the LTA's meet all of the thermnl design criteria for Cycle 5 operation using either the TAFY or TACO-1 codes. On the basis of this statement, we find the LTA furl system design analysis acceptable as submitted. for Cycle 5 operation.

For operation during subsequent cycles, it must be shown that the fuel design criteria continue to be met with the TAFY, TACO-1, or if approved, TACO-2 models.

In addition to the safety analysis of the LTA fuel design, we believe i

that a substantial level of fuel surveillance is necessary to support the irradiation of these assemblies. The reason for this position is that the. surveillance of the lead test assemblies will be required in support 'of other full-core reloads using the Mark BEB design, whereas l

the same surveillance would not (generally) be required to assure the

.sa fety o f Cycl e 5.

The simple fact of successful irradiation of LTA's I

without detailed technical examination, would not be sufficient to support a full-core reload of the LTA design.

In response to our concern on surveillance, the licensee has stated that all four LTA's are to be extensively characterized before irradiation and examined after each cycle of operation. We have discussed details of the examination program an'd. find the level of surveillance commensurate with a lead prototype irradiation.. We conclude.that the design and irradiation of the four LTAs 'in AN0-1 Cycle 5 is acceptable.

. 2.8 Nuclear Design Cycle 5 of ANO-1 has a design length of 435 effective full power days (EFPD). This compares to a Cycle 4 design length of 387 days and a 329 day actual length. The physics parameters for Cycle 5 were based on the actual Cycle 4 length.

The extended Cycle 5 length requires that burnable poison assemblies be included in the design.

In addition, for Cycle 5 the allowable linear heat generation limits in the lower portion of the core have been reduced for the first 50 EFPD of the cycle. To facilitate meeting these revised limits the mode of operation has been changed from the present rodded operation to feed and bleed operation.

We have reviewed the nuclear design and transient and accident analysis portions of the submittal. The physics parameters for the Cycle 5 core h' ave been evaluated by the same techniques that have been employed in previous ANO-1 cycles and shown to be adequate.

The presence of burnable poison in the core has been previously treated in the first cycle of ANO-1.

The results of the physics parameter calculations are compared to those for Cycle 4 and shown to be similar. The slight differences are dde to the different cycle lengths. We conclude that the physics parameters for the core are acceptable.

The core kinetics. parameters that are the inputs to the accident and transient analyses are compared to those used in the FSAR where the consequences of the various transients and accidents are shown.

The comparison shows that in all cases the parameters fall within the values used'in the original analyses. As a_ result no recalculation of these events was necessary. On thebasis that the kinetics parameters were obtained in the same way as in previous cycles and

[

that measurements of selected parameters will be made during startup to confirm their value, we find this approach to be acceptable.

l 2.9 Thermal and Hydraulic Design l-THe thermal and hydraulic design of the reload core was reviewed to confirm that it uses acceptable analytical methods and provides acceptable margins of safety from conditions which would lead to fuel damage during normal operation and anticipated operational transients.

2.9.1 Evaluation of Thennal-Hydraulic Design The thermal-hydraulic models and methodology used are described in References 32, 33 and 34. The main differences between Cycle 5 and Cycle 4 are discussed as follows:

+

, 2.9.1.1 Lead Test Assembly, The incoming Batch 7 fuel is hydraulically and geometrically similar to the fuel remaining from the previous cycles. Among the Batch 7 fuels, four are the LTAs, which, according to the licensee, in References 1 and 31, have been designed for extenJad burnup (>50,000 mwd /litU) operation and analyzed to ensure that the LTAs are never the limiting assemblies during the Cycle 5 operation.

2.9.1.2 Rod Bow DNBR Penalty The rod bow DNBR penalty applicable to Cycle 5, according to the licensee, was calculated using the interim rod bow penalty evaluation procedure approved in Reference 27, the burnup used to calculate the penalty was the highest assembly burnup in Cycle 5, 35,309 mwd /Mtu. The calculated rod bow penalty using the procedure is 3.8 percent. Utilizing the 1 percent DNBR credit for the flow area reduction hot channel factor, the penalty is 2.8 percent.

However, according to the licensee, all plant operating limits based on DNBR criteria include a minimum of 10 percent DNBR margin available due to the plant operating limits being set at conservative values that correspond to the original method (Reference 35) of calculating rod bow penalty rather than the new procedure given in Reference 27. The licensee wants to do this for their convenience of establishing the setpoints once for all future reloads.

Therefore, we find the minimum DNBR limit of 1.43 to be conservative and acceptable.

2.9.1.3 Flux / Flow Setpoint Increase As indicated in table 1, the Cycle 5 nuclear design allowed a

. reduction of the design radial local peak from 1.78 to 1.71, resulting in an increase of the steady state design overpower minimum DNBR from 1.88 to 2.05.

This provides an additional operating safety margin. Based on the DNBR criteria, including a minimum of 10 percent DNBR margin to offset the impact of and rod. bow penalty, the flux / flow setpoint of 1.07 increased from 1.057 for Cycle 4 has been established for Cycle 5.

2.9.1.4:

Technical Specification Changes To incorporate the change c. the flux / flow setroint into the Technical Specifications, tne licensee proposed modifications to the Core Protection Safety Limits of Technical Specification 2.1 (Figures 2.1-1 to 2.1-3 of Referenca 1) for the ANO-1 Cycle 5 reload.

s b

TABLE 1 ARKANSAS RELOAD Initial Thermal-Hydraulic Conditions for Cycles 4 and 5 Cycle 4 Cycle 5 Design Power Level, MWt 2568 2568 System Pressure, psia 2200 2200 Reactor Coolant Flow, % of Design 106.5 106.5 Vessel Inlet / Outlet Coolant Temperature at 100% Power, F 555.6/602.4 555.6/602.4 Reference Design Radial-Local Power Peaking Factor 1.78 1.71 Reference Design Axial Flux Shape 1.5 cosine 1.5 cosine Hot Channel Factors:

Enthalpy Rise 1.011 1.011 Heat Flux 1.014 1.014 Flow Area 0.98 0.98 Active Fuel Length, 140.2 140.2 Average Heat Flux at 100% Power, 175 175 10 Btu /h-ft (a)

Maximum Heat Flux at 100% Power, 468 469 10 Btu /h-ft (b)

CHF Correlation BAW-2 BAW-2 Minimum DNBR: at 112% Power 1.88 2.05 at 108% Power 2.01 2.18 at 100% Power 2.30 2.39 (a) Heat flux was based on densified length (in the hottest core location).

(b) Based on average heat flux with reference peak.

. 2.9.2 Nnclusion of Thermal-Hydraulic Design According to Reference 1, the licensee has examined each FSAR transient analysis with respect to the changes in Cycle 5 para-meters to determine their effect on the plant thermal performance during the analyzed transients. The important thermohydraulic parameters are sumarized and compared in Table 1 (Table 6.1 of Reference 1) for Cycles 5 and 4.

These parameters are the same for both cycles with the exception of the higher value of design minimum DNBR for Cycle 5 (2.05) as compared to 1.88 for Cycle 4.

The licensee concluded that the results of transients for Cycle 5 are bounded by the analyses in Reference 32, which is the previously approved FSAR for ANO-1, and thermal-hydraulic models and methodology used have been approved in References 32, 33, and

34. We, therefore, conclude that this core reload will not adversely affect the capability to operate AN0-1 safely during Cycle 5 and the proposed changes to the Technical Specifications discussed in Section 2.9.1.4 are acceptable.

3.0 Evaluation of Accidents and Transients The licensee has examinci each Final Safety Analysis Report (FSAR) accident analysis with respect to changes in Cycle 5 parameters to determine their effect on the plant themal performance during the analyzed accidents and transients. The key parameters having the greatest effect on the outcome of a transient or accident are the core thermal parameters, thermal-hydraulic parameters, and physics and kinetics parameters. Fuel thermal analysis values are listed in Table 4-2 of Reference 1 for all fuel batches in Cycle 5.

Table 6-1 of the same reference compares the themal-hydraulic parameters for Cycles 4 and 5.

These parameters are the same for both cycles with the exception of the higher value of design Maximum Departure to_ l.88 for Cycle 4)g Ratio (MDNBR) for Cycle 5 (2.05 as compared from Nucleate Boilin A comparison of the key kinetic parameters from the FSAR and Cycle 5 is provided in Table 7-1 of Reference 1.

These comparisons indicate no significant changes or changes in the conservativo direction. The effects of fuel densification on the FSAR accident analyses have been evaluated.

A generic Loss of Coolant Accident (LOCA) analysis for the B&W 177-fuel assembly, lowered loop Nuclear Steam Supply System (NSSS) has been performed using the final acceptance criteria Emergency Core Cooling System (ECCS) evaluation model (Reference 30 ).

That analysis used.the limiting values of key parameters for all plants in the 177-FA lowered loop category, and therefore is bounding for the ANO-1 Cycle 5 operation.

We conclude from the examination of Cycle 5 core thermal and kinetic properties, with respect to acceptable previous cycle values and with respect to the FSAR values,.that _this core reload will not adversely affect the ANO-1 plant's ability to operate safely during Cycle 5.

..C)

. 4.0 Technical Specification Changes The proposed modifications to the Technical Specifications have been reviewed and based on the fact that the specified safety and operating limits were obtained by the same means successfully employed in earlier cycles of ANO-1 and in other B&W supplied reactors, we find these proposed TSs to be acceptable.

5.0

. Environmental Consideration We hava determined that the amendment does not authorize a change in effluent types or 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 551.5(d)(4), that an environmental impact statement, or negative declaration and environ-mental impact appraisal need not be prepared in connection with the issuance of the amendment.

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

(1) because the amendment does not involve a significant increase in the probability 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 conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.

Dated:.bbrch 9, 1981 u -.

O

. REFERENCES 1.

D. C. Trimble (AP&L) letter to R. W. Reid (NRC) dated January 30, 1981, and transmitting Arkansas Nuclear One Unit ! Cycle ~5 Reload Report (BAH-1658), January 1981 and requesting amendment to the ANO-1 operating license for Cycle 5 operation.

2.

D. C. Trimble (AP&L) letter to R. W. Reid (NRC) dated February 12, 1981.

3.

D. C. Trimble (AP&L) letter to R. W. Reid (NRC) dated February 26, 1981.

4.

BPRA Retainer Design Report, Babcock & Wilcox Conpany Report BAH-1496, hay 1978.

W. P. Stewart (Florida Power Corporation) letter to C. Helson (1:RC) on 5.

" Crystal River Unit'Three Status Report - May 1,1978," dated liay 4,1978.

-6.

T. M. Novak (NRC) menorandum to E. L. Jordon (!!RC) dated Decerber 22, 1980.

7.

Standard Review Plan, Section 4.2 (Rev.1), " Fuel System Design," U. S.

Nuclear Re;ulatory Commission Report NUREG-75/087.

8.

C. D. Morgan and H. S. Kao, "TAFY-Fuel Pin Temperature and Gas Pressure Analysis," Babcock and Wilcox Company Report BAW-10044, May 1972.

9.

" TACO-Fuel Pin Performance Analysis," Babcock and Wilcox Company Report BAW-10087P-A, Rev. 2, August 1977.

10.

D. F. Ross, Jr., (NRC) letter to J. H. Taylor (B8W) dated January 18, 1978.

11.

J. H. Taylor (B&W) letter to P. S. Check (NRC), dated July 18, 1978.

12.

W. L. Bloomfield, et.al., "ECCS Analysis of B&W's 177-FA Raised-Leop NSS,"

Babcock and Wilcox Company Report BAW-10105, June 1975.

13.

R. O. Meyer (NRC) merorandum to L. S.' Rubenstein (NRC) on "TAFY/ TAC 0 Fuel Performance Models in B&W Safety Analyses," dated June 10, 1980.

14.

J. H.- Taylor (B&W) letter to L. S. Rubenstein (NRC) dated September 5,1980.

15.

L. S. Rubenstein (HRC) letter to J. H. Taylor (B&W) dated October 28, 1980.

16.

D. C. Trimble ( AP8L) letter to R. W. Reid (NRC) dated February 19, 1981.

t s e

'O,

. 17.

D. C. Trimble (AP&L) letter to R. W. Reid (NRC) dated February ?6,1081.

18.

D. C. Trimble (AP&L) letter to R. W. Reid (NRC) dated March 4, 1981.

19.

R. W. Reid (NRC) letter to W. Cavanaugh (AP&L) dated November 23, 1979.

20.

D. C. Trimble (AP&L) letter to R W. Reid (NRC) dated October 21, 1980 and transnitting Control Rod Guide Tube Wear Measurement Program (BAW-1623)

June 1980.

21.

T. D. Murray (Toledo Edison) letter to J. G. Keppler (NRC/ REG. III) dated May 23,1980.

22.

J. A. Hancock (Florida Power) letter to J. P. O'Reilly (NRC/ Reg. II) dated May 29, 1980.

23.

W. O. Parker, Jr., (Duke Power) letter to J. P. O'Reilly (NRC/ Reg. II) dated June 6,1980.

24.

D. C. Trinble ( AP&L) letter to R. W. Reid (NRC) date February 13, 1981.

25.

L. S. Rubenstein (NRC) memorandum for T. M. Novak (NRC) on "B&W Fuel Assembly Holddown Spring Failures" dated December 20, 1980.

26.

R. W. Reid (NRC) letter to W. Cavanaugh (AP8L) dated February 9,1981.

27.

L. S. Rubenstein (NRC) letter to J. H. Taylor (88W) on " Evaluation of Interim Procedure for Calculating DNBR Reduction due to Rod Bow," dated October 18, 1979.

28.

D. F. Ross and D. G. Eisenhut (NRC) memorandum to D. B. Vassallo and K. R. Goller (NPT) on " Interim Safety Evaluation Report on the Effects of Fuel Rod Bowing on Thorral Margin Calculattens for Light Water Reactors" dated December 8, 1976.

l 29.

Y. H. Hsii, et al.; " TACO-2:

Fuel Pin Perfomance Analysis", Babcock &

l Wilcox Company Report BAW-10141P, January 1979 l

30.

ECCS Analysis of B&W's 177-FA Lowered-Loop NSS, BAN-10103, Rev. 3 Gaocock

& Wilcox, September 1975.

l l

. < 31.. Extended-Burnup Lead Test Assembly, BAW 1626, dated October,1980.

32. Arkansas Nuclear _0ne, Unit. 1 - Final Safety Analysis Report, Docket 50-313, Arkansas Power and Light.

33. Arkansas Nuclear One, Unit 1 - Cycle 4 Reload Report, BAW-1504, Babcock and Wilcox, October,1978.

34. Arkansas Nuclear One, Unit 1 - Fuel Densification Report, BAW-1391, Babcock and Wilcox, June 1973.

35.- ' Letter from D. F. Ross and D. G. Eisentut (NRC) to D. B. Vassallo and K. R. Goller (NRC), " Revised Interin EER on the Effects of Fuel Rod Bowing on Therral Margin Calculations for Light Water Reactors," dated February 16, 1977.

I i