/ o UNITED STATES! ~ ,t NUCLEAR REGULATORY COMMISSION{ $ W ASHING TON, D. C. 20555

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION

NORTHERN STATES POWER COMPANY

PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT NOS.1 AND 2

DOCKET NOS. 50-282 AND 50-306

REMOVAL OF ROD CLUSTER CONTROL GUIDE TUBE THIMBLE PLUGS

1.0 INTRODUCTION

By letter dated April 22, 1985 Northern States Power Company, the licenseefor Prairie Island Nuclear Generating Plant Units 1 and 2, provided theresults of an analysis (Reference 1) for the removal of rod cluster control(RCC) guide tube thimble plugs. This removal is scheduled for the next reloadcycles (Cycle 11 for Unit I and Cycle 10 for Unit 2) and is to be done under10 CFR 50.59. The proposed change involves the removal of the 90 RCC guide

tube thimble plug assemblies from the core (each RCC assembly contains 16thimble plugs). New Westinghouse Optimized Fuel Assemblies (OFA) will beinserted in the next reload cycles. The OFA have smaller diameter guidetubes and, therefore, the existing thimble plugs will not fit in. A numberof reasons were presented for removing the RCC guide tube thimble plugs.These included: costsaving(avoidingpurchaseandinstallationofnewthimble plugs), reduction in radiation dose (from plug repair anddisposition), and reduction in potential for loose parts.

The removal of the thimble plugs effects a decrease in active core flow sincethe hydraulic resistance through the bypass region is decreased, therebyincreasing the bypass flow. There will be an offset due to an increase intotal flow because of a decrease in overall core resistance. However, the

net effect is a decrease in active core flow. ,

The' decrease in active core flow decreases the fuel thermal margins for

minimum departure from nucleate boiling ratio (MONBR), fuel melt and peak l'

cladtemperature(PCT). Also, the increase in total core flow reduces themechanical design flow margin affecting lift off. The licensee haspresented information (Reference 1) that relates to the effect of the

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flow changes due to removal of the RCC guide tube thimble plugs.on the thermalhydraulic design analysis, accident and transient analysis, and mechanical and

j structural analysis (lift off). Our evaluation follows.

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| 2.0 EVALUATION -

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|2.1 Core Bypass Flow

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| The current design bypass core flow fraction for Prairie Island Units 1 and 2plants is 4.5% as stated in Reference 2. The major bypass flow paths and theapproximate bypass flow include: (1) holes in the internal support ledge to

j cool the head (0.3%), clearance between the discharge nozzle in the upper corebarrel and the reactor vessel (1%), flow through the guide thimbles in the,

! fuel assemblies (3.2%). The licensee presented information on bypass flowj calculations for two types of fuel (Exxon TOPROD and Westinghouse Optimized)I with the guide thimble plugs removed. These calculations indicated, as shown

in Table 1, that the increase in bypass flow with a full core of Exxon TOPROD,

3 fuel is 1.32% and for a full core of Westinghouse Optimized is 1.20%. These

calculations include 10% uncertainties on the K form loss coefficients takenin a conservative direction. The calculation is dependent on fuel type and iscycle dependent. In order to bound all reloads with any combination of the two

; fuel types the licensee proposes to use a conservative value of 1.5% for the

|bypass flow increase. Therefore, a new bypass flow fraction of 6.0% (4.5% +1.5%) is to be used. The staff has reviewed the calculations for the core bypass,

i flow increase and has found them acceptable.J,

j In addition, the licensee has conservatively calculated that with thej assumption of a 6.0% core bypass flow when the thimble plugs are remove <1, thei reactor coolant system (RCS) flow is increased a maximum of 0.6% (affects fuelj assembly lift-off forces).

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2.2 Thermal Hydraulic Design Analysis

The licensee has provided the results of their thermal hydraulic designanalyses for Prairie Island Unit 1 Cycle 10 and Prairie Island Unit'2 Cycle 9in which the RCC guide thimble plugs are removed. The design criteria for theExxon Nuclear Company (ENC) reload fuel are as follows:

(1) The minimum departure for nucleate boiling ratio (MONBR) will be 1.3 atoverpower using the W-3 correlation with corrections for non-uniform axialheating, cold wall effects, and a reduction in MONBR due to fuel rodbowing.,

(2) The fuel must be thermally and hydtaulically compatible with the existingfuel and the reactor core throughout the life cycle of the fuel,

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(3) The maximum fuel temperature at design overpower shall not exceed the,

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j fuel melting temperature.!

4 (4) The cladding upper temperature limits shall not exceed:!

Inner surface temperature 850*F,

Outer surface temperature 675"F

Average volumetric temperature 750*F,

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i 2.3 Core Hydraulic Compatibility

In Reference 3, the hydraulic compatibility of the reload fuel (TOPR00 andSTANDARD) is discussed. Because of similarities of the fuel assemblies the

j hydraulic campatibility of the fuel is not affected by the removal of the RCC: guide thimble plugs and is acceptable. -

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2.4 Thermal Margin

The most limiting operational transient was found to be the slow rod :

withdrawal. Table 2 provides the reference conditions for this analysis. "

Table 3 provides the thermal margin results for both the conditions of plugsin and plugs out and with and without rod bow penalty. The results in Table 3are bounding for Prairie Island Unit 1 Cycle 10 and Prairie Island 2 Cycle 9.The most limiting value for MONBR is 1.317 which represents the conditionswith the plugs out and includes a rod bow penalty. The rod bow penalty hasbeen calculated using an approved method (Ref. 4). The thermal margin isabove the allowable limit of 1.3 and is, therefore, acceptable.

The effect of rod bow on LOCA is independent of the bypass flow fraction andtherefore there is no change in effect because of the removal of the thimbleplugs.

2.5 Fuel Temperature Analysis

Analyses performed by the licensee (Ref. 3) have shown that increasing thecore bypass flow to 6% due to thimble plug removal will cause the active coremoderator temperature to increase less than 0.5'F. This will result in anincrease in the fuel temperature by approximately the same amount which isless than a 0.02% increase in peak centerline temperature. This is found tobe a negligible amount and is acceptable.

;:.'ety Limit Curvesi.

The Prairie Island safety limit curves, along with the associated overpowerand overtemperature AT setpoints are given in the Technical Specifications inSection 2.1. Tech Spec Figure 2.1-1 shows the currently acceptable safetylimit curves which are calculated as a loci of points where 100BR = 1.30.The staff asked questions relative to operation with the thimble plugs removedand the effect of this (core bypass flow increased from 4.5% to 6%) on thesafety limit curves. The licensee provided the results (Ref. 9) of an

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analysis which used values of core bypass flow = 6.0%, FaH= 1.55 and FQ = 2.32.A spectrum of seven bounding points along the constant pressure lines of 2400psia, 2250 psia, and 1700 psia of the Technical Specification Safety LimitsFigures TS.2.1-1 were recalculated. These ranged in percent of rated powerfrom 92% to 120%. The calculated MDNBR values ranged from 1.361 (p = 1700 psia,120% rated power) to 1.492 (p = 2400 psia, 92% rated power). These are wellabove a DNBR value of 1.3. Also, an additional point of 59% rated power(P = 1700 psia) was presented for which the DNBR was 2.255. We therefore find

the current safety limit curves (Technical Specification Figure TS. 2.1-1)acceptable as they are bounding for operation with the thimble plugs removed.

3.0 ACCIDENT AND TRANSIENT ANALYSIS

3.1 Plant Transient Analysis

The input conditions (flow, temperature pressure, power) for the analyses foroperational transients without the RCC guide thimbles removed are provided inReferences 6 and 7 for Prairie Island Unit 1 Cycle 10 and Unit 2 Cycle 9,respectively. The analyses were done using a conservative value for FAH of1.65 whereas the current Technical Specifications limit the FaH to 1.55.

The removal of the thimble plugs, which increases the core bypass flow, doesnot significantly affect the plant transient system analysis. The systemanalysis is based on core average conditions. These are not significantlychanged as the moderator temperature change is less than 0.5'F. Therefore, theinput conditions for the previous analyses without thimble plugs removal (Refs.6 and 7) are still valid. Also, removal of the RCC guide thimble plugs willnot affect the relative significance of the transients. The same transientswhich were found to be limiting will remain limiting for the plugs removedanalyses. However, the hot channel analyses response can be affected by smallchanges in core flow as is the case with the removal of thimble plugs. The

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effect of this is modeled in the plant transient analyses that were redone.

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Table 4 shows the thermal hydraulic parameters used for full power operation.An axial peaking factor, FZ, of 1.365 was located at X/L = 0.7.

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3.1.1 Transient Analysis Results

The limiting Class II and III transient was found to be the slow rodwithdrawal. The limiting Class IV accident was found to be locked rotor.These are discussed further below.

3.1.1.1 Slow Rod Withdrawal

Transient response information was calculated for slow rod withdrawal by thelicensee for the case with the plugs removed and showed that the DNB ratiodrops from an initial value of 1.792 to a minimum value of 1.363 at 38.6

seconds. See Table 5 for a comparison of MDNBR with and without the plugsremoved. These values are acceptable for operation with the thimble plugsremoved as the calculated MDNBR for the limiting Class II and III transient,the slow rod withdrawal, is above the acceptable minimum DNBR limit of 1.3 andthe maximum reactor coolant and main steam pressures do not exceed 110% of

their design values. These results bound both the Prairie Island Unit 1 Cycle10 and Unit 2 Cycle 9 operation.

3.1.1. 2 Locked Rotor

Transient response information was calculated by the licensee for the lockedrotor for the case where the plugs are removed and the number of fuel rodscalculated to experience DNB was found to be 18.2%. This is less than the

' acceptable FSAR limit of 20%. This value bounds both Prairie Island Unit 1i Cycle 10 and Unit 2 Cycle 9. Table 5 shows a comparison for MDNBR and percent; failed fuel for both the conditions with the plugs in and the plugs out.!

The acceptance criteria for the locked rotor analysis are as follows:|

I 1. The maximum reactor coolant and main steam system pressures must notexceed 110% of the design values.

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2. The number of fuel rods calculated to experience a DN8R of less than 1.3

should not exceed the number which would cause the doses due to releasedactivity to exceed a fraction of the limits of 10 CFR 100. The limit is

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currently the maximum number of failed fuel rods (20%) calculated in theFSAR.

3. The maximum clad temperature calculated to occur at the core hot spotmust not exceed 2700*F.

The transient meets all acceptance criteria, including the FSAR limits for thenumber of fuel rods calculated to experience DN8, and is therefore acceptablefor operation with the thimble plugs removed.

3.2 Rod Ejection Analysis

Rod ejection analyses were performed by the licensee for Prairie Island Unit 1Cycle 10 and Unit 2 Cycle 9 and are discussed in References 6 and 7,respectively, for the core with the thimble plugs in. The analyses assumed ahot channel flow of 86% total average channel flow whereas the actual hotchannel flow as determined from a COBRA analysis by the licensee wasapproximately 90% average. For the case where the thimble plugs are removed,the bypass flow is increased by 1.5% which is calculated to result in a hotchannel flow of approximately 89%. This is bounded by the more conservative

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value of 86% assumed in the original analyses and therefore operations with '

the thimble plugs removed is supported by the original analyses.

3.3 LOCA-ECCS Analysis

The licensee is currently performing LOCA-ECCS analyses for the two types of '

fuel (Exxon TOPROD and Westinghouse OFA) for operation with the thimble plugsremoved (6% bypass flow). The analyses are scheduled for completion beforethe starup of Prairie Island Unit 1, Cycle 11. The thimble plugs will remainin the core until that time.

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3.4 ATWS Analysis

The ATWS issue for Prairie Island is currently under review by NRC. Removalof the thimble plugs has no impact on the analyses which have been requiredat this time.

4.0 Mechanical and Structural Analysis

The effect of removing the thimble plugs increases the total vessel flow butdecreases the active core flow. The fuel assembly mechanical design and thereactor vessel internal design are based on upper limit bounding RCS flows.For the Prairie Island plants, the licensee stated (Reference 9) that bothExxon and Westinghouse have calculated the liftoff flow for their fuel to be206,000 gpm or greater (References 9 and 10). The licensee stated (Reference8) that the current best estimate RCS flow rates for Prairie Island Units 1and 2 are 196,883 and 195,331 gpm respectively. Tt.ase flows will increase0.6% due to removal of the thimble plugs and approximately 0.6% due to thelower resistance of the Westinghouse fuel (based on a total Westinghouse core).Also a 2.0% measurement and instrument uncertainty must be applied. The revisedflow rates are then 203,237 and 201,635 gpm for Prairie Island Units 1 and 2respectively. These flow rates are bounded by the calculated liftoff flowfor the fuel of 206,000 gpm. Additional information was presented by thelicensee in Reference 8 which indicated that the removal of the thimble plugswill drop the differential pressure across the core and thus reduce the liftoff

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forces acting on the fuel. This also decreases the possiblity of fuel liftoff.

The original hot functional tests were run at 140% rated flow (249,200 gpm)

and showed no adverse affects on the reactor vessel mechanical vibrationdesign. Also, a review of the Westinghouse design analysis shows that removalof the thimble plugs will not have any effect. '

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In consideration of the above information the staff finds the mechanical andstructural design acceptable for the removal of the thimble plugs.

I 5.0 Application to Westinghouse Fuel '

fThe evaluation discussed for the effect of removal of the RCC thimble plugs,

i has been for the current Prairie Island Unit 1 Cycle 10 and Unit 2 Cycle 9! which has Exxon TOPROD fuel. However, in the next cycles of operation of

Prairie Island Units 1 and 2, Westinghouse Improved Optimized Fuel Assemblies(OFA) will be inserted. As shown in Table 1, there is less bypass flow for9

OFA fuel in comparison with TOPROD fuel when the RCC thimble plugs are removed..

This is in the conservative direction for lif t-off forces. The licensee has statedthat the Westinghouse LOCA/ECCS analysis will be based on the same input parameters,

I as for Exxon TOPROD fuel. A reevaluation must be made by the licensee beforeuse of Westinghouse OFA fuel in future reloads.

| 6.0 Core Physicsi1

The following reasons were presented by the licensee to show that the core,

; physics calculations are not significantly affected by the removal of the RCC| guide thimble plugs.

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(a) The thimble plugs do not extend into the active fuel region.

(b) The actual core active flow will decrease by approximately 1% whichj translates to approximately a 0.5'F change in core moderator fuelI temperature. This is considered insignificant for reactivity.|

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(c) The actual core bypass flow will increase by approximately 2% which;

translates to less than a 2*F change in moderator temperature. .This is;

; considered insignificant in terms of reactivity.!.

The reasons support the conclusion that there will be no significant change,

in core physics due to thimble plug removal.!

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7.0 SUMMARY AND CONCLUSION

Information has been presented to show that the effect of removing the RCC -

guide thimble plugs for Prairie Island Units 1 and 2 using TOPROD foel willresult in an increase in the core bypass flow (decrease in active core flow)and an increase in total RCS flow. The licensee performed safety analysesafter a review to examine where the present analyses bound operation with thethimble plugs removed. Those accident and transient analyses not bounded wererecalculated. From this information it was concluded that operation ofPrairie Island Units 1 and 2 with the RCC thimble plugs removed is acceptableas it will not result in any adverse reduction in safety margin.

A LOCA-ECCS reevaluation must be made by the licensee before insertion of theWestinghouse OFA fuel in forthcoming Prairie Island reloads.

Principal Contributor:H. Balukjian

Date: October 18, 1985

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TABLE 1

Bypass Flow Calculations-

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EXXON TOPROD

j Best Estimate Bypass Flow" (% total desian)

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j Plugs In 2.26! Plugs Out 3.64

Increase 1.02

Including Uncertainties;

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! Plugs In 2.50i

] Plugs Out 3.82

! Increase 1.32 L

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WESTINGHOUSE OPTIMIZED '

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Including Uncertainties1

Plugs In 2.39 |

1 Plugs Out 3.59Increase 1.20

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| * Defined here as the flow through the non-rodded| (currently plugged) guide tubes.|

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TABLE 2 '

Prairie Island Thermal Hydraulic Reference Conditions.

Reactor Conditons Nominal;

Rated Core Power (MWt) 1650 (100%)

Total Reactor Flow Rate (M1b/hr) 68.2

Active Core Flow Rate (M1b/hr) 64.1Core Coolant Inlet Temperature (*F) 530.5

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Core Pressure (psia) 2250.0

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Power Olstribution

Total Peaking (F ) 2.32*q

Enthalpy Rise (Fg) 1.65*Axial 1.365 :Engineering Factor 1.03

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* Current Technical Specification Limit are FQ = 2.32 and FAH = 1.55,*

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TABLE 3

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Slow Rod Withdrawal.

. Transient and Thermal Margin ResultsI

Pluas In Pluas Out

PONBR 1.413 1.363NBo 0.034 0.034g

MON 8R 1.365 1.3178

These results bound PI 1 Cycle 10 and PI 2 Cycle 9.

MONBRNB = Non bowed MONBR

MON 8Rg = Bowed MON 8R

08 = Fractional DNBR reduction due to rod bow

MON 8R8 = MON 8Ryg (1 - O)g

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TABLE 4

Parameter Values Used in Full Power Transient Analysis.

Analysis

InputValue

Core

Total Core Heat Output, Mw (102%) 1,683.0Heat Generated in Fuel, % 97.4System Pressure, psia 2,220*

Hot Channel FactorsTTotal Peaking Factor, F 2.32g

Enthalpy Rise Factor, FAH 6Total Coolant Flow, Ib/hr 68.20 X 10

6Effective Core Flow, 1b/hr 64.11 x 10 .,

Reactor Inlet Temperature, "F 543.5Steam Generators

6Calculated Total Steam Flow, Ib/hr 7.26 x 10Steam Temperature, 'F 510.8Feedwater Temperature, 'F 427.3

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* Locked Rotor is initiated from 2280 psia** 94% Total Coolant Flow

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! TABLE 5

:| Summary of Plant Transient Analysis Results

i MDNBR % Failed FuelTransient Plugs In Plugs Out Pluas In Pluas Out

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j Slow Rod Withdrawal 1.413 1.363 - -

I Locked Rotor <1.3 <1.3 16.46 18.20:I

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These results bound PI 1 Cycle 10 and PI 2 Cycle 9.4

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8.0 REFERENCES

1. Letter from D. Musolf, Northern States Power, to Director, NRR, NRC,dated April 22. 1985. '

2. " Prairie Island Nuclear Generating Plant Units 1 and 2, Updated SafetyAnalysis Deport," Docket Numbers 50-282, 50-306.

3. XN-NF-80-61, " Prairie Island Nuclear Plants TOPROD Safety AnalysisReport," Revision 1, March 1981.

4. XN-75-32(P)(A), Supplements 1, 2, 3, and 4, " Computational Procedure forEvaluating Fuel Rod Bowing," dated October 1983.

5. Deleted.

6. " Prairie Island Unit 1 Cycle 10 Final Reload Design Report (RSE)",NSPNAD-8411P, October 1984.

7. " Prairie Island Unit 2 Cycle 9 Final Reload Design Report (RSE)",NSPNAD-8404P, Revision 2, May 1984.

8. Letter from D. Musolf, Northern States Power, to Director, NRR, NRC,1

dated August 30, 1985.

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i 9. Telecopy NLG:020:85, N. Garner (Exxon) to T. M. Parker, Northern States

Power Company " Evaluation of Increased Coolant Flow at Prairie Island",August 26, 1985.

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10. Letter 855-G-023 R. T. Meyer (Westinghouse) to T. M. Parker, Northern

I States Power Company " Release of Westinghouse Proprietary Information,i

to NRC", August 22, 1985.,

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