ML20116C926

Safety Evaluation for Rod Cluster Control Guide Thimble Plug Removal

ML20116C926
Person / Time
Site:	Prairie Island
Issue date:	04/30/1985
From:	Kapitz J NORTHERN STATES POWER CO.
To:
Shared Package
ML20116C911	List: ML20116C907 ML20116C926
References
NSPNAD-8412, TAC-57544, TAC-57545, NUDOCS 8504290215
Download: ML20116C926 (34)

Text

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. PRAIRIE ISLAND UNITS 1 AND 2 SAFETY EVALUATION FOR RCC GUIDE THlMBLE PLUG REMOVAL NSPNAD-8412 April 1985 Prepared by / - v. - d' f. 4

,, - Date

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Reviewed by -0">7 Date I p .- .

Approved by 1 ' ' ' I' / ~C '

A is . Date '/ /s 8504290215 850422 PDR ADOCK 05000282 P PDR Page 1 of 34

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Proprietary Data Clause This document is the property of Northern States Power Company (NSP) and contains proprietary information developed by NSP. Any reproduction or copying of such information requires the express written consent of NSP.

Disclosure of such infomation to parties outside of NSP requires the express written consent of NSP.

Page 2 of 34

LEGAL NOTICE This report was prepared by or on behalf of Northern States Power Company (NSP). Neither NSP, nor any person acting on behalf of NSP:

a. Makes any warranty or representation, express or implied, with respect to the accuracy, completeness, usefulness, or use of any information, apparatus, method or process disclosed or contained in this report, or that the use of any such information, apparatus, method, or process may not infringe privately owned rights; or

b. Assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method, or process disclosed in the report.

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EXECUTIVE

SUMMARY

This report summarizes'the calculations, completed by NSPNAD, to evaluate the effect of removing the RCC guide thimble plugs on Prairie Island Units 1 and 2.

There are 90 RCC guide thimble plugs currently in each core. The plugs serve to limit the bypass flow through the RCC guide thimbles in fuel assemblies which do not contain either control rods or source assemblies. -

Removal of the RCC guide thimble plugs at Prairie Island will decrease the active core flow while increasing the total RCS flow.

The effect on the current safety analyses was evaluated by either redoing the ca'1culation or by showing that the present analysis bounds operation with the thimble plugs removed.

It is concluded that operation of the two Prairie Island units with the RCC guide thimble plugs removed will not result in any reduction in safety margin and hence does not constitute an unreviewed safety question per 10CFR.50.59.

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.c TABLE OF CONTENTS Page

~.

1.0 INTRODUCTION

8 2.0 FLOW CALCULATIONS 12 2.1 Core Bypass . 12 2.2 RCS 12 3.0 THERMAL HYDRAULIC DESIGN ANALYSIS 14 3.1 Design Criteria 14 3.2 Core Hydraulic Compatibility 14 3.3 Thermal Margin 15 3.4 Effect of Fuel Rod Bow on Thermal Hydraulic Performance 15 3.4.1 Rod Bow as Applied to DNBR Analysis 15 3.4.2 Effect of Rod Bow on LOCA Limits 15 3.5 Fuel Temperature Analysis 15 3.6 Safety Limit Curves 16 4.0 ACCIDENT AND TRANSIENT ANALYSIS 19 4.1 Plant Transient Analysis 19 4.1.1 Input Parameters 19

. 4.1.2 Transient Analysis Results 20 4.1.2.1 Slow Rod Withdrawal 20 4.1.2.2 Locked Rotor 20 4.1.3 Conclusions 21 4.2 Rod Ejection Analysis 21 4.3 LOCA - ECCS Analysis 22 4.4 ATWS Analysis 22 5.0 MECHANICAL AND STRUCTURAL ANALYSIS 25 5.1 Fuel Mechanical Design and Fuel Liftoff Forces Analysis 25 5.2 Reactor Vessel and Internals Mechanical & Vibration Analysis 25 6.0 APPLICATION TO WESTINGHOUSE FUEL 26 7.0 CORE PHYSICS 27 8.0

SUMMARY

AND CONCLUSIONS 28

9.0 REFERENCES

29 APPENDIX A - CORE BYPASS FLOW CALCULATIONS 30 Page 5 of 34

LIST OF TA8LES

'. Pace 2.1 Bypass Flow Calculations 13

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3.1 Prairie Island Thermal Hydraulic Reference Conditions 17 3.2 Slow Rod Withdrawal Transient and Thermal Margin Results 18 4.1 Summary of Plant Transient Analysis Results 23 4.2 Parameter Values Used in Full Power Transient Analysis 24 i

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LIST OF FIGURES i

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1.1 Thimble Plug Assembly 10 1.2 Fuel Assembly and Control Cluster Cross Section ' 11

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A.1 Exxon Guide Tube Dimensions 33 A.2 Westinghouse Guide Tube Dimensions 34 Page 7 of 34 l

1.0 INTRODUCTION

This report summarizes the calculations to evaluate tiie effect of removing the RCC guide thimble plugs on Prairie Island Units 1 and 2.

There are 90 RCC guide thimble plugs (Figure 1.1) <;urrently in each core.

The plugs serve to limit the bypass flow through the ACC guide' thimbles in fuel assemblies (Figure 1.2) which do not contain either contrgl rods or source assemblies.

Northern States Power is seeking removal of the thimble plugging inserts from the core at Prairie Island Units 1 and 2 for the following reasons.

1. The new Westinghouse Optimized Fuel Assemblies (OFA) will not accept the present thimble plugs, due to the smaller diameter guide tubes, so that new ones would have to be purchased. The cost for a new replacement set would be in excess of $600,000 for both units. Removal of the plugs will eliminate this cost.

2. Presently 8 to 12 extra hours each refueling are spent changing thimble plugs from old assoaslies to new assemblies. Removal of the plugs would result in a savings of $40,000 to $60,000 per outage.

3. Since the thimble plugs must ultimately be disposed of as intermediate level waste, their elimination will lower waste generation and result in disposal

cost savings. ,

i

4. The potential presently exists for a time loss during refueling due to bent or damaged thimble plugs, resulting in a time loss of a few hours. This could cost up to 515,000 if it occurs (approximately 12 occurrences to date l on critical path).

5. There would be a radiation dose reduction by removing the plugs due to I

reduced fuel shuffle time, elimination of plug repair when they are bent or damaged, and the elimination of the ultimate disposition and handling exposure.

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6. By removing the inserts from the core the potential for loose parts generation is reduced. Some instance of cracked, plug springs have been reported by other utilities, and loose plug fingars are a possibility; they have nuts lock-wired onto the plug and tack-welded. Removal of the plugs will' result in an overall reduction of complexity of the core and internals assembly.

When the thimble plugs are removed, the, hydraulic resistance through the bypass region of the core decreases thereby increasing the bypass flow and decreasing the active core flow. A decrease in active core flow will decrease the fuel thermal margins, i.e. MON 8R, fuel melt and PCT margins. The decrease in active core flow will be partially offset by an increase in total flow due to a decrease in the.overall core resistance. Increasing the total core flow will reduce the mechanical design flow margin. This report will evaluate the effect of these flow changes on the thermal hydraulic d Mign analysis, the accident and transient analysis, and the mechanical and structural analysis, i

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!! I Figure 1.2 Fuel Assembly and Control Cluster Cross Section Page 11 of 34

2.0 FLOW CALCULATIONS 2.1 Core Bypass Flow .

The current design bypass core flow fraction is given in Reference 1 Table 3.2-3 as 4.55. Approximately 0.3% is directed through 24 holes 15' apart drilled at different angles in the internals support ledge to cool the head,1% flows through the clearance between the discharge nozzle in

,, the upper core barrel and the reactor vessel, 3.2% flows through the guide thimbles in thi fuel assembifes. The decrease in active core flow fraction due to removal of the guide thimble plugs is calculated to be u- less than 1.55. This calculation is described in Appendix A. This calculation is dependent on fuel type and is cycle independent. Results for Exxon T0pR00 and Westinghouse OFA are presented in Table 2.1. A new design bypass flow fraction of 6.0% will therefore bound all reloads with any combination of these two fuel types.

2.2 RCS Flow i

Assuming the core bypass flow increases to 6.0%, when the thimble plugs are removed, the RCS flow is conservatively calculated to increase a maximum of 0.6%

Page 12 of 34 1

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TABLE 2.1 ,

Bypass Flow Ca'1culations EXXON TOPROD Best Estimate Bypass Flow" (% total design)

Plugs In 2.62 Plugs Out 3.64 Increase 1.02 i

Including Uncertainties Plugs In 2.50 Plugs Out 3.82 Increase 1.32 WESTINGHOUSE OPTIMIZED Including Uncertainties Plugs In 2.39 Plugs Out 3.59 Increase 1.20

! Defined here as the flow through the non-rodded l

(currently plugged) guide tubes.

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3.0 THERMAL HYORAULIC DESIGN ANALYSIS This section provides results of the thermal 5)draulf'idesign analyses for Prairie Island 1 Cycle 10 and Prairie Islano 2 Cycle 9 with the RCC guide thimble plugs removed.

3.1 Desion Criteria The thermal and hydraulic design performance requirements for ENC reload fuel design are as follows.

1) The minimum departure for nucleate boiling ratio (MONBR) will be 1.3

. at overpower using the W-3 correlation with corrections for non-uniform axial heating, cold wall effects, and a reduction in MONBR due to fuel rod bowing.

2) The fuel must be thermally and hydraulically compatible with the existing fuel and the reactor core throughout the life cycle of the fuel.

3) The maximum fuel temperature at design overpower shall not exceed the fuel melting temperature.

4) The cladding upper temperature limits shall not exceed:

Inner surface temoerature 850 *F ,

Outer surface temperature 675 'F Average voltmetric temperature 750 'F 3.2 Core Hydraulic Comcatibility The hydraulic compatibility of the Prairie Island Unit 1 Cycle 10 and Unit 2 Cycle 9 reload fuel is discussed in Reference 3. The hycraulic comcatibility of the fuel is not affected by removal of the RCC guide thimble plugs, cue to the similarities of the fuel assemblies.

e Page 14 of 34

a 3.3 Thermal Margin The most limiting operational transient for Prathie Island Unit 1 Cycle 10 and Unit 2 Cycle 9 is the slow rod withdrawal. For this event the MDNBR is calculated to be 1.363 which bounds both PI 1 Cycle 10 and PI 2 Cycle 9.

Table 3.1 provides reference conditions for this analysis. Details of the plant transient analysis are given in Section 4.0.

3.4 Effect of Fuel Rod Bow on Thermal Hydraulic Performance 3.4.1 Rod Bow as Applied to DNBR Analysis The calculation of the DNBR reduction as a result of rod bow is described in Reference 4. This procedure for evaluating fuel rod bowing will not be affected by removal of the thimble plugs, hence 68 , the fractional reduction in MONBR due to fuel rod bowing, will not change. Table 3.2 provides a comparison of key results and rod bow penalties from the analysis of the slow rod withdrawal event, in Section 4.0, for PI 1 Cycle 10 and PI 2 Cycle 9. The bowed and unbowed results are well above the allowable 1.3 limit for the case with and without thimble plugs.

3.4.2 Rod Bow as Applied to LOCA Limits The effect of rod bow on LOCA limits for Prairie Island is given in Reference 3. This effect is independent of bypass flow .

fraction hence there will be no change due to removal of the thimole plugs.

3.5 Fuel Temperature Analysis The fuel temperature analyses for Prairie Island 1 Cycle 10 are given in Reference 3. Increasing the core bypass flow to 6.0% due to thimble plug removal will not significantly impact the analyzed fuel temperature. The active core moderator temperature will increase less than 0.5 *F which will increase the fuel temperature by approximately the same amount. This is less than a 0.02% increase in peak centerline temperature.

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3.6 Safety Limit Curves The safety limit,curjv s. for Prairie Island, giveit in Technical Specifications Section 2.1, along with the associated overpower and overtemperature AT setpoints were calculated using FAH=1,55 and FQ=2.51 (Reference 5). Current Technical Specification Ifnits are FAH=1.55 and RM2.32. The margin between the analyzed and actual FQ IImit is more than sufficient to account for removal of the thimble plugs.

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TABLE 3.1 -

Prairie Island Thermal Hydraulic Reference Conditions Reactor Conditions- Nominal Rated Core Power (MWt) 1650 (100%)

Total Reactor Flow Rate (M1b/hr) 68.2 Active Core Flow Rate (M1b/hr) 64.1 Core Coolant Inlet Temperature (*F) 530.5 Core Pressure (psia) 2250.0 Power Distribution Total Peaking (Fg ) 2.32*

Enthalpy Rise (FAH) 1.65*

Axial 1.365 Engineering Factor 1.03 Current Technical Specification Limits are FQ = 2.32 and FAH = 1.55 Page 17 of 34

TABLE 3.2 '.

Slow Rod Withdrawal Transient and Thermal Margin Rer Its plues fn pluos out MONBR N8 1.413 1.363 og 0.034 0.034 MONSRg 1.365 1.317 These results bound PI 1 Cycle 10 and PI 2 Cycle 9.

MONBRNB = Non bowed MON 8R MON 8Rg = Bowed MON 8R 8 = Fractional DN8R reduction due to rod bow B

. MON 8Rg

= MON 8Rh3 (1 - og )

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4.0 ACCIDENT AND TRANSIENT ANALYSIS 4.1 Plant Transient Analysis .

This documents the analysis of plant operational transients for Prairie Island Unit 1 Cycle 10 and Unit 2 Cycle 9 with the RCC guide thimbles removed. The analyses without the plugs removed are presented in References 6 and 7. These analyses were -done at an FAH of 1.65.

Current Technical Specifications limit FAH to 1.55.

Increasing the core bypass fraction due to thimble plug removal will not significantly affect the plant transient system analysis. This is because the systems analysis is based on core average conditions which are not significantly changed, i.e. less than 0.5 'F in moderator temperature.

Therefore the systems analyses were not rerun and the analysis in References 6 and 7 remain valid.

The hot channel analyses will be affected by removal of the thimble plugs.

Small changes in the active core flow can significantly change the hot channel response.

Removal of the RCC guide thimble plugs will not affect the relative significance of the transients. Therefore the transients which were found to be limiting in the previous analyses (References 6 and 7) will remain limiting for the plugs removed analyses.

4.1.1 Input Parameters Table 4.2 summarizes the thermal hydraulic parameters for full power operation.. For full power operation, an axial peaking factor, FZ, of 1.365 located at X/L = 0.7, an enthalpy-rise factor, FAH, of 1.65, and a tctal peaking factor, FQ, of 2.32 are assumed. References 6 and 7 give a more detailed listing of the input parameters assumed.

Page 19 of 34

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l 4.1.2 Transient Analysis Results 4.1.2.1 Slow Rod Withdrawal ',

The transient response of the NSSS for this case is shown in Reference 7 Figures 8.8 through 8.11. The DNS ratio drops from its initial value of 1.792 to a minimum value of 1.363 at 38.6 seconds. These results e

boundbothPIICycle10andPI2 Cycle 9 operation.

The acceptance criteria are that the minimum DNBR be not less than 1.3 and that the maximum reactor coolant and main steam system pressure not exceed 110% of their design values. This transient meets all acceptance criteria.

4.1.2.2 Locked Rotor The transient response of the NSSS for this case is shown in Reference 6 Figures 8.18 through 8.23. The number of fuel rods statistically calculated to experience DN8 for this transient is 18.2% These results bound both PI 1 Cycle 10 and PI 2 Cycle 9 operation.

The acceptance criteria for the locked rotor analysis .

are as follows.

1. The maximum reactor coolant and main steam system pressures must not exceed 1125 of the design values.

Page 20 of 34

2. The number of fuel rods calculated to experience a DNBR of less than 1.3 should not exceed the number which are' required to fa.il in order th'at the doses due to released activity will exceed the limits of 10CFR 100. The limit is currently the maximum ,

number of failed fuel rods calculated in the FSAR.

3. The maximum clad temperature calculated to occur at the core hot spot must not exceed 2700 'F.

This transient meets all acceptance criteria.

4.1.3 Conclusions The analysis showed the calculated MON 8R, with the thimble plugs removed, for the limiting Class II and III transient, the slow rod withdrawal, is above the acceptable minimum DNBR of 1.3. The limiting Class IV transient, the pump seizure, showed a calculated MONBR of < 1.3. The number of fuel rods which potentially experience DNS in the transient is calculated to be less than the acceptable limit of 20%. Table 4.1 summarizes the hot channel results. Therefore all plant operational transients meet the acceptance criteria with the thimble plugs removed.

4.2 Rod Ejection Analysis The Rod Ejection analyses for PI 1 Cycle 10 and PI 2 Cycle 9 are discussed in References 6 and 7, respectively. These analyses assume a hot channel flow of 86% total average channel flow. The actual hot channel flow is approxi-sately 90% average. This number is determined from the COBRA analysis.

Therefore, increasing the bypass flow by 1.5% total flow will reduce the hot channel flow to approximately 89% which is bounded by that assumed in the analyses. Therefore, the original analyses remain valid.

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4.3 LOCA-ECCS Analysis The LOCA-ECCS analysis for Prairie. Island is given in Reference 8. This analysis assumes Technical Specifications design flow, i.e. 178,000 gpa, 4.5% bypass flow, and 55 steam generator tube plugging. The actual RCS flow is much greater than the design flow, approximately 110% design flow. This excess flow will offset the effect of removing the thimble plugs.

The actual core flow with 55 steam" generator tube plugging and 6.0% bypass (thimble plugs removed) is conservatively calculated to be 69.0 and 67.4 M1be/hr for PI 2 and PI 1 respectively. The current analysis assumes a core flow of 65.5 M1bs/hr. Therefore, the current analysis bounds the

, actual plant operation with the thimble plugs removed and no reanalysis is necessary. This argument remains valid unless a design change is made at the plant which significantly reduces actual RCS flow, excluding steam generator tube plugging. This type of change is not anticipated. A bypass flow fraction of 6.0% will be , included in all future ECCS-LOCA analyses.

4.4 ATWS Analysis The ATWS analysis for Prairie Island is given in reference 9. This analysis remains valid based on the same arguments discussed in Section 4.3. The actual core flow after removal of the thimble plugs bounds that assumed in the analysis.

page 22 of 34

TABLE 4.1 '.

. Summary of Plant Transient Analysis Results Transient MDNBR  % Failed Fuel plugs in plugs out plugs in olugs out Slow Rod Withdrawal 1.413 1.363 - -

Locked Rotor < 1.3 < 1.3 16.46 18.20 These results bound PI 1 Cycle 10 and PI 2 Cycle 9.

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TABLE 4.2 .

Parameter Values Used in Full Power Transient Analysis Analysis Input Value Core Total Core Heat Output, Mw (102%) 1,683.0 Heat Generated in Fuel, % 97.4

. System Pressure, psia - 2,220*

Hot Channel Factors Total Peaking Factor, F0 2.32 Enthalpy Rise Factor, F 1.65 AH Total Coolant Flow, Ib/hr 68.20 x 10 6 Effective Core Flow, Ib/hr 64.11 x 100 **

Reactor Inlet Temperature, *F 534.5 Steam Generators Calculated Total Steam Flow, Ib/hr 7.26 x 10 6 Steam Temperature, 'F 510.8 Feedwat.ar Temperature, 'F 427.3 Locked Rotor is initiated from 2280 psia 94% Total Coolant Flow 4

I Page 24 of 34 9

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5.0 MECHANICAL AND STRUCTURAL ANALYSIS The fuel assembly mechanical design and the vessel intarnals structural design assume upper limit bounding flows. Removal of the thimble plugs decreases the active core flow and increases the total vessel flow due to the decreased bypass flow resistance. .

5.1 Fuel Mechanical Design and Fuel Liftoff Forces Analysis Exxon has verified the mechanical design of their fuel up to 201,000 gpm and the liftoff forces up to 205,800 gpm.

Westinghouse has verified the design of the fuel to to 200,600 gpm for the Prairie Island plants.

It is conservatively calculated that the actual RCS flow will increase to 198,700 and 200,400 gpm for Prairie Island Units 1.and 2 respectively due to thimble plug removal. These numbers include measurement uncertainty and are based on zero SG tube plugging. Plugging steam generator tubes increases system flow resistance and hence decreases total RCS flow. It is conservatively calculated that the increase in RCS flow due to removal of the thimble plugs will be offset by the decrease in core flow associated with plugging 0.9% of the tubes. Current plugging levels are approximately 1.0% and 2.0% for PI 1 and 2 respectively. Therefore, the design analysis will remain bounding after the thimble plugs are removed.

5.2 Reactor Vessel and Internals Mechanical and Vibration Analysis A review of the Westinghouse reactor vessel and internals mechanical and vibration design analysis shows that thimble plugs are not a consideration.

I In addition, original hot functional tests at Prairie Island were run at 140% rated flow (249,200 gpm) and showed no adverse effectc. Therefore the design analysis will remain bounding after the thimble plugs are removec.

l Page 25 of 34 9

6.0 APPLICATION TO WESTINGHOUSE FUEL ,

This evaluation was performed for Prairie Island Unit) Cycle 10 and Unit 2 Cycle 9 with Exxon TOPR00 fuel.

The Westinghouse Improved Optimized fuel is bounded by the Exxon TOPROD with respcct to bypass flow, i.e. removal of the thimble plugs will'eause less of a bypass flow increase in the Westinghouse assemblies than in the Exxon assemblies (see Table 2.1).

The Westinghouse Improved Optimized fuel is also bounded by the Exxon TOPROD with respect to ONBR margin. This is due to use of the new Westinghouse WR8-1 correlation.

The Westinghouse LOCA/ECCS analysis is based on the same input parameters as Exxon currently uses.

Therefore, all calculations and qualitative arguments presented in this report bound or apply to the Westinghouse Improved Optimized fuel assemblies.

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7.0 CORE PHYSICS The core physics calculations will not be significant{y affected by the removal of the RCC guide thimble plugs for the f'ollowing reasons; The thimble plugs do not extend into the active fuel region.

Actual core active flow will decrease by less than 1%. This translates to less than a 0.5 *F change in core moderator and fuel temperature.

This is insignificant in terms of reactivity.

Actual core bypass flow will increase by approximately 2%. This translates '

to less than a 2 *F change in bypass moderator temperature. This is insignificant in terms of reactivity.

Therefore there will be no significant reactivity changes due to thimble plug

. removal.

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8.0 Supt %RY AND CONCLUSIONS Removal of the RCC guide thimble plugs at Prairie Islitnd will decrease the active core flow while increasing the to'tal RCS flow.

The effect on the current safety analyses was evaluated by either redoing the calculation or by showing that the present analysis bounds ope' ration with the thimble plugs removed.

It is concluded that operation of the two Prairie Island units with the RCC guide thimble plugs removed will not result in any reduction in safety margin and hence does not constitute an unreviewed safety question per 10CFR 50.59.

This evaluation was performed for Prairie Island Unit 1 Cycle 10 and Unit 2

Cycle 9 with ENC TOPROD fuel. Sufficient conservatisms have been incl'uded in these evaluations so that they generically bound all cycles with Exxon TOPR00 and Westinghouse Improved Optimized fuel assemblies.

Page 23 of 34

- - - -n-, n- _ . _ . _ _ . - - - -

9.0 References

1. " Prairie Island Nuclear Generating ' Plant Units 1 & 2, Updated Safety Analysis Report", Docket Numbers 50-282, 50-306.

j 2. " Reload Safety Evaluation Methods for Application to PI U'its" n NSPNAD-8102P, December 1982.

s

3. XN-NF-80-61, " Prairie Island Nuclear Plants TOPROD Safety Analysis Report",

Revision 1, March 1981.

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

5. WCAP 8090, " Fuel Densification Prairie Island Nuclear Generating Plant Unit No. 1," March 1973.

6. " Prairie Island Units 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 Rev.2, May 1984.

8. XN-NF-83-38, " Prairie Island Units 1 and 2 Limiting Break LOCA/ECCS Analysis using EXEM/PWR" May 1983. -

9. WCAP-8330, " Westinghouse Anticipated Transients Without Trip Analysis.

. 10. " Flow of Fluids Through Valves, Fittings, and Pipe," Crane, Technical Paper No. 410.

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APPENDIX A Core Bypass Flow Calculations Exxon Fuel '.

The increase in bypass flow is calculated as follows; Initial assumptions; Design flow = 68.2 E+6 lbs/hr Bypass flow fraction = 4.55 No heating in the thimble tubes Nominal inlet conditions, 2250 psia, 534.5 'F Constant pressure drop across all channels Guide tube dimensions are shown on Figure A-1 AP = KpV8/2g p = density = 48.07 lbs/ft' gg = conversion factor = 4.17E+8 lbe ft/lbf hr2 V = coolant speed = W/pA ft/hr

. A = cross secttonal area = 2.0189 ft*

W = coolant mass flow = lbs/hr K =

loss coefficient = KLOSS

• EFRICTION K

FRICTION

= f/ De)

L = channel length = 152.0 in 0, = channel equivalent diameter = 0.507 in i

f = friction factor = 0.184 Re -0.2 Re = Reynolds number = 0,Vp/u .

y = coolant viscosity ='O.24 lbs/ft hr KLOSS (plugs in) = form loss coefficient = 148.5 KLOSS (plugs out) = fem loss coefficient = 73.2 The form loss coefficients are calculated based on Reference 10.

The inlet form loss coefficient assumes five parallel inlet flow', paths, i.e. the screw cap and the four reliaf holes.

AP (plugs in) = pressure drop not including head loss

= 20.88 psi  !

AP (plugs in) is taken from the HFP CCBRA IIIC/MIT model (Reference 2) which is developed as part of the normal reload process page 30 of 34

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Solving for W (plugs in) gives; W (plugs in) = 1.789E+6 lbm/hr

= 2.62% design flow- 1 It is assumed that; AP (plugs out) = AP(plugsin)*[W8 ACTIVE (plugs out) / W8 ACTIVE (PI"95I")3

=

20.88 * (68.2E+6 [1-(.045 .0262)] - W)*

/ [(68.2 E+6) (1 .045)]*

where WACTIVE = Active core flow = lbm/hr Solving for W (plugs out) gives; W (plugs out) = 2.481 E+6 lbm/hr

= 3.64% design flow In order to assure a conservative calculation, a 10% uncertainty is included on the K #*Ct""

LOSS A +10% uncertainty is assumed on KLOSS (plugs in) and a -10%

on KLOSS (plugs out). This maximizes the increase in bypass flow between the two cases.

Solving gives; W (plugs in) = 2.50%

W (plugs out) = 3.82% .

A increase = 1.32k So that WBYPASS (plugs ut) = 4.5 + 1.32 = 5.82 The new design bypass flow is therefore assumed te be 6.0%

Page 31 of 34

Westinghouse Optimized Fuel The same calculation was performed for the Westinghouse Opt.imized fuel. The guide tube dimensions are shown on Figure A-2. Solving gives; W(plugsin) = 2.39%

W (plugs out) = 3.59%

A increase = 1.20%

So that the We stinghouse fuel is bounded by the Exxon fuel with respect to bypass flow.

If we assume that the rodded guide tubes have the same resistance as the plugged guide tubes then the total guide tube bypass flow for a Westinghouse core can be calculated as; Wgypg33 (guide tubes) =

(2.39) (121/90)

= 3.21%

This agrees with the value calculated by Westinghouse of 3.2% (see page 12).

Page 32 of 34 9

4

FIGURE A.1 EXXON H(in) D(in) Khss**

.,________ 155.125 0.471 53 -

1_ I

_______. 148.77 0.507 23 Q ___. .________ 26.02 0.094 1269 24' Y  ; /_(([~(([ 25.52 0.447 1 O

-e-1.280 0.3125 2 .

I /

', t

/

e '

__________ l.122 0.04 24970 i ,

l /

!  %,h _________.. 0.872 100* 8870 t /

M

___..______ 0 0.125 135 1

Note: Not to scale

• • Note: All K factors relative to 0.507

. Page 33 of 34

.,. e s

FIGURE A.2-WESTINGBOUSE H(in) D(in) Kg33 **

': J.-------- 154.360 '0.454 50

= -------- 148.0 0.490 21 Q ----. -------- 25.54,25.36 0.091 1261

---

15'

---

. 24.865 0.4445 1 O

/-

..-----.---. 1.550 22156

~

-.......... 0 420 0.040, 118' 8688

---

. 0.170 0.098, 118* 210 U ----------. 0 0.1875 23 Note: Not to scale

• • Note: All I Factors relative to 0.490 i

Page 34 of 34

.. . _ _ . _ _ . _ _ .