Text

Nonproprietary Replacement Pages to HI-91738,consisting of Section 5.0, Thermal-Hydraulic Analysis
	ML20205L054
Person / Time
Site:	Nine Mile Point
Issue date:	04/01/1999
From:	; HOLTEC INTERNATIONAL
To:	;
Shared Package
ML20205L053	List: ML20205L050; ML20205L054;
References
	HI-91738-ERR, NUDOCS 9904140115
	Download: ML20205L054 (49)
	v • d • e

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ATTACHMENT A Section 5.0 Thermal-Hydraulic Analysis l

9904140115 990401 7 PDR ADOCK 05000220 -

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l TABLE OF CONTENTS SECTION PAGE 1.0 INTRODUCTI ON. . .. . . . . . ... .... . .. ... . ... .. . ... . . ... . . . ..........................................................1-1 2.0 MODULE LAYOUT FOR INCREASED STORAGE..... ...................... .. .... . .. ...... 2-1 2.1 Heavy Load Considerations for the Proposed l

R eracking Operatio n .... . . ... .. .. .. . . . . .. . . . . . .. . . . .. . . . . . . . . . .. . . . . . . . . . .. . . . . . . . . . . . . . . . .. . . .. . . . . 2-2 2.2 Intennediate Configuration Considerations . .......... ......... ...... ..................... ... 2-7 3.0 RACK FABRICATION AND APPLICABLE CODES . .......................... . ... .. . ... . 3-1 3.1 D esi gn Obj ective ... . . . . . . . ... .. . . . . . .. . . .. . . . . . . . .. . . .. ..... . . .. . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . .. . . .. . 3 - 1 3.2 Anatomy o f the Rack Mod ule . ... .......... . . . .... . . ............. .... ... .... ..... . .. ........ . . . 3-2 3.3 Materials o f Construction . ... .......... ... .. . ... ... ... ............ . . .. ..... . ... .. . . ... . ....... . 3-5 l 3.3.1 In trod u c t i o n . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . .

I 3.3.2 Stmetural Materials. . . . . . . . . . . . . ................................................3-5 3.3.3 Poi so n M at eri al . . . .. . . . . . . .. . .. . .. . . . . . . . .. . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . .. . . . .. . . .. . 3 -5 3.3.4 Compatibility with Coolant.......... . .............................................3-7 3.4 Codes, Standards, and Practices for the Spent Fuel Pool Modification.. .. ............ ........... .. . .. ....................3-8 4.0 CRITICalTY SAFETY CONSIDERATIONS ................... .................... .............4-1 l

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l 4.1 Int rod u c t i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. ..... . . . .4-. .1. . . . . . . . . . . . . . . .'

4.2 Abnormal and Accident Conditions........................................... ........................ 4-2 4.3 Inp ut Param et ers . . .. . . . .. . . . . .. .. . . . . ... .. . . .. . . . . . .. . . . . . . ... .. .. .. . .. . . . . . .. . . ......................4-3 4.3.1 Fue? A ssembly Design Specifications................................ ..... .... ......... 4-3 4.3.2 Storage Rack Cell Specifications . ......................................... . .... ......... 4-3 i 4.4 Analysi s M ethod ol ogy . . . . .. . .. .. . . . . . .. ... . . .. . . .. .. . . . . . . . . . . . . . . .. . . . . . . . . .. . . . .... . . . . . . . . ... 4-4 !

4.5 Criticality Analyses and Tolerance l Variations ....... ........ ..... ........ . ..............................................................4-5 l 4.5.1 Nom i nal D esign Case ... . . . . .. ... .. ... . .. . ... . . ... . . . . . ... . . . . . . . . . . .. . . . . .. .. . . . . .. . .... 4-5 !

4.5.2 Uncertainties Due to Manufacturing To l eran c e s . . ... . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . .4-6 .....

4.5.2.1 Boron Loading Variation ...... ........ ...... .. .......... ......... ........ .. .......... . 4-6 :

4.5.2.2 Boral Wiuth Tolerance Variation.............. ...................... .... ....... ... ... 4-6 l 4.5.2.3 Storage Cell Lattice Pitch !

Variation ............... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . .. 4-6 l f \ projects \911080\ license \propriet I

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i 4.5.2.4 Stainless Steel Thickness Tolerances.. . . . . . . ... . . . . . . .... . 4-7 l 4.5.2.5 Fuel Enrichment and Density Variation .. .. . . . . . . . .. . . . . . . . . . . . . . . . ... . . .... 4-7 l i

4.5.2.6 Zirconium Flow Channel . . ..... . . . . . . . . . . ... .. . . . . . . . . 4-7 I 4.5.3 Uncertainty in Depletion Calculations.. . . . . . .. . .. . . ...... 4-8 4.6 Higher Enrichments and GE-11 Fuel. .. . .. ... . . . . . . ... . .. . . . . 4-8 4.7 Abnormal and Accident Conditions.. . .. . . . . .... . . . . . . . . . . . . 4-9 4.7.1 Temperature cad Water Density Effects. .. . . . . . .. .. . .... 4-9 4.7.2 Abnormal Location of a Fuel Assembly.. . . . . .. .. .. . . 4-9 4.7.3 Eccentric Fuel Assembly Positioning. . .. . . . . . . . . . . . . . 4-10 4.7.4 Zirconium Fuel Channel Distortion. . . . . . . . . . . . . . .4-10 4.7.5 Dropped Fuel Assembly.. . . . . . . . . . . . . . . . . . . . .. . . . . . . . . .. . 4-10 4.7.6 Fuel Rack Lateral Movement.. . .. . . . . . . . . . . . . . . . . . . . . ... 4- 10 4.8 Comparison with Other Recently Licensed US Plant. . . . .. . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . .4-11 4.9 Summary and Conclusions.. .. . .. . . . . . . . .. . . . . . . . . . 4-12 4.10 References for Section 4. . . . . ... ..... . . . . . . . . . . . . . . . . . . . . . . .4-13 Appendix A to Section 4 Appendix B to Section 4 5.0 THERMAL-HYDRAULIC CONSIDERATIONS . .. .... .. . . .. . .. . . . .. .. . . .... 5-1 5.1 Introduction.... . ...... . .. .. .. . . . . . . . . . . . . . . . . . . . . . . . . . . . ... 5-1 5.2 Spent Fuel Cooling System Description. .. . . . . .. .... ... .. . . . . . . 5 -2 5.3 Discharge Scenarios.. ... . . . . . . . . . .. . . . . . .. . . ... 5-3 5.4 Discharge Constraints. .. ... . . . . . . .. .. .. 5-6 5.5 Bulk Pool Temperatures .. . . . . . . .. . .. .. . . . . . . . . . . . . . . . . . 5-7 5.6 Local Pool Water Temperature. .. . . . ... .. . . . . . . .5-11 5.6.1 Basis . . . . . . . . , . . . . .. . . ..... . . . . ... . 5- 1 1 5.6.1 Local Temperature Evaluation Methodology . . . . . . .........5-11 5.6.3 Computations. ... . . . . . .. . . . . . . . . . . . . . . 5-14 5.7 Cladding Temperature.. .. . . . . . . . . . .. .. ... . . .. . . . .. .. . 5- 15 5.8 Results.. . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . 5-17 5.9 References for Section 5.. . . . . . . . . . . .. . . . . . . . . . . . . . . . 5-19 l

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1 6.0 SEISMIC / STRUCTURAL CONSIDERATIONS . ............... .... ................. . 6-1 ,

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6.1 Intrnduction .................... ... . . . . . . . . . . . . . . . . . . .....................................6-1 I 6.2 Analysi s O utlin e.. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . .. . . . . . . . . . ........... .. .......... 6-1 6.3 Artificial Time-Histories.. . . . . . . . . . . . . . . ................................6-6 i 6.4 Rack Modeling for Dynamie Simulations ..... ... ....... .. ... .. .... . ......... 6-7 l .

6.4.1 G eneral R emarks .. . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . .. . . . . . .. . . . 6-7 6.4.2 The 3-D 22 DOF Model for l Single Rack Module... .. ........ .... . ............ .. .... ..... 6-9 6.4.2.1. Assumptions... . . . . . . . . . . . . . . . . . . .

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6.4.2.2 Model Details .. . . . ... . . ...... ... . . . ... ..... .. ...................6-11 i 6.4.2.3 Fluid Coupling Details ., ........... .. .... ..... .. . . ..... . .... .. .... .... . ...... 6- 12 l 6.4.2.4 Stiffness Element Details.. ....... . . . . ..... ......... . ... ...................6-13 !

1 6.4.3 Whole Pool Multi-Rack (WPMR)

M od el . .. . . . .. . . . . .. . . . . .. . . . . .. . . . . . . . . . . . . . . . . . ..... 6-14 6.4.3.1 General RemarP . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-14 6.4.3.2 Whole Pool Fluid Coupling . .. ... ..... . . . . . . . . . . . . . .. . . . .... . . 6- 14 6.4.3.3 Coefficients of Friction . . . . . ..... . .. . .. .. ... .. . . . . . . . . . . .. 6-15 l 6.4.3.4 Modeling Details.... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . 6-16 6.5 Acceptance Criteria, Stress Limits and Material Properties.. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . 6-16 6.5.1 Acceptance C;iteria..... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 6-16 6.5.2 Stress Limits aor Various Conditions..... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . 6-18 6.5.2.1 Nermal and Upset Conditions (Level A or Level B). . . . . . . . . . . . . . . . . . . . . . . . . . ... . .. . . 6- 18 6.5.2.2 Level D Service Limits . ......... .. . ........ .. . . ... . ...................6-21 6.5.2.3 Dimensionless Stress Factors.. .. . .. . . . . . . . . . . . . . . . . .. 6-21 6.5.3 Material Properties....... .. . .... .. . . .. .....................................6-22 6.6 Governing Equations of Motion . .. .... .. . . . . . . . . . . . . . . . . . . . . . . . .. . .. . . . 6-22 ;

6.7 Results of 3-D Nonlinear Analyses of i Single Racks .. .. . ............. .. . . . . .... . ....................................6-24 6.7.1 Impact Analyses. . ..... ...... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .6-25 i 6.7.2 Weld S tresses .. .. .... ..... . .. . . . . . ... . .. . . . . . . .. . . . . . . . . . . .. . 5-26 1 6.8 Results from Whole Pool Multi-Rack Analyses.. . . ..... . . . . . . . . . . . . . . . . . . .6-28 6.9 Bearing Pad Analysis .. .. . .... ....... ....... ... . . . . . . . . . . . . . . . . . . . . .. . . . . . . 6-3 0 ;

6.10 Fatigue Analysis of Racks.... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .... 6-31 f:\ projects \911080\ license \propriet 111 t .

TABLE OF CONTENTS SECTION PAGE 6.11 Conclusions.. . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . .... . . . 6-3 3 6.12 References for Section 6. .. . ... . .... . .... . . . . . . . . . . . . . . . . . . . . .6-33 7.0 ACCIDENT ANALYSIS AND MISCELLANEOUS STRUCTURAL EVALUATIONS.. .... .. . . . . . . . . . . .. .. . . . . . . . . . . . . . . . . . . . . . . 7- 1 7.1 Introduction... . ...... . . .. . . . ... ... .. . . . . . . . . . . . . . .... . 7-1 7.2 Refueling Accidents. . .. . . . . . . . . ... . . . .. . . . . .. . 7- 1 7.2.1 Drepped Fuel Assembly... . . . . ... ..... . . . . . . ... ... 7-1 7.3 Local Beckling of Fuel Cell Walls . ... .. .. . . . . . . . . . .7-2 7.4 Analysis of Welded Joints in Rack due to Isolated Ho'. Cell . ... . ..... ........ . .... . . . . . . . . . . . . . . 7-3 7.5 Drop of An Empty Rack From Top of Pool to Liner .. . . . . . . . . . .. . .... . . . . . .. ...... .. .. . 7-4 7.6 Conclusions.. .. .. ...... .... ..... . . . . . . . . . . . . . . . . . . . . . . . . . . . .7-5 7.7 References for Section 7... . . . .. .. . . . . . . . . . . . . . . . . . . . . . . . .. .7-6 8.0 FUEL POOL STRUCTUiW INTEGRITY CONSIDERATIONS .... .. . . . . . . . 8-1 8.1 Introd uction ... . ...... . . .. . .. ... .... .. . . . . . . .. . . . . . . . . . . . . . . . .. . . ... 8- 1 8.2 Description of Spent Fuel Pool Structure . , . ..... . .. ...... .. . . . . . . . . . . . . . 8-1 8.3 ' Definition of Loads.... . . . . . .. ... . . . . . . . . . . . . . . . . . . . . . . . . .. . 8-2 8.3.1 Static Loading . . ... ... .. . . . . . . .. .. . . . .. ...... . 8-2 8.3.2 Dynamic Loading.. .. . . . . . . . . .. ... . .. . . . . . .8-3 8.3.3 Thermal Loading.. . . . . . . . . .. .. .. . . . . . . . 8-3 8.3.4 Cask Drop Loading... . ... . . . . . . . . . . . ..., . ..... ....... 8-4 8.3.5 Embedment Loads...... . . . . . . . . . . . . .. . .. . 8-4 8.4 Analysis Procedures. . . . . . . . . . . . . . . . . . . . . . . . ..... . ........... 8-4 8.4.1 Finite Element Analysis Model.. . . . . ... . . . . . . . . . . . . . .... .. 8-4 8.4.2 Analysis Methodology... . . . . . . . . . . . . . . .. . . . . . . . . . . . 8-5 8.4.2.1 Structure Loadings. . . .. . . . . . . . . . . . . . . .... . . 8-6 8.4.2.2 Concrete Cracking... . . . . . . . . . . . . . . . ... . . . . . 8-7 8.4.3 Load Combinations. . . . . . . . . . . . . . . . . .. . . . . . . . . .8-7 8.5 Results of Analyses.. ... . . ... . . . . . . . . . . . . . . . . . . . . . . . . . . .. . 8-8 8.5.1 Design Basis Load Combinations. .. . . . ... . . 8-8 8.5.2 Spent Fuel Pool Boiling. . . ...... . . .......... . . . . . . . . . . . . . . . . .8-9 fr\ projects \911080\licensc\propriet IV

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8.6 Pool Liner.. .. .. . . . . . . . . . . . . . . . . . . . . . . . . . ............................8-10 8.7 Conclusions ,. .. .. .. ..... ... . .. . .. .. .. .. .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-11 ;

8.8 References for Section 8. .. . .............. ... .... . . ...............................8-11 )

9.0 RADIOLOGICAL EVALUATION. . ... .. . . . . . . . . . . . . . . . . . . . . . . . . .............9-1

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9.1 Solid Radwaste... . . ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . 9-1 9.2 Gaseous Releases .... ..... . . . ... .. .... . .. . .. ....... . . . . . . . . . . . . . . ... 9-1 9.3 Personnel Exposures. . ... . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . ... . . 9-1 9.4 Ar.ticipated Exposure During Re-Racking .. . . . . . . . . . . . . . . . . . . . . . . . . . . .9-2 9.5 Conclucions... .. ...... ......... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . 9-3 10.0 BORAL SURVEILLANCE PROGRAM . . ..... ...... ..... ...... . . .. ..... . . .. . . ... 10-1 i I

1 10.1 P u rpose. . . . . . . .. . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . .. . . . . . . . . . . . . 10- 1 l 10.2 Coupon Surveillance Program .. ..... .. .. . .. . . . .. .............................10-2 10.2.1 Coupon Description .... .. . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-2 10.2.2 Surveillance Coupon Testing '

Sched ule .. ... . . . . ... .. .. . . . . . . . . . ..........................10-3 10.2.3 M easurement Program . . . .... .... .. ............ ..... ..... ..... .... ........ ... ... 10-4 l 10.2.4 Surveillance Coupon Acceptance Criteria . .. . .... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 0-4 10.3 In-Service Inspection (Blackness Tests) .. ............. . .... . . . . .. . . . . . 10-5 10.4 References for Section 10. ........ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 0-6 11.0 ENVIRONMENTAL COST-BENEFIT ASSESSMENT.. ... . . ..... . . . . . . . . .... . . . I 1 - 1 11.1 Introduction ..... . . .... .. .. ... . .. ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .Il-1 11.2 Imperative for Reracking ...... ..... . . .. . .... ..... .... .. .. .. . . . . . . . . . . . . .Il-1 11.3 Appraisal of Alternative Options. .... . .... . . . . . . . . . . . . . . ... 11-1 11.4 Cost E.etimate . . ... ... . . . . . . . . . . . . . . . . . ......... . . . . . . . . . . . . ... 11-7 11.5 Resource Commitment...... .. . .. . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . 11-7 11.6 Environmental Considerations.. .. .. ... .. . . . . . . . . . . . . . . . . . . . . . . . . . .... 11-8 11.7 References for Section 11. .. ... ............... .. .... .... . . . . . . . . . . . . . . . . . . . . 11-8 f:\ projects \911080\ license \propriet v

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l- 5.0. THERMAI HYDRAULIC CONSIDERATIONS 5.1 Introduction This section provides a sununary of the methods, models, analyses and numerical results to demonstrate the compliance of the reracked NMP-1 spent fuel pool with the provisions of Section III of the USNRC "OT Position Paper for Review and Acceptance of Spent Fuel Storage and Handling Applications", (April 14, 1978), including modification letter dated January 18, 1979.

Similar methods of thermal-hydraulic-analysis have been used in over 35 previous licensing efforts on high density spent fuel racks for the past 20 years, including Fermi 2 (Docket 50-341),

Quad Cities 1 and 2 (Dockets 50-254 and 50-265), Rancho Seco (Docket 50-312), Grand Gulf Unit 1 (Docket 50 416), Gyster Creek (Docket 50-219), Virgil C. Summer (Docket i),

Diablo Canyon I and 2 (Dockets 50-275 and 50-323), Byron Units 1 and 2 (Dockets 50-454 and 50-455), St. Lucie Unit One (Docket 50-335), Millstone Point 1 (Docket 50-245), Vogtle Unit 2 (Docket 50-425), Kuosheng Units 1 & 2 (Taiwan Power Company), Ulchin Unit 2 (Korea Electric Power Company), and J.A. FitzPatrick (New York Power Authority).-

The analyses to be carried out for the thermal-hydraulic qualification of the rack array may be broken down into the following categories:

i. Evaluation of the pool decay heat generation rate and pool bulk temperature variation with time. This includes determination of in-core hold time reouirements.

l ii. Determination of the maximum pool local temperature at the instant when the bulk temperature reaches its maximum permitted value.

iii. Evaluation of the maximum fuel cladding temperature to establish that nucleate boiling at any location resulting in two phase conditions environment around the fuel is not possible.

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iv. Evaluation of the time-to-boil if all heat rejection paths from the cooler are lost.

v. Compute the effect of a blocked fuel cell opening on the local water and maximum cladding temperature.

The following sections present a synopsis of the methods employed to perform such analyses and fmal results.

5.2 Spent Fuel Cooling System Description The Spent Fuel Pool Cooling System (SFPCS) contains two parallel loops. Each loop consists of <

one full capacity (600 gpm) pump, one filter, and one approximately 6.0x10 6Btu /hr he.,t exchanger. The system takes suction from the pool surge tanks and circulates the pool water l 1

through the system and retums the cooled and purified water to the pool through diffusers. The SFPCS is designed as NMP-1 Seismic Category I. I Cooling water is supplied to the heat exchangers from the Reactor Building closed loop cooling I (RBCLC) water system. The maximum value of the RBCLC water is bounded by 95 F per the ;

FSAR. However, the RBCLC water temperature can be maintained between 40 F and 95 F depending on the lake water temperature and the coincident heat load to the SFPCS heat l exchangers. '

Initial filling and level maintenance in the spent fuel pool and surge tanks is from the condensate transfer system. The total volume of the surge tanks is approximately 2000 cubic feet. They will normally run at a level of approximately 1000 cubic feet. The difference in surge tank volume allows for the displacement of water when a shipping cask or any other object is placed in the spent fuel storage pool.

l Makeup water is provided by the condensate transfer system. Nomially, makeup is directly l provided to the spent fuel storage pool. Makeup to the spent fuel storage pool is automatically 111-91738 SI!ADED REGIONS ARE IIOLTEC PROPRIETARY INFORM ATION 5-2

initiated when the surge tank volume decre.ises to 800 cubic feet and stops when the volume l reaches 1000 cubic feet. If the makeup to the spent fuel storage pool is not sufficient to maintain surge tank volume, makeup water can be provided directly to the surge tanks. The available rate of makeup (to either the spent fuel storage pool or the surge tanks) is 75 gpm or more. Makeup water can also be supplied directly to the spent fuel pool through fire water hoses.

Any particles that enter the pool either sink to the bottom to be removed by a portable vacuum cleaner or float about in the pool and eventually enter the skimmers, surge tanks and filtering loop. Provision is made for transferring water to the liquD. waste disposal system for processing if the pool water becomes highly contaminated.

The makeup water valve to the spent fuel storage pool is powered from the diesel-generator units. This ensures the supply of makeup water in the event ofloss of both nonnal and reserve a-O power.

5.3 Discharae Scenarios l

Three dkebrge scenarios are considered: the first two scenarios, presented in the following, are intended to demonstrate compliance of the NMP-1 spent fuel pool cooling system with the '

l NUREG-0800. SRP 9.1.3 provisions, although NMP-1 is not an SRP plant. The final scenario "

corresponds to the actual end of cycle plant refueling discharge practice.

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The heat load from the previously stored fuel is computed using the data provided in Table 5.4.1.

It is recognized that the actual discharge dates in the future will deviate from those assumed in Table 5.4.1. However, the effect of such deviations on the heat load and pool water temperature profile is negligible and therefore will not be required to be revisited in the future.

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i. Case 1: SRP Normal Discharge Scenario This scenario simulates the SRP discharge condition labeled as the normal scenario with appropriate modification for 24 month operating cycles. This is not intended to represent the actual refueling of the NMP-1 reactor, it is intended to l illustrate compliance with the provisions of SRP 9.1.3.

The fuel pool is assumed to contain three batches of discharged fuel with 6 years of exposure at full power. The previous two discharges occurred at scheduled 24 month intervals. The final discharge is assumed to occur one year after the second discharge.

l The discharge of fuel assemblies to the pool begins after 150 hours0.00174 days 0.0417 hours 2.480159e-4 weeks 5.7075e-5 months of decay in the reactor, and proceeds at the rate of 120 assemblies for each 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period. Only one fuel pool cooler is assumed to be operating. It is recessary to demonstrate that the maximum bulk pool water temperature is less than 140 F.

ii. Case 2: SRP Full-Core Offload Scenario This case simulates the full-core ofiload scenario of the SRP. This case is intended to illustrate compliance of the spent fuel pool cooling system with the provisions of SRP 9.1.3. It does not represent an actual discharge scenario at NMP-1.

The pool contains three batches of discharges. The first normal batch is assumed to have 24 months decay and the second normal batch (200 assemblics) has decayed for 36 days. The full core.is transferred to the pool after 150 hours0.00174 days 0.0417 hours 2.480159e-4 weeks 5.7075e-5 months decay (at the rate of 120 assemblies per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period). It is assumed that both coolers are operating. The pool water temperature should be kept below boiling to comply I

with the SRP 9.1.3 stipulations.

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W*# %.-m iii. Case 3: Actual End-of-Cycle EOC) Refueling This scenario corresponds to the actual discharge practice at NMP-1. To bound the heat load in the pool, the calculations are carried out at the point in time when I t

the stored fuel inventory is such that the addition of a nonnal batch to the pool I i

will leave it with insufficient capacity to accept another batch while maintaining i the full-core discharge reserve capability. The discharge of fuel is assualed to have occurred in accordance with the schedule presented in Table 5.4.1.

The discharge consists of first transferring the entire core to the pool followed by re-transfer to the reactor of all except the "bumed" normal batch (200 assemblies). I i

i One fuel pool cooling train is assumed to be available and operating. The fuel I pool coolers, however, are assumed to be fouled to their design maximum and l

conservatively assumed to have 5% of the tubes plugged.

l In view of the fact that the RBCLC coolant temperature to the fuel pool ecolers can be varied within a wide range (from 40 F minimum to 95 F maximuc))

depending on the month of the outage and the coincident heat load, the bulk pool water temperature peak will vary. Therefore, it is appropriate to stipulate a maximum temperature limit of 140*F on the peak bulk pool water temperature to ensure that the SRP 9.1.3 nomial temperature limits are not exceeded and that the demineralizer resins in the spent fuel cooling system are not damaged.

Calculations, therefore, are performed to determine T (t is the time-after-reactor-shutdown when all assemblies are transferred to the pool) as a function of the RBCLC water temperature t,i such that the 140 F peak bulk temperature limit is not violated. This calculation, as summarized in the discussion of results section of this chapter, was done for a varying range of RBCLC water temperatures.

111-91738 SII ADED REGIONS ARE IIO1,TEC PROPRIETARY INFOR31 ATION 5-5

L 5.4 Discharce Constraints The following constraints are applicable to all full-core discharge operations to the fuel pool:

i. 140*F bulk pool temperature limit with one cooler running for actual end of cycle (EOC) refuelit.g discharge practice (temperature taken at pump suction)

, . ii. Design basis values of pool water and RBCLC water flow rates to the fuel pool coolers assumed.

iii. The decay heat load calculation is performed in accordance with the provisions of "USNRC Branch Technical Position ASB9-2, " Residual Decay Energy for Light I

Water Reactors for Long Term Cooling", Rev. 2, July,1981.

! iv. The discharge to the pool will begin at no sooner than (t - 89) hours, assuming i the rate of transfer to the pool is limited to a maximum of 6 per hour averaged l over any 24-hour period. However, a more expedited offload may be performed if the plant conditions exist to maintain the pool water temperature at or below 140 F with one SFPC train operating or if higher than assumed flows to spent fuel

cooling are available.

i

v. In view of the fact that the RBCLC coolant temperature to the fuel pool coolers can be varied within a wide range (40*F minimum to 95*F maximum) depending i on the month of the outage and the coincident heat load, cycle specific evaluations shall be done for those times when the RBCLC temperature, within the RBCLC l ,

i temperature limits, to the fuel pool coolers varies from the temperatures addressed in this submittal. These cycle specific evaluations will ensure that with specific RBCLC temperatures and the decay heat from inventory in the pool, the pool bulk i i

111-91738 SHADED REGIONS ARE HOLTEC PROPRIETARY INFORMATION 5-6 ,

i i

['

l l

1 l

l water temperature will be maintained at or below 140 F with one spent fuel cooling train operating. The need for a different hold time, if required, will be addressed as part of this evaluation.

The above constraints to fuel discharge operations will be incorporated in the appropriate plant procedures.

5.5 Bulk Pool Temperatures In order to perform the analysis conservatively, the spent fuel pool heat exchangers are assumed to be fouled to their design maximum and 5% of the tubes are assumed to be plugged. Thus, the temperature effectiveness, p, for the heat exchanger utilized in the analysis is the lowest postulated value calculated from heat exchanger thermal-hydraulic codes. The value of "p" is assumed to remain constant in the calculation.

The mathematical formulation can be explained with reference to the simplified heat exchanger alignment of Figure 5.5.1.

Referring to the spent fuel pool cooler system, the governing differential equation can be written by utilizing consen>ation of energy:

dT \

- Gyr (5.5.1) I C dr =Qt 1 l

Qt = P,, + Q(r)- Gy(T,t,)

where:

1 C: Thennal capacitance of the pool (net water volume times water density and times heat capacity), Btu /"F.

Qc. Heat load to the heat exchanger, Btu /hr.

111-91738 SIIADED REGIONS ARE IIOLTEC PROPRIETARY INFORMATION I

5-7

Q(t): Heat generation rate from recently discharged fuel, which is a specified function of time, T, Btu /hr.

Peon. = B Po: lleat generation rate from "old" fuel, Btu /hr. (Po = average assembly

, operating power, Btu /hr.)

1 Qux: Heat removal rate by the heat exchanger, Btu /hr.

Qey (T,t.): Heat loss to the surroundings, which is a fun tion of pool temperature T and ambient temperature t., Btu /hr.

Qux is a nonlinear function of time if we assume the temperature effectiveness "p" is sonstant during the calculation. Qux can, however, be written in terms of effectiveness "p" as follows:

Qux = W,C,p(T-1,) (5.5.2) t, - r,

" T - t, where:

W:i Design point coolant flow rate, Ib./hr.

C:i Coolant spesific heat, Btu /lb. F.

p: Temperature effectiveness of heat exchanger with 5% of the tubes plugged.

T: Pool water temperature, F ti: Coolant inlet temperature, F to: Coolant outlet temperature, F "p"is obtained by rating the heat exchanger on a Holtec proprietary thermal-hydraulic computer code. Q(t) is specified according to the provisions of"USNRC Branch Technical Position ASB9-2, " Residual Decay Energy for Light Water Reactors for Long Term Cooling", Rev. 2, July, 1981. Q(t) is a function of decay time, number of assemblies, and in-core exposure time. During the fuel transfer, the heat load in the pool will increase with respect to the rate of fuel transfer and equals Q(;) after the fuel transfer.

111 91738 SilADED REGIONS ARE IlOLTEC PROPRIETARY INFOlG1 ATION l 5-8 1

J

l Qev is a nonlinear function of pool temperature and ambient temperature. Qev contains the heat evaporation loss through the pool surface, natuial convection from the pool surface and heat conduction through the pool walls and slab. Experiments show that the heat conduction takes only about 4% of the total heat loss (5.5.1], and can therefore be neglected. The evaporation heat and natural convection heat loss can be expressed as:

Qev = m F A, + he A, O (5-3) where:

m: Mass evaporation rate, Ib./hr. ft.2 F: Latent heat of pool water, Btu /lb.

A: Pool surface area, fl.2 he: Convection heat transfer coefficient at pool surface, Btu /ft.2 hr. F G = T-t.: The temperature difference between pool water and ambient air, *F The mass evaporation rate, m, can be obtained as a nonlinear function of O. We, therefore, have m = ha (0) (Wp, - W..) (5-4) where:

Wp.: llumidity ratio of saturated mo'a air at pool water surface temperature T.

W: liumidity ratio of saturated moist air at ambient temperature to ho(O): Diffusion coefficient at pool water surface. ha is a nonlinear ftmetion of O, Ib./hr.

fl.2 op To increase the conservatism in the evaluations presented herein, all beneficial cooling effects of evaporation are conservatively neglected. This is a substantial conservatism, as evaporation heat losses can be relatively large for large pool surface temperatures.

111-91738 SIIADED REGIONS ARE IlOLTEC PROPRIETARY INFOR51 ATION

, I l 5-9

l The nonlinear single order differential equation (5-1) is solved using Holtec's QA validated 1

numerical integration code "ONEPOOL". j The next step in the analysis is to determine the temperature rise profile of the pool water if all forced indirect cooling modes are suddenly lost. Makeup water is provided from the condensate transfer system.

Clearly, the most critical instant of loss-of-cooling is when the pool water has reached maximum value. The governing enthalpy balance equation for this condition can be written as dT C = P_ + G(r + r,,,,)- Gy (5.5.5)

In the foregoing, water is assumed to have specific heat of unity and latent heat L, and the time coordinate T is measured from the instant maximum pool water temperature is reached. T. is the time coordinate measured from the instant of reactor shutdown to when maximum pool water l temperature is reached. T, the spent fuel pool water temperature, is the dependent variable.

l 1

To increase the conservatism in the evaluations preser.ted herein, all beneficial cooling effects of evaporation are conservatively neglected. This is a substantial conservatism, as evaporation heat losses can be relatively large for the large pool surface temperatures that occur as the pool temperature approaches the saturation temperature.

l A QA validated numerical quadrature code is used to integrate the foregoing equation. The pool water heat-up rate, time-to-boil, and subsequent water evaporation-time profile are generated and compiled in order to perform a safety evaluation.

111-91738 SIIADED REGIONS ARE IIOLTEC PROPRIETARY INFOlGIATION 5-10

5.6 Local Pool Water Temperature i

In this section, a summary of the methodology, calculations and results for local pool water temperature is presented.

l 5.6.1 Basis i i

i The local water temperature analysis uses bounding bulk pool heat load conditions coincident with peak bulk temperature response. In order to determine an upper bound on the maximum I local water temperature, a series of conservative assumptions are made. The most important j assumptions are listed below:

The minimum rack-to-wall and rack-to-floor gaps are used. l The sparger pipe is conservatively assumed to be truncated at an elevation of 336'- i 6"(actual cut to occur at approximately elevation 333*).

l No downcomer flow is assumed to exist between the rack modules.

Decay heat loads shown in Table 5.6.2 are used for the local water temperature analysis. The values used are conservative, bounding all values calculated and l applied for both previously and freshly discharged fuel assemblies.

5.6.2 Local Temperature Evaluation Methodolouv To determine the maximax (maximum in time and space) temperature in the NMP-1 fuel pool, it is necessary to rigorously quantify the velocity fic: a the pool created bs the interaction of buoyancy driven flows and water injection / egress. A Computational Fluid Dynamics (CFD) analysis for this demonstration is required. The objective of this study is to demonstrate that the principal thermal-hydraulic criteria of ensuring local subcooled conditions in the pool is met for the limiting fuel loading scenario. The local thermal-hydraulic analysis also considers the effect of a fuel assembly dropped above the racks and laying horizontally on top of the hot fuel rack 111-91738 SIIADED REGIONS ARE IlOLTEC PROPRIETARY INFOIO1 ATION 5-11

I cells. An outline of the CFD approach is described in the following.

I There a e several significant geometric and thermal-hydraulic features of the NMP-1 spent fuel pool wluch need to be considered for a rigorous CFD analysis. From a fluid flow modeling standpoint, there are two regions to be considered. One region is the bulk pool region where the classical Navier-Stokes equations are solved with turbulence effects included. The other region is the heat generating fuel assemblies located in the spent fuel racks located near the bottom of the spent fuel pool. In this region, water flow is directed venically upwards due to buoyancy forces through relatively small flow thannels formed by the BWR fuel assemblies in each rack ;

cell. This situation shall be modeled as a porous solid region in which fluid flow is governed by the classical Darcy's Law:

8P p

=-

g{j) V, - @, / 2 where Sp/SXi is the pressure gradient, K(i), Vi and C are the corresponding permeability, velocity and inertial resistance parameters and p is the fluid viscosity. The permeability and inertial resistance parameters for the rack cells loaded with BWR fuel is determined based on friction factor conelations for laminar flow conditions typicallf encountered due to low buoyancy induced velocities and small size of the flow channels.

The NMP-1 pool geometry requires an adequate portrayal oflarge scale and small scale features, spatially distributed heat sources in the spent fuel racks and water inlet / outlet configuration.

Relatively cooler bulk pool water normally flows down through the narrow fuel rack outline to pool wall liner clearance known as the downcomer. Near the bottom of the racks, the flow tums from a vertical to horizontal direction into the bottom plenum supplying cooling water to the rack cells. Heated water issuing out of the top of the racks mixes with the bulk pool water. An adequate modeling of these features on the CFD program involves meshing the large scale bulk pool region and small scale downcomer and bottom plenum regions w:th sufficient number of computational cells to capture the bulk and local features of the flow field.

H1-91738 SilADED REGIONS ARE IlOLTEC PROPRIETARY INFORM ATION 5-12

The distributed heat sources in the spent fuel pool racks are modeled by identifying distinct heat generation zones considering full-core discharge, bounding peak effects, and presence of background decay heat from old discharges. Three heat generating zones were modeled. The first consists of background fuel from previous discharges, the remaining two zones consist of fuel from a full-core-discharge scenario. The two full-core discharge zones are differentiated by one zone with higher than average decay heat generation and the other with less than average decay heat generation. This is a conservative model, since all of the fuel with higher than average decay heat will be placed in a contiguous area. .A uniformly distributed heat generation rate is applied throughout each distinct zone. The analysis has been performed for a limiting full-core offload decay heat scenario which bounds all of the offload scenarios listed earlier.

The CFD analysis is performed on the industry standard FLUENT [5.6.4] fluid flow and heat transfer modeling program. The FLUENT code enables buoyancy flow and turbulence effects to be included in the CFD analysis. Turbulence effects are modeled by relating time-varying "Reynolds's Stresses" to the mean bulk flow quantities with the following turbulence modeling options:

(i) x-c Model; (ii) RNG x-c Model; (iii) Reynolds Stress Model The x-c Model is considered most appropriate for the NMP-1 CFD analysis. The x-c turbulence model is a time-tested, general purpose turbulence model. This model has been demonstrated to give good results for the majority of turbulent fluid flow phenomena. The Renormalization Group (RNG) and Reynolds Stress models are more advanced models that were developed for situations where the x-c model does not provide acceptable results, such as high speed flow and supersonic shock. The flow regime in the bulk fluid region is such that the x-c model will provide acceptable results.

Rigorous modeling of fluid flow problems requires a solution to the classical Navier-Stokes 111-91738 SIIADED REGIONS ARE IIOLTEC PROPRIETARY INFORM ATION 5-13 i

\

{

m l

equations of Guid motion [5.6.1]. The goveming equations (in modified form for turbulent flows with buoyancy effects included) can be written as:

< v l Sp,u i + O ,u uPij a af, de i Ot &; Dx, _\Dx,

Br Dp OP kui 'u j g - P.p(T- T,)g, +

where u; are the three time-averaged velocity components. p(ui uj ) are time-averaged Reynolds stresses derived from the turbulence induced fluctuating velocity components ui , po is the fluid density at temperature To, p is the coefficient of thermal expansion, is the fluid viscosity, gi are the components of gravitational acceleration and x; are the Cartesian coordinate directions. The Reynolds stress tensor is expressed in tenns of the mean flow quantities by defining a turbulent viscosity pi and a turbulent velocity scale k"* as shown below [5.6.2]:

/. .i as. ad j P\ui u> l = 2 I 3pkS g - p, +

The procedure to obtain the turbulent viscosity and velocity length scales involves a solution of two additional transport equations for kinetic energy (k) and rate of energy dissipation (c). This methodology is known as the k-c model for turbulent flows as described by Launder and Spaiding [5.6.3].

5.6.3 Computations For conservatism in the CFD analysis, the full-core off-load is discharged in the poisoned rack region near the end wall opposite from the water ingress region. This modeling scenario of remotely locating hot fuel away from cooling water ingress maximizes resistance to supply of relatively cooler incoming water to the hot fuel assemblies and is therefore considered to be a conservative scenario. The conservatively imposed fuel loading scenario thus provides bounding 111-91738 SilADED REGIONS ARE IlOLTEC PROPRIETARY INFORM ATION 5-14

local temperature results. The fuel pool water inlet near the west wall is conservatively discharged downwards at a point in the bulk pool region above the racks. In this manner, no credit is taken for the relatively cooler water entry into the bottom plenum. A two-dimensional CFD model of the NMP1 spent fuel pool is considered for analysis. Radial peaking effects (see Table 5.7.1) are included by requiring that a third of the full-core off-load discharged assemblies are generating higher than average decay heat. The balance of assemblies will therefore be generating decay heat at levels lower than average. In the CFD model, all the hotter than average assemblies are stored in neighboring locations so that any flow starvation effects are conservatively encouraged. The balance of older assemblies in the spent fuel pool are considered to be generating heat at a uniform background level. These threc distinct heat generating regions are identified as Hot, Warm, and Old. In Table 5.6.2, the corresponding volumetric heat generating levels calculated based on input data and NhiPI spent fuel pool layout are provided.

Results of CFD analysis are presented in Figures 5.6.1 through 5.6.5. In Figure 5.6.1, the computational grid applied to the problem is depicted. In Figure 5.6.2, the actual scaled grid in the physical domain is depicted. The porous heat generating racks region is shown in blue and the inlet / outlet regions are labeled in Figure 5.6.1 as 11 and 12, respectively. The temperature contour and velocity vector plots are shown in Figures 5.6.3 and 5.6.4. In Figure 5.6.5, an axial local temperature plot in the hottest fuel cell is shown. Note that the highest local temperature is encountered at the exit of the cell. The maximum fuel cladding temperatures were calculated and found to be limited to 47.9 F above these temperatures. In conclusion, the rigorous CFD analysis demonstrates that localized subcooled boiling in the NMP1 spent fuel pool racks is a noncredible possibility.

5.7 Claddine Temperature l In this section, the method to calculate the temperature of the fuel cladding is presented.

l 111-91738 Sil ADED REGIONS ARE IIOLTEC PROPRIETARY INFORMATION 5-15

1 The maximum specific power of a fuel array qr can be given by:

l 9a = qF,,

where:

Fxy = radial peaking factor q =

average fuel assembly specific power I l

The peaking factors are given in Table 5.4.2. The maximum temperature rise of pool water in the most 6 advantageously placed fuel assembly, defined as one which is subject to the highest local pool water temperature, is computed for all loading cases. Having determined the maximum local water temperature in the pool, it is now possible to determine the maximum fuel cladding temperature. A fuel rod can produce F, times the average heat emission rate over a small length, where F is the axial rod peaking factor. The axial heat distribution in a rod is generally a maxin um in the central region, and tapers off at its two extremities. Thus, peak j cladding heat flux is given by the equation:

qF,,F, q, =

4 where Ac is the total cladding extemal heat transfer area in the active fuel it igth region.

Within each fuel assembly sub-channel, water is continuously heated by the cladding as it moves axially upwards from bottom to top under laminar flow conditions. Rohsenow and Hartnett

[5.7.1] report Nusselt-number based heat transfer correlation for laminar flow in a heated channel. The film temperature driving force (ATr) at the peak cladding flux location is calculated as follows; h,D, i Kw = Nu (5.7.4)

ATf

= 1f 111-91'738 SIIADED REGIONS ARE IIOLTEC PROPRIETARY INFORMATION 5-16

where, hr is the waterside film heat transfer coefficient, Dh is sub-channel hydraulic diameter, Kw is water thermal conductivity and Nu is Nusselt number from heat transfer correlation.

In order to introduce some additional conservatism in the analysis, we assume that the fuel cladding has a crud deposit resistance Re (equal to 0.0005 fl2 -hr *F/ Btu), which covers the entire surface. This value is assumed because of the lack of any value known by GE at this time. This value is assumed to be conservative and has been used in many other previous licensing submittals. Thus, including the temperature drop across the crud resistance, the cladding to water

~

local temperature difference (ATc) is given by:

AT, = ATf + R,q, (5.7.5) 5.8 Results Table 5.8.1 presents the major design input for bulk pool temperature analysis. For the first two presented scenarios the maximum bulk pool temperature is detemiined, as are the minimum time-to-boil and the corresponding maximum boil-off rate. For the fmal presented scenario, the in-core hold time requirements to prevent exceeding the temperature limit are calculated, as are I the minimum time-to-boil and the corresponding maximum boil-off rate, for the four RBCLC water temperatures. The maximum local water and fuel cladding temperatures are calculated for the evaluated scenario with the highest bulk temperature peak.

1 The results of the first scenario, corresponding to the SRP normal discharge scenario, are presented in Tables 5.8.2 and 5.8.3. Calculations show that the maximum SRP normal temperature is limited to 119.8 F for single train operation, which is below the SRP limit of 140'F. The maximum decay heat load is 5.99x106 Btu /hr. The minimum time-to-boil for this scenario is 37.39 hours4.513889e-4 days 0.0108 hours 6.448413e-5 weeks 1.48395e-5 months and the maximum boil-off rate is 11.61 gpm. This calculated maximum boil-off rate is well within the 75 gpm available makeup rate.

111-91738 SilADED REGIONS ARE IlOLTEC PROPRIETARY INFOlG1ATION 5-17

l The results of the second scenario, corresponding to the SRP full core discharge scenario, are also presented in Tables 5.8.2 and 5.8.3. Calculations show that the maximum SRP emergency !

temperature is limited to 138.3*F for dual train operation, which satisfies the SRP acceptance criteria of no boiling. The maximum decay heat load is 20.14x106 Btu /hr. The minimum time- ;

to-boil for this scenario is 8.37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months and the maximum boil-off rate is 42.35 gpm. This ;

calculated maximum boil-off rate is also within the 75 gpm available makeup rate.

The results of the third scenario, corresponding to the actual NMP-1 end of cycle refueling i discharge scenario, is also presented in Tables 5.8.2 and 5.8.3. For this case, the maximum temperature limit is fixed and the in-core hold time required to prevent exceeding this limit is calculated based on a varying RBCLC coolant water temperatures of 40 F,60 F,80 F, and 95 F maximum. The minimum in-core hold time is 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , govemed by radiological requirements.

Shoner in-core hold times will not be evaluated, even if the corresponding peak pool bulk temperature is less than the maximum temperature limit. Note that this in-core hold time is for a 6 assembly per hour transfer rate. As long as the fuel transfer does not end earlier than 161 hours0.00186 days 0.0447 hours 2.662037e-4 weeks 6.12605e-5 months after shutdown (72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in-core plus 89 hours0.00103 days 0.0247 hours 1.471561e-4 weeks 3.38645e-5 months transfer), the in-core hold time shall be extended if a faster late is used (i.e.,90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months in-core and 7.5 assemblies per hour). The minimum time-to-boil for these scenarios is 8.87 hours0.00101 days 0.0242 hours 1.438492e-4 weeks 3.31035e-5 months and the maximum boil-off rate is 43.72 gpm. This calculated maximum boil-off rate is within the 75 gpm available makeup rate.

The maximum bulk pool temperature reached in any of these three scenarios is during the actual end of cycle refueling conesponding with a RBCLC coolant water temperature of 95*F. The peak bulk pool temperature for this condition is 140.0 F. The corresponding maximum local water and fuel cladding temperatures (see Table 5.8.4) are 150.7'F and 198.6'F, respectively.

The hydraulic resistance used in the CFD model was increased to bound the 50% blockage scenario. No nucleate boiling is indicated at anylocation.

In conclusion, the thermal-hydraulic analysis of the NMP-1 spent fuel pool cooling system 111-91738 SIIADED REGIONS ARE IIOLTEC PROl'RIETARY INI/ORMATION 5-18

l l

demonstrates that the system complies with the cooling capability requirements of SRP 9.1.3.

Further, the maximum pool bulk temperature during an actual end of cycle refueling scenario can be maintained below the respective acceptable temperature limits, even if a single active failure )

is considered. The coincident local pool water and fuel cladding temperatures were shown to be below the nucleate boiling point. Therefore, the fuel cladding is subject to negligible themial tress and long-term integrity of the spent fuel assemblies is ensured.

5.9 References for Section 5

[5.5.1] " Heat Loss to the Ambient From Spent Fuel Pools: Correlation of Theory with Experiment", Holtec Report HI-90477, Rev. O, April 3,1990.

[5.6.1] Batchelor, G.K., "An Introduction to Fluid Dynamics", Cambridge University Press,1967.

[5.6.2] Hinze, J.O., " Turbulence", McGraw Hill Publishing Co., New York, NY,1975.

[5.6.3] Launder, B.E., and Spalding, D.B., " Lectures in Mathematical Models of )

Turbulence", .\cadernic Press, London,197 .

l

[5.6.4] "QA Documentation and Validation of the FLUENT Version 4.3 CFD Analysis Program", Holtec Report HI-961444. l

[5.7.1] Rohsenow, N.M., and Hartnett, J.P., " Handbook of Heat Transfer", McGraw Hill Book Company, New York,1973.

l 111-91738 SIIADED REGIONS ARE IlOLTEC PROPRIETARY INFORMATION

! 5-19

i.

I i Table 5.4.2 i DATA FOR DISCHARGE SCENARIOS Number of assemblies in full core 532 Number of fuel pool coolers in parallel 1 Maximum fuel exposure time, hours 52,600 .)

l 1

l l

~ 111-91738 SHADED REGIONS ARE HOLTEC PROPRIETARY INFORMATION 5-21 l

L

Table 5.4.1 NMP-1 EXISTING AND PROJECTED FUEL DISCHARGE SCIIEDULE ...

End of Cycle Bundles Total Number Discharged Permanently Bundles Date Discharged Discharged 01a 17 17 901 Olb 31 48 402 01e 104 152 403 02 148 300 3n4 03 200 $00 905 04 160 660 307 05 168 828 3n9 0/ 200 1,028 3/81 07 216 1,244 3/84 08 200 1,444 3/86 09 176 1,620 12/87 10 192 1,812 2/93 11 200 2,012 2/95 12 188 2,200 3/97 13 200 2,400 3/99 14 200 2,600 3/2001 15 200 2,800 3/2003 16 200 3,000 3/2005 17 200 3,200 3/2007 18 200 3,400 3/2009 19 200 3,600 3/2011 Note: All fuel bundles in the paol are assumed to have 6 years full power operation in reactor.

111 91738 SHADED REGIONS ARE HOLTEC PROPRIETARY INFORMA' LION 5-20

/

4 Table 5.6.1 l

l DATA FOR LOCAL TEMPERATURE Fuel Assembly Array Size 8x8 Fuel cladding outer diameter, in. 0.484 Fuel cladding inside diameter, in. 0.414 Storage cell inside dimension, in. 5.90 l Active fuel length, in. 144 Number of fuel rods / assembly. 62 Operating power per fuel assembly Po x 104 , 11.873 Btu /hr Cell pitch, in. 6.06 Cell height, in. 165 Bottom height, in. 5.75 Note: For the analysis, inertia resistance was increased by 50% and permeability was reduced by 15% for conservatism in order to account for any' slight discrepancies in the.

dimensions in the above table for ott'er fuel types.

l 111 91738 SilADED REGIONS ARE IlOLTEC PROPRIETARY INFORMATION 5-22

[

Table 5.6.2

SUMMARY

OF VOLUMETRIC HEAT GENERATION LATE CALCULATIONS Hot Region 6

Decay Heat 11.03 x 10 Btu /hr 3

Volume 729.88 R Volumetric Heat Generation 15,112.8 Btu /n3-hr Warm Region 6

Decay Heat 8.36 x 10 Btu /hr Volume 1,458.07 fl 3 Volumetric Heat Generation 5,682.9 Blu/fl3 -hr Old Recion 6

Decay Heat 3.82 x 10 Btu /hr Volume 14,652.13 R 3 3

Volumetric Heat Generation 260.6 Btu /A -hr 111-91738 SHADED REGIONS ARE HOLTEC l'ROPRIETARY INFORMATION l 5-23

i l

i Table 5.7.1 PEAKING FACTORS FACTOR VALUE Radial 1.70 Total 3.02 l I

1 1

l l

l

, l l 1 l I l

l ,

1 I

l 111-91738 SIIADED REGIONS ARE IIOLTEC PROPRIETARY INFORMATION 5-24

l Tal,le 5.8.1 MAJOR DESIGN INPUT Net water volume of pool, ft' 36,774 l Fuelpoolthemialcapacity,106Btu / F 2.25 6

l. Average operating power of a fuel assembly,10 Btu /hr 'l1.873 1

Coolant (RBCLC) inlet temperature, 'F 95 max.; 40. min.

l Coolant (RBCLC) flow rate,106 lb/hr 0.3 l

l l

111-91738 SHADED REGIONS ARE 110LTEC PROPRIETARY INFORMATION l-5-25

l l Table 5.8.2 SFP BULK POOL TEMPERATURES f

Case Number Maximum Pool Coincident Time Maximum (RBCLC Temp.) Temperature After Reactor (Coincident) Net (Required Hold Shutdown Heat Load Timel) _

Case 1 119.8 F 208 hrs 5.99 MBtu/hr (95 F) (n/a) (5.70 MBtu/hr)

Case 2 138.3*F 266 hrs 20.14 MBtu/hr (95*F) (n/a) (19.92 MBtu/hr)

Case 3 130.1 F 177 hrs 21.46 MBtu/hr (40 F) (72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ) (20.72 MBtu/hr)

Case 3 140.0 F 250 hrs 18.91 MBtu/hr (60*F) (141 hours0.00163 days 0.0392 hours 2.331349e-4 weeks 5.36505e-5 months ) (18.39 MBtu/hr) ,

Case 3 140.0 F 573 hrs 14.05 MBtu/hr (80'F) (458 hours0.0053 days 0.127 hours 7.572751e-4 weeks 1.74269e-4 months ) (13.80 MBtu/hr)

Cue 3 140.0 F 1129 hrs 10.48 MBtu/hr (95 F) (1008 hours0.0117 days 0.28 hours 0.00167 weeks 3.83544e-4 months ) (10.35 MBtu/hr)

Notes:

1. Required hold time (t - 89) is based on 6 assembly per hour fuel nansfer rate. Slower fuel transfer rates may allow shoner hold times. Coincident time is De time when the maximum temperature is reached.

HI-91738 SHADED REGIONS ARE HOLTEC PROPRIETARY INFORMATION 5-26 l

l L

r-Table 5.8.3 RESULTS OF LOSS-OF-COOLING Case Number Time-to-Boil (Without Maximum Evaporation j Makeup Water) Rate l 1 37.39 hrs 11.61 gpm l 2 8.37 hrs 42.35 gpm l

l 3 (40 F) 8.97 hrs 43.72 gpm 3 (60 F) 8.87 hrs 39.05 gpm 3 '80 F) 11.79 hrs 29.41 gpm 3 (95*F) 15.70 hrs 22.09 gpm l

l l'

l l

l-i l

l-l 111-91738 SHADED REGIONS ARE HOLTEC PROPRIETARY INFORMATION 5-27 l

l

I' Table 5.8.4 1

MAXIMUM LOCAL POOL WATER AND FUEL CLADDING TEMPERATURES Maximum Bulk SFP Temperature (*F) 140.0 Maximum Cell Temperature Rise (*F) 10.7 Maximum Local Water Temperature ( F) 150.7 Maximum Fuel Clad Superheat ( F) 47.9 Maximum Fuel Clad Temperature ( F) 198.6 i ,

i The peak cladding temperatures reported herein are based on 144" active fuel length. Minor (16") variations in the active fuel length will produce extremely small changes in the peak cladding temperatures.

111-91738 SHADED REGIONS ARE IIOLTEC PROPRIETARY INFORMATION 5-28

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r ATTACHMENT B r

I l

l I

l Additional Information Concerning Question No.10 l

1 Ountian No.10 Currently, the Spent Fuel Pool (SFP) high temperature alarm setpoint is 113*F. NMPC procedure N1-ARP-L1, " Control Room Panel L1," provides direction if the high temperature alarm is received. This procedure directs the operator to increase cooling flow per N1-OP-6,

" Fuel Pool Filtering and Cooling System," in order to maintain SFP temperature between 75*F and 105*F.

NMPC procedure N1-OP-34, " Refueling Procedure," contains the SFP cooling requirements to ensure that the SFP temperature design basis limit is not exceeded during a core offload. Prior to a refueling outage, the SFP temperature alarm setpoint is administratively lowered to 100*F. If the SFP temperature reaches 100*F, the system temperature will be monitored and logged at 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months intervals With the core fully offloaded to the SFP, if the temperature is expected to reach 120*F in four hours or less, the operators are directed to: 1) operate both loops of SFP Cooling at maximum flow,2) secure all non-essential Reactor Building Closed Loop Cooling (RBCLC) system loads, 3) manually control the RBCLC Temperature Control Valve (TCV) to establish the lowest RBCLC header temperature possible, and 4) monitor SFP temperature at one-hour intervals. If the SFP temperature does not stabilize within two hours, increased cooling water flow is provided to the RBCLC heat exchangers.

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