ML20093L117

Revised Sar,Incorporating Proposed Section 9.20.8

ML20093L117
Person / Time
Site:	University of Virginia
Issue date:	10/31/1995
From:	VIRGINIA, UNIV. OF, CHARLOTTESVILLE, VA
To:
Shared Package
ML20093L108	List: ML20093L104 ML20093L117
References
NUDOCS 9510250199
Download: ML20093L117 (38)

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Table of Contents Page l

1 1.0 Introduction .............. .... ......... ..... ................ 1-1 l

1 1.1 Summary of Previous Documentation (Revised 10/95) . . ................. 1-2 2.0 General Description of Facility ...... ... .. . ...................... 2-1 i 2.1 Reactor Site . . . ............... . .. ......... .. ......... . 2-7  ;

1 2.2 Reactor Building ......... ......... . ................... ... 2-7  !

2.3 New Construction .. .................. ........ . . ......... 2-7 l 2.4 Wind Direction and Velocity ..................................... 2-13 i 2.5 Hydrology .............. ............. ....................2-13 j 2.6 S e ism o logy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2- 15 )

i 3.0 Reactor Components and Control ... ........ ..... .. ............ 3-1 j 3.1 Reactor Assembly .......... .... .. ... ..... ... .... ....... . 3-1 l 3.2 Fuel Elements ....... .... . ............ . . ........... ... 3-4 3.3 Fuel Plates . . . . . . . ..................... .. ........... ... . 3-9 3.4 Control Rods and Drives ... ... ............ ... .... ..........3-9 3.5 Reactor Reflectors ... .. .... . . ..... . . ..... . . . 3-12 1 3.6 Core Loadings .. .. .......... .... ... ..... ..... ... 3-12 3.7 Fuel Storage Facilities .... ............ ..... . .. ....... . 3-14 ,

3.8 Reactor Data . . . . .......... .................. .. ..... .. 3-16 l 3.9 Reactor Kinetics .... .. .... . . ........ ................ . 3-20 3.10 Fission Product Inventory . . . . . . . . .. ......... ... ......... 3-21 3.11 Nuclear Instrumentation . . . . . . . .......... . .. ............. 3-21 3.11.1 General Description ........ . . ... . ...... ..... 3-21 3.11.2 Source Range Circuit ... .......... ..... ............ 3-24 3.11.3 Intermediate Range Circuit . . .......... . ..... .... . 3-26 3.11.4 Power Range Drawer . .. . . . . . ....... . . . . . 3 -29 3.11.5 Scram Logic Drawer ... ... ....... ............ .... 3-31 3.12 Scrams, Interlocks and Ahrms ...... . ...... .... ......... . 3 38 3.12.1 Scrams .. ..... . . . . . ......... . .... . . . . 3-3 8 3.12.2 Interlocks . . . . . . . . . . ..... ........................... 3-39 3.12.3 Alarms .............................................3-39 3.13 Automatic Control for Maintaining Constant Power . . . . . . . . . . . . . . . . . . . . 3-40 4.0 Reactor Systems .... ... ... .... .. ....... . . . .. . . ..... 4-1 4.1 Pool . .. .. ....... ........ .... ... ..... .......... 4-1 4.2 Filling and Draining the Pool . ... .. .... ..... . .. . .. ... 4-1 4.3 Primary Cooling System . .... . . . . .. . . 43 4.4 Measurement of Temperature Differential . ..... .. . .. . . . . . 4 -6 4.5 Secondary Cooling System ......... .. . . . .. 4-6 4.6 Design Specifications (Revised 10/95) ............ . .. . . . 4-7-A 4.6.1 Replacement of Heat Exchanger System Components (10/95) . . . 4-7-A 4.7 Water Purification (Revised 10/95) ...... . .. . .......... ..... 4-7-B 4.8 Liquid Waste Disposal System (Revised 10/95) . ... ... . . .. . 4-7-B 4.9 Building Ventilation System and Airborne Eftluents . . .. . .. ... . 4-12 4.10 Core Spray System . . . . .... ... . . .. . . . . . . . . . . 4-26 9510250199 951019 PDR ADDCK 05000062 P PDR

ii Table of Contents (cont.)

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5.0 Exocriment Facilities ...................... .......... .............. 5-1 l 5.1 Beamports .......... ................... .... . .. .......... 5-1 j 5.2 Large Access Facilities .. ......................... . ... ...... 5-6 l 5.3 Rabbit Facility . . . . . . . . . . . . . . . . . . ........... .. .......... 5-7  ;

5.4 Fueled Experiments ................ .... ... ... .. ........ . 5-9 I

6.0 Rabbit Hazards ................ . .......... ....................... 6-1 6.1 Confinement .......................... .... ..... .......... 6-1 6.2 Shielding .................... ............... ... ......... 6-6 6.3 Hazards During Normal Operations ....... ............... ....... 6-13 7.0 Health Physics ... .............. ....... . .. . .................. 7-1 7.1 General Information .................. .... ......... ....... 7-1 7.2 Education in Health Physics .... .. ... .... . ... ... ........ 7-1 7.3 Personnel Monitoring and Protection . . . .. ....... ........ .... 7-2 7.4 Permanent Monitoring and Surveys . . . . . .... ... . .......... 7-2 )

7.5 Prohibitions and Sanctions ......... .. ... . . . . .... ........ 7-3 l 7.6 Waste Disposal .... .. . ... .... .... . . .. ..... 7-4 1 7.7 Shipping and Transport .. . . . .... .............. ........ . 7-4 l l

8.0 Administration . . . . . . .... .... . .... .. .... . ............... 8-1 8.1 General Organization . . . . . ... ... .... ........... . ....... 8-1 8.2 Radiation Safety Committee ..... .... ... .. ........... ...... 8-1 8.3 Reactor Safety Committee ................ ....................... 8-1 8.4 Procedures .... . .. ......... ............ ... ........... 8-3 9.0 Safety Analysis . . . . ........... .. .. ..... .. . . ... ........ ... 9-1 9.0 Safety Analysis . . . . . . . . . . . . ...... .... .. ................... 9-1 l 9.1 Thermal Hydraulic Analysis of the UVAR .... .. .. .......... ..... 9-1 9.2 Forced Convection Heat Transfer . ............. ...... ........... 9-4 9.3 Prediction of incipient Boiling ... .. ............ ...... .. . . 9-5 9.4 Burnout Heat Flux . .... .. ...... . ..... ................... 9-6 9.5 Flow Instability ...... ..... . . . ............. ......... . . 9-12 9.6 Burnout Ratio ........ .... .... .. ...... ........... . . . 9-18 9.7 Nomenclature Used in Thermal Hydraulic Analysis . .. ...... ....... . 9-19 9.8 Hot Channel and Minimum Core Loading . . . ............... ... 9-21 9.9 Allowance for Error in the Burnout Determination . . .. .. . . 9-24 9.10 Safety Limit ..................... .... . . .. ..... 9-24 9.11 Limiting Safety System Settings and Measurement Errors . . . . . . 9-31 a) Coolant Inlet Temperature .............. . .. .. . ... . . . 9-31 b) Flow Rate ..... ...... .... .. . . .. . . 9-31 c) Reactor Power .......... .. .. .. .. . ..... . ..... 9-32 9.12 Short Period Transient .. ..... ........... .. . .. . 9-32 9.13 Loss of Flow Transient and Natural Convection . .. . . . . . 9-35 9.14 HEU Analysis for Loss of Flow Transient .... . . .. .. ....... 9-36 9.15 HEU Analysis for Natural Convection . . . ... ... ... .. . . ....... 9-50 9.16 Maximum LEU-22 Fuel Temperatures Following LOCA .. ...... . . . . 9-51 a) Introduction ... .......... .. ..... . ... . . . . . . . . . 9-51 b) Calculation of Peak Fuel Temperature Following LOCA .. . . . . . . . . 9-5 4

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Table of Contents (cont.T Page 9.17 - Emergency Core Spray System Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 ,

9.18 Time to Uncover Core Following a LOCA .......................... 9-65 ,

a) Flow Rate Without Frictional Losses . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-66 l b) Flow Rate With Frictional lesses ..............................9-69 c) Time to Uncover Core With Double-Ended Pipe Break . . . . . . . . . . . . . . 9-71 d) Time to Uncover Core With Crack in Pool Wall ....................-9-72 9.19 Heat Exchanger Secondary Tube Plugging Analysis (10/95) ............... 9-73 Risk Associated with Plug Failure .........................

9.19.1 9-73

, 9.19.2 Probability of Plug Failure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-74 l

9.19.3 Conclusion Regarding Risk of Plug Failure . . . . . . . . . . . . . . . . . 9-76 l 9.20 Heat Exchanger Primary-to-Secondary Leak Analysis (10/95) . . . . . . . . . . . . 9-77 l 9.20.1 Introduction . .......................................9-77 9.20.2 Basis for Primary-to-Secondary Leak Ratc . . . . . . . . . . . . . . . . . . . 9 -77 9.20.3 Conditions Prevailing at Start of Scenario . . . . . . . . . . . . . . . . . . . 9-7 8 9.20.4 Initiating Scenario Event ' . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-78 9.20.5 Primary-to-Secondary Leak Progression ..................... 9-79 9.20.6 Mitigation of Heat Exchanger Leak . . . . . . . . . . . . . . . . . . . . . . . . 9-79 9.20.7 Calculated Release Rates ............................... 9-80 9.20.7.1 Calculation of August 1995 Heat Exchanger Leak Rate . 9-82 9.20.7.2 Airborne Release Calculation (Case 1) . . . . . . . . . . . . 9-83 9.20.7.3 Water-Borne Release Calculations (Case 2) ..... .. 9-84 9.20.7.4 Release Analyses Conclusions .................. 9-86 9.20.8 Double-Ended-Heat-Exchanger-Tube-Break leak Analysis . . . . . . . . 9-87 9.20.8.1 Introd uction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-87 9.20.8.2 Conditions Prevailing at Start of Scenario .......... 9-88 9.20.8.3 Initiating Scenario Event . . . . . . . . . . . . . . . . . . . . . . . . 9-88 9.20.8.4 Primary-to-Secondary Leak Progression and Mitigation . 9-89 9.20.8.5 Scenario Consequences . . . . . . . . . . . . . . . . . . . . . . . . 9-89 9.20.8.6 Airborne Release Calculations (Case 1) . . . . . . . . . . . . 9-90 9.20.8.7 Water-borne Release Calculations (Case 2) . . . . . . . . . 9-92 9.20.8.8 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-94 9.21 Heat Exchanger Secondary-to-Primary Leak Analysis (10/95) ............. 9-96 References for Chapter 9 ............. ........ .......... ..... 9-97

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I iv List of Ficutes Figure Page 2-1 Aerial View of Reactor Site and Immediate Vicinity (1967) . . . . . . . . . . . . . . . . . . . . 2-2 2-2 Aerial View of Reactor Site and Immediate Vicinity (1964) . . . . . . . . . . . . ....... 2-3 2-3 Contour Map of UVAR Site with Exclusion Fence ........... .............. 2-5 2-4 Map of Charlottesville and Vicinity . . . . . . . . . . . . . . . . . . . . . . . . . . . ..... .... 2-6  !

2-5 U Va. Research and Training Reactor Facility . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-8 2-6 First Floor Plan of UVAR Section of Building ............................. 2-9 2-7 Mezzanine Level of UVAR Section of Building . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-10 2-8 Ground Floor Plan of UVAR Section of Building . . . . . . . . . ................ 2-11 l 3-1 Cross-Section View of Reactor Pool and Reactor Room ...................... 3-2 3-2 8 x 8 UVAR Gridplate . . . . ........ ... ............................. 3-3 3-3 Top View of UVAR Standard and Partial LEU Fuel Element .... . ..... .... 3-5 3-4 Side View of UVAR Standard and Partial LEU Fuel Element ...... .. ... .. 3-6 3-5 Top View of UVAR Control Rod LEU Element . .. ...... ..... ......... 3-7 3-6 Side View of UVAR Control Rod LEU Element ......... ....... . . . . . . . . . 3-8 3-7 Nuclear Instrumentation System .. ... ... ...................... ..... 3-22 3-8 Source Range Drawer ........................... .................. 3-25 3-9 Intermediate Range Drawer . . . . . . . . . . .. .... .......... .. ...... . 3-28 3-10 Power Range Drawer ....... ........... ............. . . . . . . . . . . . . 3 -30 3-11 Scram Logic Drawer . ...... . .. ...... .......... .......... . . 3-32 3-12 High Power Trip ............. .. . . ........................ . 3-36 4-1 Cooling System Flow Diagram (revised 10/95) . . . . . . . . . .................... 4-4 4-2 Exhaust System to Stack . . . . . . . . . . . . . . . . . . . . . . ..... ..... .......... 4-13 4-3 Spray Header Mock-up ...... ... .... ....... . . . . . . . . . . . . . . . . . . . 4 -2 8 4-4 Core Spray System Elevation View ............ .. .......... . . . . . . . . . 4-29 4-5 Core Spray System Plan View . . . . . . . . . . . . . . ..........................4-30 5-1 Beam Hole Detail ...... ......................... ......... .. .... 5-2 5-la Top View of North Beamport Shielding and Access Control Walls ...... ...... .... ............. . . . ...... ....... 5-3 5-lb North Beamport Drain Fill System .......... ............... .. ..... ... 5-4 5-2 Large Access Facility . . . . . . . . . . . . . . . . . . . . ...... .... . ........ .. 5-5 5-3 UVAR Optimum Configuration ........ ...... ....... ... ........ . 5-8 6-1 Personnel Door .... .......... ....... .... ..... ........ ... .... 6-2 6-2 Exit Manhole . . . . . . . . . . . . . . . . ..... ..... ........... ... ....... 6-3 6-3 Pressure-Tight Air Duct .............................. . ... . . . 6-5 6-4 UVAR Confinement Room Count Rate Data .. . ............. ... .. .. 6-7 6-5 Thermal Neutron Fluxes ............ .. ... .... ... .. ............. 6-8 6-6 Fast-Neutron Dose Rates . . . . . . . . . . . ... ..... . . . .. .. . . . .... 6-9 6-7 Gamma Ray Dose Rates .... ........ .. ..... . . . . .... .. . 6-10 6-8 Dose Rates Near Surface of Pool . . . . . . . . . . . . . ... . . , . ... . . 6-12 8-1 Organizational Structure of U.Va. Reactor Facility .......... .. .. ......... 8-2

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9-1 H eat Flux Distributions . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-3 :

, 92 Cale, vs. Exp.t Incipient Boiling in the ORR ............................ ,. 9-7 l 9-3 Calculated and Experimental Burnout Heat Fluxes . . . . . . . . . . . . . . . . . . . . . . . . . 9-11 ,

9-4 Core Loading Configuration (showing peak flux location) . .. . . . . . . . . . . . . . . . . . . 9-22 95 Vertical Flux Traverse . . . . . ........................................ 9-23 .

9-6 LEU Core Power versus System (Core) Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-30 9-6a HEU Core Power versus System (Core) Flow . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . 9-37 9-7 Flow and Power Coastdown (Heauer Up) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-38 9-8 Flow Header Jammed in Cocked Position . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-4 2 - >

9-9 Flow Reversal After LOF Transient from 3.45 MW . . . . . . . . . . . . . . . . . . . . . . . . . 9-43 .t 9-10 Comparison Between OWR and UVAR Correlations . . . . . . . . . . . . . . . . . . . . . . . 9-53 9-11 Maximum Fuel Temperature After LOCA, LEU-22,4x4 . . . . . . . . . . . . . . . . . . . . . 9-60 1 '

9-12 Power of Hottest Element & ECCS Cooling After Shutdown . . . . . . . . . . . . . . . . . . 9-64 9-13 Prim a ry Pi pi n g . . . . . . . . . . . _. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ' 9-67 9-14 Geometry For Calculating V2 . . . . . . . . . . . ............................. 9-68 l

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, vi List of Tables Page Table  !

2.1 Hourly Wind Speeds ..............................................2-14 2.2 Hourly Wind Direction ........................... . . . . . . . . . . . . . . . . 2- 14 3.1 LEU-22 Reactor Data I. Typical 4x4 and 4x5 Core Parameters .........................3-17 II. Fuel Element Parameters .................................. 3-17 III. Fuel Plate Parameters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-18 i IV. Side Plate Parameters . . . . . . . . . . . . . . . . ....................3-18 V. Guide Plate Parameters ...................................3-18 i VI. Control Rod Parameters ...................................3-18 j

1. Safety (Shim) Rods . . . . . . .............................3-18

2. Regulating Rods . . . . . . . . . . . . . . .......................3-19 VII. Feedback Coefficients . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-19 Vill. Kinetic Parameters ........... . ........ ........... . . . . 3-19 l 3.2 Prompt Neutron Lifetimes for LEU and HEU Fueled UVAR Cores . . . . . . . . . . . . 3-20 4.1 Heat Exchanger Design Specifications (revised 10/95) ....................... 4-8 4.1.A Heat Exchanger Secondary Tube Plug Design Specification (10/95) . . . . . . . . . .. 4-8-A l 4.2 Cooling Tower Specifications .........................................4-9 ,

4.3 Secondary Pump Specifications . . . ............. .. .... . . . . . . . . . . . 4 - 10 j 4.4 Primary Pump Specifications ........................................4-11 '

4.5 Ar-41 Releases ..... ..................... .....................4-18 4.6 Radioactive Gases ................ ... ............... ........... 4-24 4.7 Iodine Concentrations ...... . .... ....... ....... ......... . . 4-25 l

5.4.1 Exclusion Radius Fractional Exposures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-14 5.4.2a Reactor Room-5 min. exposare .......................................5-15 j

. 5.4.2b Experimental-5 min. exposure .......................................5-16 i 9.1 LEU-22 Data And Parameters ..................... . ............... 9-27 9.2 CR Time to Drop from Predetermined Position ............ . . . . . . . . . . . . . 9-3 3 9.3 Maximum Peak Fuel Temperature Following LOCA . . . . . . ...... . . . . . . . . 9-5 9 9.4 Characteristics of Core Spray System .................................. 9-63 9.3 Time to Uncover Core for Various Leakage Mechanisms .............. .... 9-65 9.20.1 Calculated Primary Water Equilibrium Activity Concentrations (10/95) . . . . . . . . . . 9-78 9.20.2 Air Activity Concentrations Compared with Regulatory Limits (10/95) ......... 9-84 9.20.3 Water Activity Concentrations's Compared with Regulatory Limits (10/95) . .... 9-85 9.20.4 Air Activity Concentrations Compared with Regulatory Limits (10/95) ......... 9-91 9.20.5 Secondary Water Concentrations Compared with Regulatory Limits (10/95) . . . . . 9-93 ,

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Revised 10/95 4.6. Desien SDeCificationS The design specifications for the heat exchanger, cooling tower, secondary pump and primary pump are given in Tables 4.1 through 4.4. The data in these tables are updated as components are repaired, modified or replaced.

4.6.1 ReDiacement of Heat Exchanner System ComDonents When a heat exchanger system component, e.g., primary pump, secondary pump, heat exchanger, secondary tube plug, cooling tower, associated piping and valves, is to be repaired, modified, or replaced, a Title 10 Code of Federal Regulations 50.59 analysis shall be performed to determine suitability of the replacement equipment. Because only a limited number of system specifications are critical, most components can be replaced with a fairly wide range of substitutes. Consideration must always be given to material compatibility. For example, wetted parts of pumps must be aluminum-compatible. Also, care must be taken to maintain the required primary coolant flow rate. Finally, replacement components should be able to withstand the system pressures, with reasonably large excess margins.

It is unlikely that exact replacements of system components decades old can be located.

In the cases of the heat exchanger and cooling tower, the replacement component is likely to be of a different style, design, or type from the original. Such differences are acceptable once a "10 CFR 50.59 analysis" finds that the replacement is capable of performing its intended function and meets applicable 50.59 criteria. With that qualification, heat 4 A

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Revised 10/95 exchanger system component changes do not pose 10CFR50.59 "unreviewed safety questions". Changes of these components are permissible once a 10 CFR 50.59 analysis performed by the Reactor Staffis reviewed and approved by the Reactor Safety Committee.

4.7 Water Purification The pool water purity is maintained by circulating it at a rate of 20 gallons per minute through a carbon filter and a mixed-bed ion exchange demineralizer. The water is normally maintained at a pH of 6.0 to 7.0 with a conductivity ofless than 5 micromhos / cm.

4.8 Lieuid Waste Disposal System The Reactor Facility can collect radioactive liquid waste in two underground retention tanks of 5000 gallons each located outside of the Reactor Facility building, but within the site area. The waste is circulated and filtered, as well as given decay time before it is )

either discharged into the pond or discharged along with the pond as normal procedure. l l

Other storage tanks within the Reactor Facility may also be used to temporarily store I liquid waste. The option for sanitary sewer releases exists. All radioactive releases are made in conformance with applicable regulations. Two additional tanks of 250 gallons receive all waste from the Hot Cell. These tanks were installed as underground retentiori tanks in the l l

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Revised 10/95 TABLE 4.1 Heat Exchanger Design Specifications i Heat Transfer Rate: 6.83 E6 BTU /h (2 MW) l l

Materials: Aluminum 6061 Alloy. All materials must be compatible with aluminum. For this reason, no copper-containing alloys can be used.

Maximum length: 18 feet. l Number of Secondary Tubes: 712 Secondary Tube Dimensions: 5/8" O.D. with 18 Ga. wall thickness Fabricated in accordance with ASME Code,Section VIII, Division 1. l Inspected, certified, and stamped with the Code U-Symbol.

Hydraulic Specifications: Shell(Primary) Side Tube (Secondary) Side Fluid circulated High-purity water Cooling Tower Water l

Nominal Fluid flow rate 1100 gpm 1200 gpm l l

NominalInlet Temperature 110.2 Deg. F. 82.0 Deg. F  !

Nominal outlet Temperature 95.0 Deg. F 93.4 Deg. F Pressure Drop pump-dependent pump-dependent Design Pressure 50 psi 50 psi Test Pressure 75 psi 75 psi j i

Design Temperature 150 Deg. F. 150 Deg. F Inlet and Outlet Pipe Dia. 8 inch 8 inch 4-8

4 .1 4 6- n.- e ws- n J-- Jr u ~4 mi Revised 10/95 l TABLE 4.1.A Heat Exchanger Secondary Tube Plug Design Specifications "D*^2*

Plug Sketch: di o n 1/2 on Woshet 3, g., Wcsher i

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1/4"-20 , H M i bolt  ; i (- 5 / 8" -> 1/2" 2" y -

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i Primory seol Comptession Nut Screw, nuts, and interior washers may be stainless steel or aluminum.

Retaining washer must be aluminurn. No dissimilar metal shall be in contact with the heat exchanger tubes.

Expandable / Compressible Tubing Material: Norprene or equivalent rubber.

Shaft (Screw): 1 pcs.,1/4"-20 by 2".

Nuts: 2 pcs.,1/4"-20, I locking, I normal.

Washer 1: 3 pcs.,1/2 0.D., stainless steel between seals. .

Washer 2: 1 pcs.,5/8" O.D., aluminum only, plug positioninj; washer Installation: against tube sheet support only.

Maximum Number to be Installed: To be determined, based upon allowable heat I exchanger secondary-side working pressure, and secondary pump flow.

Testing and final installation torque: 12 inch-pounds.

Test Pressure: 150 psi (checked in a bench-rig).

Surveillance interval: annual, for in-situ visual inspection and installation torque check of  !

all installed plugs. i l

l 4-8-A l 1

I l

Revised 10/95 9.19 Heat Exchanner Secondary Tube Pluerine Analysis When heat exchanger secondary tubes are found to leak, they may be plugged with the type of replaceable expansion plug specified in TABLE 4.1.A. Tubes may be plugged up I F

to a maximum number at which the heat exchanger secondary maximum working ,

pressure is reached, as specified in TABLE 4.1. These pressures shall be monitored with 1 heat exchanger inlet and outlet pressure gauges shovn in Figure 4.1, following plug insertion. Secondary pump operating specifications.,in TABLE 4.3, should also be i consulted to set a reasonable lower limit on minimum steady-state flow through this l

pump.  ;

1 l

l Secondary tube plug installation does not affect primary coolant flow. This is because I

primary flow in the heat exchanger is through the shell side. The heat transfer rate is not of safety consideration. Reduction of cooling capacity resulting from tube plugging may affect the length of time the UVAR is operated on hot summer days until maximum pool water temperature (scram set-point at 105 Deg. F.) is reached.

9.19.1 Risk Associated with Plue failure The risk associated with tube plug failure is equal to the product of the following: [the probability of failure of either of the two plugs in a tube] and [the number of plugged tubes] and [the consequences of a failure]. Qualitative evaluation of these quantities is discussed in the following two sections. No significant mode of interaction between failed plugs and intact plugs is recognized. As explained below, the probability of single l

or multiple plug failures is negligibly small.

9 - 73

r I Revised 10/95.. '!

L 9.19.2 Probability of Plum Failure l_

l' Failure of a plug will occur when the plug fails to maintain separation of primary and

( secondary water at its installed location. The type of plug used in the UVAR heat exchanger is of design and materials that are the same as or comparable to those used by j the National Institute of Standards and Technology (NIST) to repair both their aluminum l l

and stainless steel tube and shell heat exchangers (for the NBS Reactor at NIST). Failure j of an installed plug is posited via a number of mechanisms:

1) degradation of rubber components because of N-16 gamma-ray j interactions;

2) loosening of torqued nut-bolt connections because of flow-induced vibrations;

3) galvanic corrosion of SS plug components,in association with contacting j materials; i

4) removal of the plug from installed location by pressure gradient.

Any of the above mechanisms could result in the plug coming loose and possibly being ejected from its tube. It is shown in this section that the mechanisms potentially leading to plug failure are all very unlikely.

Degradation of Norprene rubber components, due to irradiation by N-16 gamma-rays which originate in the primary side of the heat water exchanger, occurs at a very slow rate. NIST has operated such plugs for a period of four or five years without failure by this mechanism. A service life of this length can be expected in the UVAR heat exchanger. To test the durability of Norprene, the UVAR staff subjected a sample of the material to a Co-60 gamma-ray dose of 0.15 Mrad, which was 9 - 74

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

i

[-

Revised 10/95 l'

l. calculated to be equivalent to the dose that would be received by the material in an l

l installed plug during 1 year of continuous operation of the UVAR at 2 MW. No deterioration of material flexibility or other properties was observed as a result.

Therefore, the probability of plug failure during a surveillance interval of a year by this mechanism is essentially zero.

i Loosening of torqued nut and bolt connections, due to vibration and stress-relaxation cannot be ruled out completely. However, NIST has never experienced a plug failure  !

1 due to this mechanism. Precaution is taken against extreme loosening by use of SS l

threaded components, lock-nuts, and lock-washers. Regular yearly surveillance over tightness should be adequate to ensure that the probability of failure by this mechanism is l essentially zero.

1 Galvanic corrosion, arising from direct contact of stainless steel plug components and the aluminum heat exchanger would occur at a negligible rate. This is because the corrosion potential between aluminum and stainless steel is small. However, plug design ensures that dissimilar metals will not come into contact with the tubes. Periodic surveillance is adequate to ensure that the probability of mechanical failure due to corrosion is negligible.

Removal of a plug from its installed location by a pressure gradient during reactor operations can be ruled out easily by considering the maximum service pressure differential from primary-to-secondary. This pressure gradient should be 9 - 75

Revised 10/95 about 15 psi., whereas plugs are tested before installation at 150 psi. This ten-fold margin, coupled with the design-feature plug positioning washer, is ample to ensure that the probability of plugs being removed from their installed locations by normal water pressure is essentially zero.

9.19.3 Conclusion Renardine Risk of Plue Failure l

The consequence of a tube plug failure is the slow recreation of a pre-existing leak, a

'l possibility which is included in the safety analysis envelope in the subsections of Section 9.20. Since it is shown above why there is essentially a zero probability that a l plug will fail, it is concluded that the risk associated with plug failure is also l

approximately zero.

l i

l 9 - 76

L . Revised 10/95 l

9.20 Heat Exchanner Primary-to-SecondaryLeak Analysis L

9.20.1 Introduction The worst-credible primary-to-secondary heat exchanger tube leak rate that could develop before discovery is 1 ml/sec (about 1 gph) and would be caused by pitting corrosion.

This type of leak starts small, grows slowly, and resultant secondary water activity is detectable before a 1 ml/sec leak rate is reached. The following scenario begins with conservative assumptions being made about reactor system configuration and operation.

Progress of the scenario is followed to determine the significance,if any, of environmental releases of diluted pool water. The basis for this leak is presented first.

Finally, it is shown clearly that the worst credible primary-to-secondary leak rate will not violate air or water effluent release limits.

9.20.2 Basis for Primary to-Secondary Leak Rate In August of 1995, a UVAR heat exchanger leak was observed by the reactor staff following cleaning of secondary tubes. The magnitude of that leak was determined using the activity of sodium-24 measured in primary and secondary water samples. Applying corrections to these data for the operating history of the reactor and secondary water blowdown (water draining from the secondary to the sewers) rate, a leak rate of 1 ml/sec was calculated. This leak rate resulted in sodium-24 activity in the secondary water equal to a factor of eight times the minimum detectable activity in cooling tower water.

Thus, a smaller leak could be detected.

9 - 77

Revised 10/95 Primary water gamma-ray spectroscopy data provided a basis for calculating the equilibrium activity concentrations ofisotopes normally expected in pool water as a consequence of operation. These concentrations are given in TABLE 9.20.1.

TABLE 9.20.1 Calculated Primary Water Eauilibrium Activity Concentrations Nuclide Equil. Conc. Rad. Decay Const. (A)

Activity IuCi/ml..ter] I 1/ see 1 H-3 4.0 E-4 1.79 E-9 Na-24 2.5 E-3 1.28 E-5 Mg-27 5.0 E-5 1.22 E-3 Cl-38 1.5 E-5 3.10 E-4 Mn-54 1.0 E-5 2.57 E-8 Cr-51 5.0 E-5 2.89 E-7 Sb-122 5.0 E-6 1.22 E-3 W-187 5.0 E-6 8.05 E-6 9.20.3 Conditions Prevailine at Start of Scenario The UVAR is assumed to have been in operation at 2MW for a long time, compared to the radionuclide half-lives ofinterest. The primary and secondary pumps and cooling tower fan are operating normally. Pool water contains saturation levels of typical radionuclides.

9.20.4 Initiatine Scenario Event A 1 ml/sec primary-to-secondary leak is assumed to begin instantaneously in the heat exchanger through a pin-hole leak in a secondary tube. Assumption of sudden leak 9 -78 l

l 1

1 Revised 10/95 l

l 1

initiation is conservative. For simplicity and conservatism, it is also assumed that l equilibrium cooling tower water activity concentrations are reached instantaneously upon initiation of the leak. Thus, steady-state releases of radioactive material begin from the cooling tower at a constant rate.

l 9.20.5 Primary-to-Secondary Leak Proeression It is assumed that the reactor staff does not notice the leak from observations of pool level, although a leak of this magnitude (about 24 gal / day) is detectable by that method. )

Therefore, the leak is assumed to continue until discovered when the next cooling tower water sample is analyzed and sodium or other nuclides are detected. A leak will be 1

considered identified if the concentration of sodium-24 in cooling tower water exceeds 1 i E-6 pCi/ml. Leak detection could occur as long as ten days after leak inception, since cooling tower water samples are analyzed " weekly".

9.20.6 Mitimation of Heat Exchanner Leak Connauences Soon after a leak is identified, the cooling tower fan would be stopped. The UVAR would be shutdown, and primary isolation valves would be closed. Cooling tower water l l

blowdown would be disabled. These actions would effectively stop further addition of radionuclides to the secondary water and end further release of radioactive material to the air and sanitary sewer. Water remaining in the secondary system could then be disposed 1

of appropriately, 9 -79

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

Revised 10/95 9.20.7 Calculated Release Rates Equilibrium activities for normally observed radionuclides (TABLE 9.20.1) were I-L ' determined from primary water analysis and corresponding operating history. These b

l activities, along with the observed leak rate from August 1995, were used as the source y

i term for possible airborne, water, and sewer release activity concentration calculations -

and comparison to regulatory limits for each release mode.

Two bounding cases are considered:

Case 1: All activity in the primary water leaking into the secondary is postulated to become airborne instantaneously in the cooling tower exhaust.

Case 2: All activity in the primary water leaking into the secondary becomes concentrated in the secondary water through evaporation of some secondary water in the cooling tower, This water is then assumed to be released to the sanitary sewer (the normal blowdown path), or to the environment (see below).

Since the source term is fixed, the release from each of the two paths will have lower radionuclide concentrations than the releases calculated by assuming only one pathway at a time. An actual release will normally be a combination of air and sewer releases.

When the secondary pump stops, water in the upper basins flows down to the lower basin, which does not have sufficient capacity to contain about 1000 gallons of excess water. Thus, a direct release of cooling tower water to the roof of the building, and from there to the pond, is possible.

l l

l 9 - 80

i Revised 10/95

" Concentration Ratio", Cr [unitiess],is defm' ed as a measure of the degree of accumulation of a given element in cooling tower water due to cooling tower operation and blowdown. The ratio is calculated by dividing the concentration of an element in cooling tower water by the concentration of the element in make-up water. This is typically determined by using the concentrations of calcium, since this element is ubiquitous.

Equation 9.20.1. Cr[unitless] = C, / C.

where C = Element concentration in secondary water [ atoms /ml],

and C. = Element concentration in make-up water [ atoms /ml].

When the make-up rate M [gph] is known, then the blowdown rate B [gph] required to obtain a given concentration ratio can be calculated from the following relationship:

Equation 9.20.2 Cr[unitless] = M / B where M = Secondary water make-up rate [gph],

and B = Secondary water blowdown rate [gph].

Typically, concentration ratios for stable elements are controlled to be in the range of l

from 6 to 7 by a blowdown system that automatically sends secondary water to the sanitary sewer when secondary water conductivity reaches a pre-set level. It is assumed that the concentration ratio that exists for stable elements is also applicable to 1 ..

radionuclides. To be conservative, the isotope concentration ratio is assumed to be equal l to 10 in the Case 2 Calculations. ,

I 9 - 81 l

Revised 10/95 9.20.7.1 Calculation of Aunust 1995 Heat Exchanner Leak Rate The fractional primary-to-secondary leak rate in August 1995 is calculated from Equation 9.20.3, below, using the pmdominant radioisotope in primary water, sodium-24. First, l secondary water blowdown rate is assumed to have been 300 gph. Given that the secondary system contains 2000 gallons, this blowdown rate value leads to a fractional blowdown rate, Br,4.2 E-5 [l/sec], as noted below.

Equation 9.20.3 L = (A + B,)

• A." / Ap " [1/sec],

where:

l L = Primary leak rate divided by secondary volume [1/sec];

A = Decay constant of sodium-24 [1/sec];

B, = Fractional blowdown rate from secondary to sewer per unit time [1/sec];

Note: Br [1/sec] = B [gph] / V, [ gal] / (3600 sec / h),

where B is assumed = 300 [gph] (typical),

and V = 2000 gallons of secondary water volume.

A " = Calculated equilibrium sodium-24 activity concentration in secondary water

[pCi/ml..t,,] , based on August 1995 secondary water samples; A p' = Calculated equilibrium sodium-24 activity concentration in primary water

[pCi/ml.. ,,], based on August 1995 primary water samples.

Using appropriate values:

L = 1.28 E-5 [1/sec];

B, = 4.2 E-5 [l/sec];

A p' = 2.5 E-3 [pCi/ml..ier];

A," = 6.1 E-6 [pCi/ml.,,er],

9 -82

Revised 10/95 the fractional leak rate, L, is found to be 1.34 E-7 [l/sec]. Finally, multiplying this rate by the secondary volume,2000 gal, the volumetric leak rate is found to be 1 ml/sec (about I gph).

Oace the fractional leak rate is known, Equation 9.20.3 can be arranged to solve for A." .

Steady-state activity concentrations in cooling tower water for the other nuclides with source terms listed in Table 9.20.1 are found in this way. The resultant concentrations are listed in Column 2 (Secondary Water Activity [pCi/mlwer]) of TABLE 9.20.3.

Temporary ratios presented for nuclides in this table are valid foc the release duration and are equal to their secondary water activity concentration divided by the water and sewer release limits, respectively.

I 9.20.7.2 Airborne Release Calculation (Case D Airborne release of radionuclides from secondary water in the cooling tower will be J

generated and diluted by the forced air flow from the cooling tower fan. The flow rate of this fan is 5 E+7 ml/s. For this analysis, it is assumed that all activity leaking from the primary to the secondary becomes instantly airborne. Table 9.20.2 gives the airborne i

concentrations based upon this scenario, assuming the TABLE 9.20.1 activity concentrations for activation radionuclides normally present in the primary water as the source term. A ratio equal to the derived airborne concentration divided by the applicable limit given in TABLE 9.20.2 is then calculated. The sum of these temporary ratios is calculated and found to be much less than 1, assuring that regulatory release limits will not be violated by the postulated air release.

9 - 83 L

Revised 10/95 TABLE 9.20.2 Air Activity Concentrations Compared with Regulatory Limits l

Yeariv-Average TemDorary Airborne Release Ratio i 4 ADD.B Table 2 Air Activity Col.1 Limits [(Air Act. Conc.) /

Nuclide IuCi/mlel fuCi/ mig] (Air Conc. Limit)]  !

H-3 8 E-12 1 E-7 8 E-5 l Na-24 5 E-11 7 E-9 7 E-3 i Mg-27 1 E-12 1 E-9 1 E-3 Cl-38 3 E-13 6 E-8 5 E-6

! Mn-54 2 E-13 1 E-9 2 E-4 l Cr-51 1 E-12 3 E-8 3 E-5 Sb-122 1 E-13 3 E-9 3 E-5 W-187 1 E-12 1 E-8 1 E-4 Sum of Temporary Ratios 0.008 l

9.20.7.3 Water-Borne Release Calculations (Case 2)

Radionuclides in the cooling tower basin will be concentrated as water is evaporated.

Normally, the concentration ratio for elements is controlled by blowdown to be between 6 and 7. To be conservative, a concentration ratio of 10 is chosen for these calculations.

Radioactive decay of the shorter-lived radionuclides limits the maximum activity that can be obtained, independent of the concentration ratio in cooling tower water. The formula used to determine secondary water activity concentrations is Equation 9.20.3, rearranged l l

to solve for those concentrations. As explained above, with a concentration ratio of 10 and a normal average make-up of 800 gph, B, becomes 1.1 E-5 [l/sec].

I i

9 - 84

l Revised 10/95 TABLE 9.20.3 Water Activity Cone's Compared with Regulatory Limits Yeariv-Average Yeariv-Average l Secondary Water Release Sewer Release I Water ADD. H Table 2 ADD. H Table 3 Activity Col. 2 Limits Limits j Nuclide fuCi/ml .ierl IuCi/mi,,,ter] Ratio IuCi/ml. terl Ratio l

H-3 5 E-6 1 E-3 0.005 1 E-2 0.0005 Na-24 1.5 E-5 5 E-5 0.300 5 E-4 0.03 Mg-27 5 E-9 -- -- - -

l Cl-38 6 E-9 -- -- -- -

I Mn-54 1 E-7 3 E-5 0.003 3 E-4 0.00033 Cr-51 6 E-7 5 E-4 0.001 5 E-3 0.00012 i Sb-122 5 E-8 1 E-5 0.005 1 E-4 0.0005 j W-187 4 E-7 3 E-5 0.013 3 E-4 0.0013 i

! l Sum of Temporary Ratios 0.327 0.033 Because sodium-24 is the dominant and most limiting radionuclide in the primary water, i

an additional calculation was done by assuming no loss of sodium-24 by blowdown to 1 l

show that the concentration ratio (effectively infinite in this case) is not a critical l parameter in these calculations. With no loss of sodium by blowdown, the equilibrium i

sodium concentration would increase from 1.5 E-5 [pCi/ml.a.r] to 2.6 E-5 [pCi/mlwaerl.

The temporary ratio for sodium effluent would increase from 0.300 to 0.530, making the i 1

corresponding sum of temporary effluent ratios increase from 0.327 to 0.557. The temporary ratio for sodium sewerage would increase from 0.03 to 0.052, making the corresponding sum of temporary sewerage ratios increase from 0.033 to 0.055. The result remains that the postulated water releases will not violate regulatory release limits. l l

9 -85

Revised 10/95 9.20.7.4 Release Analyses Conclusions The primary-to-secondary leak was readily identified in August 1995 by means of radio-analysis of the secondary water at an activity level of about 1 E-6 Ci/ml. Any future leak should be observed easily before the leak grows to the size of the observed leak of August 1995. Sodium-24 is the most observable and most limiting radionuclide in the secondary water and can be detected through regular sampling of secondary water before release limits are reached. Radionuclides not on the list of those analyzed (TABLES 9.20.1,2 &

3) will be present at such low levels compared to the sodium-24 that they do not need to be individually considered.

I l

9 - 86

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

l l

Revised 10/95 9.20.8 Double-Ended-Heat-Exchanner-Tube-Break Leak Analysis 9.20.8.1 Introduction i

i This primary-to-secondary leak analysis is predicated upon the occurrence of.a double-l ended-guillotine-tube-break (DEGTB). Such a leak-initiating event is an applicable, '

conservative assumption for power plant safety analyses because high flow rates, pressures, temperatures and possibly flow-induced vibration occur in power-plant steam generators. In contrast, operating conditions in the UVAR heat exchanger are quite I l

gentle. In Sections 9.20 to 9.20.7.4, the leak analysis is based upon credible assumptions, j I

conditions, and industry experience (regarding leaks in heat exchangers at low l l

temperatures, pressures, and flow-rates). The analysis in this section, predicated upon a totally hypothetical model, is used to fathom the maximum impact from heat exchanger leaks on public health and safety.

In the UVAR heat exchanger, primary water flows through the shell. If a large amount of primary water were to enter the UVAR secondary system, potential cooling tower water l

overflow would be directed to a 750,000-gallon on-site effluent-hold-up pond. The pond provides dilution and decay time for radionuclides, making unplanned release to I i'

unrestricted areas in excess of regulatory limits extremely unlikely. It is emphasized that a primary-to-secondary UVAR leak would only be possible when the primary pump is on.

Since pump status is administratively controlled, the primary pump only would be on when staffis on-site.

9 - 87

l Revised 10/95 9.20.8.2 Conditions Prevalline at Start of Seanario l

The most challenging conditions would exist with staff on-site, and the UVAR just  !

shutdown following an operating time sufficiently long for equilibrium radionuclide activity concentrations to exist in pool water. These concentrations, based upon August 1995 reactor operations data, and corresponding radioactive decay constants, are given in i TABLE 9.20.1. With the UVARshutdown, the primary pump, secondary pump, and cooling tower fan could be left, and for the analysis are left, operating to cool down the i

pool. Primary pool level is assumed at 19'6" above the core, at its highest level short of overflowing. With the OVAR shutdown, the low-pool-level alarm is actuated by pool level dropping to 19'2-1/2", which is lower than the 19' 3-1/2" actuation level of the other low-pool-level alarm (and scram) enabled during reactor operation.  ;

9.20.8.3 Initiatine Scenario Event It is posited that a DEGTB occurs in the heat exchanger. The leak rate must be estimated I i

without having some necessary data in hand, for example, the driving pressure across the break, flow friction factors and the exact primary flow path around baffles. It is sufficient to make conservative assumptions. In analogy with the North Carolina State University ,

(NCSU) DEGTB analysis, it is assumed that the primary flow is through the hest b

exchanger tubes, rather than the shell. Accordingly, the leak rate is calculated at 5.8 gpm and conservatively rounded up to 10 gpm (600 gph). This leak rate is 600 times larger ,

than the corrosion-initiated leak rate postulated above. It is shown below that, with a large leak, the actual leak rate value is not a critical parameter. Rather, the bounding parameter is the total primary water radionuclide activity released, which can be fairly accurately predicted.

9 - 88

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

Revised 10/95 9.20.8.4 Primary-to-Secondary Leak Proeression and Mitination ,

i Since a reactor staff member is required to be present at the facility when the primary )

l pump is operating, the 600 gph leak could continue until primary pool level drops from 19'-6" to the 19'2-1/2" level, at which point the low-pool-level alarm would be activated. l There are 200 gallons of primary water per inch of pool water level. Therefore, it is assumed the water level lowers by 3-1/2" (from 19'6" to 19'2-1/2"), and this results in a transfer of 700 gallons of primary water to the secondary side. It is very likely, however, that the actual amount of primary water leaked would be less than 700 gallons because reactor pool level does not routinely approach 19'6".

Discovery of the assumed 10 gpm leak will occur after no more than about 1.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. At I

that time the low-pool-level alarm would sound in the reactor control room and the l facility hallway. Reactor operator (s) would respond by stopping the primary pump and l cooling tower fan. These actions stop the primary-to-secondary leak and cooling tower exhaust release immediately. The primary system may also be isolated by closing the pool isolation valves. Contaminated secondary water in the cooling tower would then be disposed ofin an appropriate manner.

9.20.8.5 Scenario Consecuences Radionuclide concentrations in cooling tower exhaust (air) and secondary system water are calculated in the manner presented in the subsections of Section 9.20.7. Again, sodium-24 data are used in illustrative calculations. Results pertaining to this and the other radionuclides listed in TABLE 9.20.1 are ebtained in the same two ways and have been listed in TABLES 9.20.4 & 5. Equilibrium primary water radionuclide 9 - 89

1 Revised 10/93 concentrations and decay constants for the calculations may be obtained from TABLE 9.20.1. The same two bounding cases discussed in Section 9.20.7 are applicable, namely:

Case 1: All activity in the primary water leaking into the secondary is postulated to become airborne instantaneously in the cooling tower exhaust.

Case 2: All activity in the primary water leaking into the secondary becomes concentrated in the secondary water through evaporation of some secondary water in the cooling tower. This water is then assumed to be released to the sanitary sewer (the normal blowdown path), or to the environment (see below).

9.20.8.6 Airborne Release Calculation (Case 1)

The rate of sodium activity in-flow to secondary water is (10 gpm) * (3800 ml/ gal) * (1 min / 60 sec) * (2.5 E-3 pCi/ml. r) i

= 1.6 [ Ci/s].

This activity in-flow is diluted by cooling tower air-flow at a rate of 5 E+7 ml/s, resulting in a sodium activity concentration in air of(1.6 pCi/s) / (5 E+7 ml/s)

= 3.2 E-8 [ Ci/ml.ir].

For all other equilibrium radionuclides listed, it follows that the appropriate multiplicative factor for conversion of primary water concentration to secondary air ,

1 concentration is (10 gpm) * (3800 ml/ gal) * (1 min / 60 sec) / (5 E+7 ml/s)

= 1.3 E-5 [ Ci/ml.ir / pCi/ml..ar]. )1 i

l l

I 9 - 90

i Revised 10/95

]

TABLE 9.20.4 Air Activity Concentrations Compared with Renulatory Limits '

Yearly Averane TemDorary l Airborne Release Ratio App.B Table 2 Air Activity Col.1 Limits [(Air Act. Conc.) /

Nuclide luCi/ml.!,-] IuCi/ml.i,] (Air Conc. Limit)] ]

l H-3 5.1 E-9 1 E-7 0.051 Na-24 3.2 E-8 7 E-9 4.57 Mg-27 6.3 E-10 1 E -9 0.600 )

Cl-38 1.9 E-10 6 E-8 0.003 Mn-54 1.3 E-10 1 E-9 0.130 Cr-51 6.3 E-10 3 E-8 0.021 Sb-122 6.3 E-11 3 E-9 0.021 l W-187 6.3 E-11 1 E-8 0.006  !

Sum of Temporary Ratios 5.4 The sum of ratios in TABLE 9.20.4 indicates that radionuclide concentrations in cooling tower exhaust durine the nostulated 1.2-hour atmospheric release are temporarily above ,

l the regulatory limit of 1, which is linked with a continuous year-lone release. This sum of temporary ratios is a valid at the cooling tower, which is about 25 meters from the site boundary. For the above calculation, no credit was taken for dilution of air release from j l

l the cooling tower to the site boundary. Dilution of the air release into the atmosphere I begins immediately, with the cooling tower exhaust expelled forcefully upward.

Diffusion of radionuclide content occurs in all directions as the plume travels to the site boundary. Intuitively, even during the release time, there will be a dilution factor of at least 10 to bring the sum of ratios to a number less than 1 at the site boundary, about 25 meters away from the cooling tower. If this dilution is ignored, it would take only about 6 additional hours of cooling tower exhaust with no measurable radionuclide content to ensure that the calendar-year-9 -91

Revised 10/95 averaged sum of ratios is reduced to less than 1. Hence, the postulated Case 1 analysis concludes that the regulatory limit pertaining to air releases will not be exceeded.

9.20.8.7 Water-Borne Release Calculations (Cafe 2)

For a DEGTB, radionuclide concentrations in secondary water are calculated using Equation 9.20.5 to evaluate A,". This equation is valid in conjunction with the conservative assumption that steady-state secondary water radionuclide concentrations -

are reached instantaneously.

Equation 9.20.5 A," = L

• Ap" / (A + B,) [pCi / ml..ar],

where, following the definitions of terms given in Section 9.20.7.1, conservative variable values are:

L = (10 gpm) * (1 min / 60 sec) / (2000 gal) = [8.3 E-5 /sec];

Ap~ = 2.5 E-3 [pCi/ml.awr] (TABLE 9.20.1);

A = Decay constant of sodium-24 = 1.28 E-5 [1/sec] (TABLE 9.20.1);

B, = 4.2 E-5 [1/sec] (typical value, again assuming a concentration factor of 10).

Substitution of the above values results in secondary water sodium-24 activity concentration of A," = 3.8 E-3 [pCi/ml..ar]. Note that this activity concentration is greater than Ap" because of concentration of sodium in the cooling secondary system.

There is no universal multiplicative factor for conversion of Ap" to A " for the other listed radionuclides because the decay constant for each isotope must be incorporated from TABLE 9.20.1 in the calculation of A,~ . The calculation is straight-forward, 9 - 92

Revised 10/95 however, and the resulting Al* for the other listed radionuclides are presented here in TABLE 9.20.5. Results of sum-of-ratios calculations for the water- and sewer-release sub-cases are also shown in TABLE 9.20.5.

TABLE 9.20.5 Secondary Water Activity Cone's Compared with Regulatory Limits Yearly-Aversee Yeariv-Averau Secondary Ef11uent Release Sewer Release Water App. B Table 2 App. B Table 3 Activity Col. 2 Limits Limits Nuclide IuCi/ml..ierl IuCi/ml.. .r] Ratio IuCi/ml..ierl Ratio H-3 2.0 E-4 1 E-3 0.20 1 E-2 0.02 Na-24 3.8 E-3 5 E-5 76 5 E-4 7.6 Mg-27 3.3 E-6 -- - -- -

Cl-38 3.6 E-6 -- -- - --

Mn-54 2.0 E-5 3 E-5 0.67 3 E-4 0.07 Cr-51 1.0 E-4 5 E-4 0.20 5 E-3 0.02 Sb-122 3.3 E-7 1 E-5 0.03 1 E-4 0.003 W-187 8.3 E-6 3 E-5 0.28 3 E-4 0.028 Sum of Temporary Ratios 77 7.7 The sum of temporary ratios, in the case of the water-only effluent release pathway is significantly larger than 1. However, there is no water effluent released to the environment because it is mostly retained in the secondary coolant [ Note: Secondary make-up shuts down automatically due to the fast leak rate which tends to improve (lower) the conductivity of water in the cooling tower.].

At worst, some cooling tower water could overflow to the on-site reactor pond, also used i for liquid waste dilution. The pond dilution factor is about 750,000 gallons / 2000 gallons, or 375-to-l. Only a fraction of the entire 2000 gallons of secondary water could overflow to the pond. Natural radionuclide concentration in the pond is negligible.

9 - 93

Revised 10/95

, With pond-water dilution, the sum of temporary ratios for water effluent is reduced to 77 / 375 = 0.21, well below the allowed calendar-year-average limit of 1. It is concluded that with a DEGTB and a water-only effluent pathway, the calendar-year-average water effluent limit would not be exceeded. Pond water is only released periodically, subsequent to radioanalysis. In the case of this event it would most likely be possible to delay a pond release long enough to achieve significant decay of the 15-hour sodium-24. ,

The sum of temporary ratios for the sanitary sewer release is marginally greater than I during a short period of time (about I hour). A great volume of cooling tower water with no measurable radionuclide content is released to the sewer by blowdown during any calendar year. This will ensure that the calendar-year-average sum of ratios will be less than 1. Thus,it is concluded for this case that sewer release regulations will not be l

exceeded. l l

9.20.8.8 Conclusions i l

The primary-to-secondary heat exchanger leak scenario involving the DEGTB analyzed above is extremely conservative. The posited 600 gph leak rate and staff discovery time of 1.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or less represents a hypothetical situation postulated for examination of the  !

potential maximum consequences to the public. Calculated radionuclide concentrations for limiting air-only and water-only release cases are sufficiently low and dilution sufficiently high that calendar-year-average release limits are not challenged. Therefore, it is concluded that even an extremely severe (hypothetical) heat exchanger tube rupture poses no significant threat to public health and safety.

9 - 94

- . - _ . - . . . .. ....-...__ ...-- . .- .. = . . . . . . -

Revised 10/95 The analysis just performed reveals that the leak rate, per se, is not a critical parameter.

While radionuclide concentrations in a release depend strongly upon the primary-to-secondary leak rate, this is a temporary situation. On a calendar-year-averaged basis, it is i

the total activity of each isotope released, i.e., the volume of primary water lost to the l environment, that is significant.

?

s r

0 9 - 95 i

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

4 Revised 10/95 9.21 Heat Exchanner Secondary-to Primary Leak Analysis During times when the primary pump is off, secondary water may pass through a leaking heat exchanger tube into the primary water system. This is possible because of the elevation difference between the water surface in the cooling tower basin and the reactor

,. pool surface. If the secondary pump is running, an enhanced secondary-to-primary leak rate occurs due to increased secondary-side pressure. ,

Abundant industry operating experience with leaks caused by corrosion of tube and shell l

heat exchangers shows that they start out at a small flow rate and then increase gradually with time. Therefore, secondary water leaking into the primary water will degrade pool water quality gradually. Overall water quality with the leak rate basis assumed in Section

! 9.20 (i.e.,1 ml/sec) can be kept easily within technical specification limits. There is a I

demineralizer system in continuous operation which mitigates the effect of secondary-to-primary leakage. It is concluded that pool water quality is monitored on a frequency (at a minimum, weekly) that is sufficient to identify leaks well before they pose water quality concerns.

l 9 - 96

~

T l 1

Revised 10/95 REFERENCES FOR CHAPTER 9 l r

1, Gambill, W.R., " Generalized Prediction of Burnout Heat Flux for Flowing, Subcooled, Wetting Liquids," Chemical . Engineering Progress Symposium Series, Vol. 59, Houston,1%2.

{

2. Bergles, A. and Rohsenow, W., "The Determination of Forced- Convection Surface  ;

Boiling Heat Transfer," ASME Paper 63-HT-22 (1963).

3. Cox, J., "ORR Operations for Period April 1958 to April 1959," CR-59-39, Oak Ridge -

National Laboratory (April 1959).  ;

4. Eckert, E.R.G. and Drake, R.M., " Heat and Mass Transfer." McGraw-Hill, New York, 1959. I

5. Private Communication from Wallace R. Gambill, ORNL.

6. Bemath, L., " Theory of Imal Boiling Burnout and its Application to Existing Data,"  ;

Ch.E. Progress Symposium Series, Storrs,1960.  :

7. Todeas, N., "Effect of Non-Uniform Axial Heat Flux Distribution on the Critical Heat f Flux," Ph.D.. Dissertation, M.I.T., Sept.1965.

8. DeBortoli, R.A., " Forced-Convection Heat Transfer Bumout Studies for Water in  ;

Rectangular Channels and Round Tubes at Pressures Above 500 psia," WAPD-188,  !

Oct.1958. ,

I

9. Gambrill, W.R., " Forced Convection Burnout," Nuclear Safety, Vol. 5, No. 2, Winter 1963-64.

10. Tong, L.S., "DNB Studies in an Open lattice Core," WCAP-3736, Aug.1964.

I 1. Perry, J.H. (ed), " Chemical Engineer's Handbook," McGraw-Hill, New York,1950.

12. Esselman, W.H., et al, " Thermal and Hydraulic Experiments for Pressurized Water Reactors," Proceedings of the 2nd United Nations International Conference on the Peaceful Uses of Atomic Energy, Vol. 7 pg. 758, United Nations, Geneva,1958.

13. Von Karman, T., NACA Technical Memo No. 611,1931.

14. Gambill, W.R. and Bundy, R., "HFIR Studies of Turbulent Water Flow in Thin Rectangular Channels," ORNL-3079, June 1961. ,

15. Reynolds, J., " Local Boiling Pressure Drops," ANL-5178, March,1954.

f 9 - 97  ;

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

1 Revised 10/95 i,

' 6. Bamett, P.G., "The Prediction of Burnout in Non-Uniformly Heated Rod Clusters from Burnout Data for Uniformly Heated Round Tubes," AEEW-R-362, November 1964. '

17. "Non-Uniform Heat Generation Experimental Program," Quanerly Progmss Repon No. ,

5, July-September 1964, BAW-3238-5, Sept.1964.

18. "Non-Uniform Heat Generation Experimental Program," Quanerly Progress Report No.

6, October-December 1964, BAW-3238-6, December 1964.

19. Sternberg, H.I., " Thermal Power Calibration and Correlation of UVAR by Foil Irradiation and Heat Balance," Master's Thesis, University of Virginia, June 1964.

20. Bmnot, W.K., "An Analysis of Fuel Element Surface Temperatum and Coolant Flow Rate in UVAR," Master's Thesis, University of Virginia, November,1%1.

21. Custer, G.A., "The Experimental Determination of the Thermal Neutron Flux in the University of Virginia Reactor Using an Aluminum Hydraulic Rabbit, Master's Thesis, University of Virginia, August 1960. t

22. Status Report on the Omega West Reactor, with Revised Safety Analysis," Ims Alamos Scientific Laboratory, LA-4192, July 1969.

23. DiNunno, J.J., et al. " Calculation of the Distance Factors for Power and Test Reactor [

Sites," TID-14844,1%2.  !

24. " University of Virginia Reactor Safety Analysis Repon," University of Virginia, UVAR-18, Revised January 1979,

25. Dahlheimer, J.A., " Thermal-Hydraulic Safety Analysis of Pool Reactors", Master's Thesis, University of Virginia,1967.

26. University of Michigan, "2DB-UM Appollo Version", Version #10, September 1986.

27. Freeman, D.L., "Neutronic Analysis for the UVAR HEU to LEU Conversion Project",

Master's Thesis, University of Virginia, January 1990.

28. Smith R.S. and Woodruff W.L., "A Computer Code, NATCON, for the Analyses of Steady-State Thermal-Hydraulics and Safety Margins in Plate-Type Research Reactors y Cooled by Natural Convection", ALN/RERTR/I'M-12, December 1988.

l.

29. " Technical Specifications for the University of Virginia", September 1982, Amended April 1988, and December 1988.

30.Telecopy from Argonne National Laboratory, " Properties of Fuel Meat Materials",

August 19,1988.

l 9 - 98 L

_ _ ~,

Revised 10/95

31. " Decay Heat Power in Light Water Reactors" American National Standard ANSI /ANS-5.1-1979, Reaffirmed July 1985, American Nuclear Society, La Grange 1979.

32. Meem, J. L, " Emergency Core Spray System Installation And Testing", Report to University of Virginia Reactor Safety Committee, September,1971.

33. Nagler, A., Gilat, J., Hirshfeld, H., " Evaluation of LOCA in a Swimming-Pool Type Reactor Using the 3D-AIRLOCA Code," paper presented at the XII International Meeting on Reduced Enrichment for Research and Test Reactors, Berlin, Federal Republic of Germany

34. ". Heat Exchanger Pressure Boundary Breach", Proposed NCSU SAR Section 13.2.1.5, pp.13-19 thru 13-23, by FAX transmittal,9/12/95.

9 - 99