ML20091E278
| ML20091E278 | |
| Person / Time | |
|---|---|
| Site: | 05000124 |
| Issue date: | 05/23/1984 |
| From: | Ellis E, Fleming H, Florian R VIRGINIA POLYTECHNIC INSTITUTE & STATE UNIV., BLACKSB |
| To: | |
| Shared Package | |
| ML20091E272 | List: |
| References | |
| NUDOCS 8406010142 | |
| Download: ML20091E278 (361) | |
Text
{{#Wiki_filter:.. .. _ _ SAFETY ANALYSIS REPORT FOR THE 4 VIRGINIA TECH ARGONAUT REACTOR Department of Mechanical Engineering . College of Engineering Virginia Polytechnic Institute and State University Blacksburg, Virginia 24061 Written, Compiled and Edited by: E. R. Ellis, Staff SRO, VTAR T. F. Parkinson, Director, VTAR H. M. Fleming E. L. Seth R. J. Florian T. S. Smithwick I P. D. Holian, Ex-Supervisor, VTAR J. D. Vassar, Jr. D. R. Krause, Supervisor, VIAR l Drawings *by Robert H. Moore Typed by Sandy Hubbard 4 O C o!h24 PDR
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TABLE OF CONTENTS - VTAR Safety Analysis Report Page Table of Contents................................................... 11 List of Figures.....................................................xvi List of Tab 1es..................................................... xvii List of Abbreviations................................................xx Chapter 1 GENERAL DESCRIPTION OF FACILITY 1.1 Introduction.................................................l 1.2 Gene ral De s c ri p t io n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.3 Comp a ri s o n Tab 1e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-1.3.1 Comparison with Similar Faciliry Designs. . . . . . . . . . . . . 8 1.3.2 Comparison of Final and Preliminary Information...... 8 1.4 Identification of Agents and Contractors. . . . . . . . . . . . . . . . . . . 12
- 1. 5 Requirements for Further Technical Inf ormation. . . . . . . . . . . . . 12 1.6 Material Incorporated by Reference......................... 12
- 1. 7 Ins t rumentat ion and Cont rol Drawings. . . . . . . . . . . . . . . . . . . . . . . 12 Chapter 2 SITE CHARACTERISTICS 2.1 Geography and Demog raphy. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 2.1.1 Si te Location and Des crip tion. . . . . . . . . . . . . . . . . . . . . . . 13 2.1.1.1 Specification of Location.................. 13 2.1.1. 2 Site Area Map.............................. 13 2.1.1.3 Boundaries for Establishing Effluent Release Limits............................. 13 2.1.2 Population Distribution............................. 18 2.1.2.1 Population Within 10 Miles. . . . . . . . . . . . . . . . . 18 2.1.2.2 Population Be tween 10 and 50 Miles. . . . . . . . . 18 2.1.2.3 Transient Popula tion. . . . . . . . . . . . . . . . . . . . . . . 20 2.1.2.4 Low Population Zo ne. . . . . . . . . . . . . . . . . . . . . . . . 20 2.1.2.5 Population Center.......................... 20 2.1. 2. 6 Population Density Around the VTAR Site.... 20 2.2 Nearby Industrial, Transportation and Military Facilities.. 21 2.2.1 Lo ca tions and Rout es . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 i 2.2.2 Descriptions........................................ 22 2.2.2.1 Description of Transportation Routes....... 22 2.2.2.2 Description of Virginia Tech Airport....... 22 2.2.2.3 Proj ections of Indus t rial Growth. . . . . . . . . . . 22 i
() 2.2.2.4 Description of Military Facilities......... 22 11
1 i i Page _/ 2.3 Meteorology and Climatology................................ 22 2.3.1 Regional Climatology. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 2 2.3.2 Local Meterology.................................... 24 2.3.2.1 Normal and Extreme Values of Meteorological Parameters................................. 24 2.3.2.2 Potential Influence of the VTAR and its Facilities on Local Meteorology............ 24 2.3.2.3 Local Meteorological Conditions for Design
- and Operating Bases 2.4 Hydrologic Engineering..................................... 31 2.4.1 Flooding............................................ 33 2.4.1.1 Flood History.............................. 33 2.4.1.2 Flo od De s ign . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 3 2.4.1.3 Effects of Local Intense Precipitation..... 34 2.4.2 Probable Maximum Flood of Streams or Rivers......... 34 2.4.3 Potential Dam Failures, Seismically Induced......... 34 2.4.4 Probable Maximum Surge and Seiche Flooding.......... 34 O- 2.4.5 Probable Maximum Tsunami Flooding. . . . . . . . . . . . . . . . . . . 34 2.4.6 Ice Effects......................................... 34 2.4.7 Cooling Water Canals and Reservoirs................. 35 2.4.8 Channel Diversions.................................. 35 2.4.9 Flooding Pro te ction Requirements. . . . . . . . . . . . . . . . . . . . 35 2.4.10 Low Water Considerations............................ 35 2.4.11 Dispersion, Dilution, and Travel of Accidental Releases of Liquid Effluents in Surface Waters...... 35 2.4.12 Ground Water........................................ 35 2.4.13 Technical Specifications and Emergency Operations Requirements........................................ 36
- 2. 5 Geology, Seismology, and Geotechnical Engineering.......... 36 2.5.1 Basic Geology and Seis mic Inf o rmation. . . . . . . . . . . . . . . 3 6 2.5.2 S i t e Ge o lo gy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 6
- 2.5.3 Vibrat o ry Ground Mo tion. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 7 2.5.4 Surface Faulting.................................... 38 2.5.5 Stability of Subsurface Materials and Foundations... 38 2.5.6 S tabili ty of S1 opes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 2.5.7 Embankme n t s and Da ms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 j
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?"Be Chapter 3 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT AND SYSTEMS 3.1 Stru'etural Design.......................................... 42 3.2 Utilities and Servicee>>................................... 47 3.2.1 Vent 11ction......................................... 47 3.2.2 Fire Protection..................................... 47 3.2.3 Flood Protection.................................... 49 Chapter 4 REACTOR 4.1 Su mma ry De s c ri p t io n. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 0 4.1.a General Reactor System Description.................. 50 4.1.2 Design and Performance Characteristics.............. 54 4.1.3 Shielding........................................... 56 4.1.4 Experimental Facilities............................. 57 4.2 Fuel System Design......................................... 57 4.3 Nuclear Designs............................................ 62 4.3.1 Flux Distributions.................................. 62 4.3.2 Xenon Transients.................................... 62 4.3.3 Other Reactivity Considerations..................... 73 4.4 Thermal Hydraulics......................................... 73 4.4.1 Primary Coolant Sys t em. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 74 4.4.2 Intermediate Heat Exchanger and Secondary Coolant System.............................................. 74 4.4.3 Cooling Tower................~....................... 74 4.4.4 Emergency Core Cooling System....................... 74 4.4.5 The rmal Hydraulic Limiting Conditions. . . . . . . . . . . . . . . 76 4.4.6 Steady-State Hydraulic Analysis..................... 76 4.4.6.1 Core Heat Balance.......................... 76 4.4.6.2 Intermediate Heat Exchanger Heat Balance... 77 4.4.6.3 Hot Channel Heat Balance................... 77 4.4.6.4 Bulk Coolant Temperature Profile in Hot Channe1.................................... 78 4.4.6.5 Fuel Clad Surface Temperature Profile...... 78 4.4.6.6 Results of Steady-State Thermal Hydraulic Analysis................................... 79 4.4.7 Transient Thermal Hydraulic Analysis................ 81 4.4.7.1 Types of Accidents Studied................. 81 4.4.7.2 Loss of Flow Accidents . . . . . . . . . . . . . . . . . . . . . 81 4.4.7.3 Loss of Coolant Accidents. . . . . . . . . . . . . . . . . . 81 4.4.8 Conclusions......................................... 82 iv
() Page 4.5 Materials.................................................. 82 4.6 Reactivity Control Systems................................. 82 Appendix 4-I Materials Incorporated ic VTAR. . . . . . . . . . . . . . . 85 Chapter 5 COOLING SYSTEM 5.1 General Description........................................ 86 5.2 Primary Coolant Sys t em. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 86 5.2.1 System 0verview..................................... 86 5.2.2 Primary Coolan t rump. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 90
- 5. 2. 3 Fl ow Mo ni t o r . . . .' . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 0 5.2.4 Heat. Exchanger.....l................................ 90 5.2.5 Inlet and Outlet Temperature Monitors............... 90 5.2.6 Primary Coolant Radiation Monito r. . . . . . . . . . . . . . . . . . . 92 5.2.7 Shield Tank / Core Tank Level Monitors................ 92 5.2.8 Du mp Va 1v e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 2 5.2.9 De mi ne ra li z e r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 2
- 5. 3 Se co nda ry Coolant Sys tem. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 3 5.3.1 System 0verview..................................... 93 5.3.2 Seconda ry Coolant Pump . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 93
() , 5.3.3 5.3.4 Th ro t t le Va1ve . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 3 Pressure Indicator.................................. 93 5.3.5 Cooling Tower....................................... 93 5.3.6 Tempe rature Monito rs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 94 5.3.7 F1owmeter........................................... 94 e 5.4 Func tional Blo ck Diagrams . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 4
- 5. 5 Cold Water Accident Operational Considerations............. 94 Appendix 5-1 Cooling System Specifications............... 100 Chapter 6 ENGINEERED SAFETY SYSTEMS 6.1 Emergency Co re Cooling Sys tem (ECCS) . . . . . . . . . . . . . . . . . . . . . . 106 6.2 Sy s t e m 0v e rvi ew . . . . . . . . . . . . . . . . . . . . . . . . . . . .' . . . . . . . . . . . . . . . 10 6 6.3 System Operation.......................................... 106 6.3.1 Emergency Coolant Pump............................. 106 6.3.2 Flow Monitor....................................... 108 6.3.3 ECCS Heat Exchanger................................ 108 6.4 Boron Injection System.................................... 108 i
l 6.4.1 System 0verview.................................... 108
- - 6.4.2 System 0peration................................... 108 l
l p) s_ . Appendix 6-1 Tables of Specifications.................... 110 V t s
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Appendix 6-11 Boric Acid Requirement...................... 112 Chapter 7 VTAR INSTRUMENTATION AND CONTROLS 7.1 Introduction.............................................. 114
- 7. 2 Identification of Safety Related Systems.................. 114 7.2.1 Console Mimic Bus.................................. 114 7.2.2 Radiation Monitoring System........................ 116 ,
7.2.2.1 APFPM..................................... 118 7.2.2.2 Area-Radiation Monitoring System.......... 118 7.2.3 Nuclear Ins t rumenta tion. . . . . . . . . . . . . . . . . . . . . . . . . . . . 123 7.2.3.1 Source Range Instruments.................. 124 7.2.3.2 Source Range Number 2. . . . . . . . . . . . . . . . . . . . . 124 7.2.3.3 Intermediate Range Ins t rument . . . . . . . . . . . . . 126 7.2.3.4 Power Range Ins t ruments. . . . . . . . . . . . . . . . . . . 126 7.2.3,5 Keichley Picoammeters . . . . . . . . . . . . . . . . . . . . . 126 7.2.4 Reactor Protective System.......................... 126 7.3 Non-Nuclear Instrumentation and Control Circuits..... ..... 127 O 7.3.1 Control Rod Drive Mechanisms....................... 129 7.3.2 Regulating-Rod Control Sys tem. . . . . . . . . . . . . . . . . . . . . . 133 7.3.3 S t ar tup /In te rlock Bus . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 3 7.3.4 Annunciator System................................. 137 7.3.5 Rod Position Indica tion Sys tem. . . . . . . . . . . . . . . . . . . . . 13 9 7.3.6 Primary Coolant Flow Monitor....................... 141 7.3.7 Secondary Coolant Flow Monitor..................... 141 Chapter 8 ELECTRICAL DISTRIBUTION 8.1 System 0verview........................................... 145 8.2 Off-Site Power Distribution............................... 145 8.2.1 440 VAC Distribution............................... 145 8.2.2 110/220 VAC Distribution........................... 145
- 8. 3 On-Site Powe r Dis t ribution. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 145 8.3.1 440 VAC Distribution............................... 145 8.3.2 110/220 VAC Dis t ribu tion. . . . . . . . . . . . . . . . . . . . . . . . . . . 145 8.4 Uninte rrup tible Power Supp1y. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 150 l 8.4.1 System Description................................. 150 8.4.2 Des cription of Sys tem Ope ration. . . . . . . . . . . . . . . . . . . . 150
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l 8.4.2.1 Input Power Converter..................... 150 8.4.2.2 Ba t t e ry Ba nk. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 150 8.4.2.3 Static Inverter........................... 152 Chapter 9 AUXILIARY SYSTEMS 9.1 Fuel Storage and Handling................................. 153 9.1.1 New Fuel Storage................................... 153 9.1.2 Spent Fuel Storage................................. 153 9.1.3 Reactor Rooe Ovarhecd Crane. . . . . . . . . . . . . . . . . . . . . . . . 153 9.2 Wa t e r Sy s t e ms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 6 9.2.1 Shield Tank Water Maintenance System............... 156 9.2.2 Primary Make-up Water System....................... 157 9.2.3 Secondary Make up Water Sys tem. . . . . . . . . . . . . . . . . . . . . 158 9.3 Reactor Room Ventilation System........................... 158 9.3.1 Description of the System.......................... 158 9.3.2 Operation of the Ventilation Sys tem. . . . . . . . . . . . . . . . 159 9.4 O t he r Sy s t e ms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 4 O 9.4.1 Pneumatic Rabbi t Sys tem. . . . . . . . . . . . . . . . . . . . . . . . . . . . 164 9.4.2 central Stringer Irradiations...................... 167 9.4.3 Beam Port Irradiations............................. 167 Chapter 10 STEAM AND POWER CONVE RSION. . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 9 Chapter 11 RADIOACTIVE WASTE MANAGEMENT 11.1 Introduction........... 4................................. 170 11.2 Solid Waste Management.................................... 170 11.3 Liquid Waste Management................................... 171
- 11. 4 Gaseous Was t e Management . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 2
- 11. 5 Co n c lu s i o ns . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 3 l Appendix 11-I Concentration of Argon-41 in Unrestricted Areas....................................... 174 Chapter 12 RADIATION PROTECTION 12.1 Health Physics Program.................................... 177 12.1.1 Responsibilities................................... 177 12.1.2 Duties of the Reactor Radiation Safety Officer.....177 12.1. 3 Routine Inspe ctions and Surveys . . . . . . . . . . . . . . . . . . . . 17 9 12.1.4 Radiation Safety Training.......................... 179
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O Page \ i 12.1.5 Conclusion......................................... 180
- 12. 2 Radiation Saf e ty Ins t ruments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 180 12.2.1 Portable Survey Equipment and Stationery Counting Systems............................................ 180 12.2.2 Installed Radiation Monitor........................ 181
- 12. 2. 3 Ins t rument Calibration and Quality Cont rol. . . . . . . . . 181 )
12.2.4 Conclusion......................................... 182 12.3 Personnel Monitoring...................................... 182 12.3.1 Occupational....................................... 182 12.3.2 Non-Occupational................................... 183 12.3.3 Conclusion......................................... 184 12.4 Radiation Surveys - Restricted Areas...................... 184 12.4.1 Shielding.......................................... 184
- 12. 4. 2 Radiation Leve1s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 184 12.4.3 Contamination Surveys.............................. 184 12.4.4 Sealed Sources..................................... 190 12.4.5 Primary Coolant Analysis........................... 191 12.4.6 Argon-41 Sampling.................................. 191
() 12.4.7 Conclusion......................................... 191 12.5 Radiation Surveys - Unrestricted Areas.................... 192 12.5.1 Shielding.......................................... 192 12.5.2 Environmental TLD's................................ 192
- 12. 5.3 Area Radiation Leve1s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 194
- 12. 5. 4 Contamination Swipes. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 194 12.5.5 Argon-41 Sampling.................................. 194 12.5.6 conclusion......................................... 194
- 12. 6 Ge ne ral Conclusion. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 198 Chapter 13 CONDUCT OF OPERATIONS 13.1 Organizational St ructu re of Applicant. . . . . . . . . . . . . . . . . . . . . 199 13.1.1 Management and Technical Support organization......199 13.1. 2 Ope rating organization. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 9 13.1.2.1 Director of the Nuclear Reactor Laboratory and Reacto r Supe rvisor. . . . . . . . . . . . . . . . . . . 199 13.1.2.2 VTAR Reactor Safety Committee............ 201 13.1.2.2.1
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l Purpose of the Reactor Safety L Committee.................... 201 viii
P, age, 13.1.2.2.2 Charter and Rules of the RSC. 201 13.1.2.2.3 Membership of the RSC........ 202 13.1.2.2.4 Review Function of the RSC... 202 13.1.2.2.5 Audit Function of the RSC.... 203 13.1.2.3 Radiation Safety Organization............ 203 13.2 Training 13.2.1 Operator Training.................................. 204 13.2.2 Replacement and Retraining......................... 204 13.2.2.1 Requalification Schedule................. 205 13.2.2.2 Lectures, Reviews and Examinations....... 205 13.2.2.3 Requalification Operatioes............... 206 13.2.2.4 Emergency Dri11s......................... 206 13.2.2.5 Absence from Authorized Acitivities...... 206 13.2.2.6 Exemptions............................... 207 13.2.2.7 Evaluation of Operators.................. 207 13.2.2.7.1 On the Job Training.......... 207 13.2.2.7.2 Grade Requirements........... 207 13.2.2.7.3 Accelerated Training......... 208 13.2.2.8 Requalification Records.................. 208 13.2.2.9 F3 qualification Document Review.......... 208 13.3 Plant Procedures ......................................... 208 13.3.1 Administrative Procedures - Access Control......... 209 13.3.2 Operating and Maintenance Procedures............... 209 13.3.2.1 Routine Operations and Records........... 209 13.3.2.2 Routine Tests, Maintenance and Monitoring.210 13.3.2.2.1 Operational Startup Checks.... 211 13.3.2.2.2 Quarterly Maintenance Checks.. 212 13.3.2.2.3 Semi-A:nual Maintenance Checks.212 13.3.2.2.3 Annual Maintenance Checks..... 212 13.4 Emergency Planning........................................ 213 13.5 Security ......................................... 213 Appendix 13-1 Charter.................................... 214 Chapter 14 INITI AL TEST PR0 GRAM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 217 14.1 Specific Information to be Included in Preliminary Safety Analysis Report.......................................... 217 O ix
() Page 14.2 Safety Information to be Included in Final Safety Analysis Report................................................... 217 Chapter 15 ACCIDENT ANALYSIS.................................... 219 15.1 In t rodu c t io n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 219 15.2 Reactivity Additions..................................... 219 15.2.1 Cold Water Accident.............................. 219 15.2.2 Sudden Addition of Maximum Excess Reactivity..... 220 15.2.3 Safety Control Blade System Malfunction.......... 221 15.3 Loss of Flow Accidents (L0FA). . . . . . . . . . . . . . . . . . . . . . . . . . . . 222 15.3.1 LOFA at 500 kW................................... 222 15.3.1.1 LOFA wi t h ARO . . . . . . . . . . . . . . . . . . . . . . . . . . 2 2 2 15.3.1.2 LOFA wi th ARI. . . . . . . . . . . . . . . . . . . . . . . . . . 2 2 3 15.3.2 LOFA at 100 kW................................... 223 15.3.2.1 LOF A wit h ARO. . . . . . . . . . . . . . . . . . . . . . . . . . 2 2 3 15.3.2.2 LOFA with ARI.......................... 228 15.4 Loss of Coolant Accidents (L0CA)......................... 228 15.4.1 LOCA with no Heat Transfer....................... 228 15.4.2 LOCA with Natural Convection..................... 228 15.5 Fuel Element Failure..................................... 229 15.5.1 Methodology...................................... 229 15.5.1.1 Basic Equations........................ 229 15.5.1.1.1 Transport of Released Radionuclides.............. 229 15.5.1.1.2 Dose Equations. . . . . . . . . . . . . 230 15.5.1.2 In p u t Da t a . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 31 1 15.5.1.2.1 Radioisotope Inventory in the Core....................... 231 15.5.1.2.2 Dispersion Factors and Leak Rate....................... 232 15.5.1.2.3 Dose Conversion Factors.... 232 15.5.1.2.4 Fuel Release Fractions..... 232 l 15.5.2 Dose Calculations................................ 232 (} 15.5.3 Results of Dose Calcula tion. . . . . . . . . . . . . . . . . . . . . . 234 X
Page 15.5.4 Selection of Site Parameters Based on Dose Calculations................................... 243 15.6 Earthquake....-......................................... 244 15.7 Fuel Handling Accide nt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 4 15.8 Storage Pit Criticality Calculati on. . . . . . . . . . . . . . . . . . . . . 248 15.9 Explosive Chemical Reactions. . . . . . . . . . . . . . . . . . . . . . . . . . . . 254 15.10 Experimental / Rabbit System Failures..................... 254 15.10.1 Sample Lodged in the Reactor................... 255 15.10.2 Sample Lodged Between Rabbit Receiver and Reactor........................................ 255 15.10.3 Ruptured or Leaking Sample Capsule. . . . . . . . . . . . . 255 15.11 Fires.................................................... 256 15.11.1 Introduction.................................... 256 15.11.2 Oxygen Sources.................................. 256 15.11.3 Fuel Sources.................................... 257 15.11.4 Igni tion Sou rces . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 5 8 15.12 Dump Va lve Fa i lur e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 5 8 15.12.1 Electrical Faults............................... 259 15.12.1.1 Dump Valve Motor Failure............. 259 15.12.1.2 Clutch Failure....................... 259 15.12.2 Mechanical Failure.............................. 259 15.12.2.1 Dump Valve Stuck open. . . . . . . . . . . . . . . . 25 9 15.12.2.2 Dump Valve Stuck in Closed Position.. 260 15.13 Secondary Coolant System Rupture......................... 261 15.14 Le aks and S p i11s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 61 15.15 Strikes by Personne1..................................... 262 Appendix 15-1 13 Plate Storage Pit Criticality . Calculations................................ 264 Chapter 16 TECHNICAL S PEC IFICATIONS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 268 f 16.1 De f i ni t ions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 6 9 16.2 Safety Limits and Limiting Safety System Settings........ 277 16.2.1 Safety Limits................................... 277 16.2.1.1 Applicability......................... 277 16.2.1.2 0bj e ct ive . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 7 7 16.2.1.3 Specifications........................ 277 xi
O
- 16.2.1.4 Bases................................. 277 16.2.2 Limiting Safe ty Sys tem Se t tings. . . . . . . . . . . . . . . . . 278 16.2.2.1 Applicability......................... 278 16.2.2.2 0bj e c tive s . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 7 8 16.2.2.3 Specifications........................ 278 16.2.2.4 Bases..........v..................... 278 16.3 Limiting Condition f or Ope ration (LCO) . . . . . . . . . . . . . . . . . . . 280 16.3.1 Reactivity L1.,itations.......................... 280 16.3.1.1 Specifications........................ 280 16.3.1.2 Bases................................. 281 16.3.2 Reactor Control and Protection Systems.......... 281 16.3.2.1 Specifications........................ 281 16.3.2.1.1 Control Rods............... 281 16.3.2.1.2 Reactivity Insertion Rates. 282 16.3.2.1.3 Nuclear Instrumentation.... 282 16.3.2.1.4 Scram Channels............. 283
-0 1e.3.2.1.s Bacx , scram channe1s...... 283 16.3.2.1.6 Bypass Conditions.......... 286 16.3.2.2 Bases................................. 286 16.3.3 Coolant Systems................................. 286 16.3.3.1 General Specifications................ 286 16.3.3.2 Primary Coolant System................ 290 16.3.3.2.1 Maj or Components . . . . . . . . . . 290 16.3.3.2.2 Flow Path................. 290 16.3.3.2.3 Specifications............ 290 16.3.3.3 Secondary Coolant System.............. 291 16.3.3.3.1 Major Components.......... 291 16.3.3.3.2 Flow Path................. 292 16.3.3.3.3 Specifications............ 292 16.3.3.4 Emergency Core Cooling System (ECCS).. 292 16.3.3.4.1 Major Components.......... 293 16.3.3.4.2 Flow Path................. 293 16.3.3.4.3 Specifications............ 293 0
xii b
1 i E*B *. l 16.3.3.5 Bases................................. 294 16.3.3.5.1 Primary Coolant System.... 294 l 16.3.3.5.2 Secondary Coolant Systhm.. 295 16.3.3.5.3 Emergency Core Cooling System.................... 295 16.3.4 Co n f i ne me n t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 9 6 16.3.4.1 Normal Confinement.................... 296 16.3.4.1.1 Operations that Require Normal Confinement........ 296 16.3.4.1.2 Equipment to Achieve Mode I Confinement............... 296 I 16.3.4.2 Emergency Confinement................. 296 16.3.4.2.1 Operations that Require Emergency Confinement (Mode 11)................. 296 16.3.4.2.2 Equipment to Achieve Mode I Confinement. . . . . . . . . . . . . . . 297 16.3.4.3 Bases................................. 297 16.3.5 Reacto r Room Ventilation. . . . . . . . . . . . . . . . . . . . . . . . 297 16.3.5.1 Components............................ 297 16.3.5.2 Specifications........................ 298 16.3.5.3 Bases................................. 298 l 16.3.6 Emergency Power................................. 298 16.3.6.1 Components............................ 298 16.3.6.2 Specifications........................ 299 16.3.6.3 Bases................................. 299 16.3.7 Radiation Monitoring System and Effluents....... 299 16.3.7.1 Fixed Radiation Monitoring............ 299 16.3.7.2 Portable Radiation Monitoring......... 302 16.3.7.3 Radiation Monitoring Bases............ 302 16.3.7.4 E f f lu e n t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 0 3 16.3.7.4.1 Argon..................... 303 16.3.7.4.2 Other Effluents........... 303 16.3.7.4.3 Bases..................... 303 () 16.3.8 -Limitations on Experiments...................... 303 xiii 1 i
1 i m Page k_) 16.3.8.1 Specifications........................ 303 16.3.8.2 Bases................................. 306 16.3.9 Building Evacuation System...................... 307 16.3.9.1 Specifications........................ 307 16.3.9.2 Bases................................. 308 16.3.10 Fuel and Fuel Handling and Storage.............. 308 16.3.10.1 Specification........................ 308 16.3.10.2 Bases................................ 309 16.4 Surveillance Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 310 16.4.1 General Requirements and Surveillance Intervals....................................... 310
- 16.4.1.1 Surveillance Intervals................ 310 16.4.1.2 General Requirements.................. 310 16.4.2 Reactivity and Co re Pa rame te rs. . . . . . . . . . . . . . . . . . 311 16.4.3 Reactor Control and Safety...................... 312 16.4.4 Coolant Systems................................. 314
{} 16.4.4.1 Primary Coolant System................ 314 16.4.4.2 Secondary Coolant Sys tem. . . . . . . . . . . . . . 314 16.4.4.3 Emergency Core Cooling System......... 315 16.4.5 Confinement..................................... 315 i 16.4.6 Reactor Room Ventila tion. . . . . . . . . . . . . . . . . . . . . . . . 315 16.4.7 Emergency....................................... 316 16.4.8 Radiaton Monitoring and Effluents............... 316 16.4.8.1 Radiation Mo nitoring. . . . . . . . . . . . . . . . . . 316 - 16.4.8.2 Effluents............................. 316 16.4.9 Experiments..................................... 316 16.4.10 Building Evacuation System...................... 316 16.4.11 Fuel and Fuel Handling and Storage.............. 316 16.4.12 Reactor Components and S tructure. . . . . . . . . . . . . . . . 317 16.4.13 S'nutdown Surveillance Requirements. . . . . . . . . . . . . . 317 l, 1 16.4.13.1 Reactor Room Ventilation. . . . . . . . . . . . . 317 ) 16.4.13.2 Emergency Power...................... 317 ' 16.4.13.3 Radiation Monitoring................. 317 16.4.13.4 Building Evacuation System........... 318 16.4.13.5 Fue l, and Fuel Handling. . . . . . . . . . . . . . 318 () 16.5 De s ign Fe a tu re s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 319 xiv
O na 16.5.1 Site............................................ 319
- 16.5.2 Reactor Facility................................ 319 16.5.2.1 Restricted Areas...................... 319 '
16.5.2.2 Reactor Room Ventilation. . . . . . . . . . . . . . 319 l 16.5.2.3 Reactor Bridge Crane.................. 319 1 16.5.3 Reactor Coolan t Sys tems . . . . . . . . . . . . . . . . . . . . . . . . . 320 16.5.3.1 Primary Coolant Sys tem. . . . . . . . . . . . . . . . 320 16.5.3.2 Secondary Coolant System.............. 320 16.5.3.3 Emergency Core Cooling System......... 321 16.5.4 React or Co re and Fue1. . . . . . . . . . . . . . . . . . . . . . . . . . . 3 21 16.5.4.1 Reactor Core.......................... 321 16.5.4.2 Reactor Fuel.......................... 322 16.5.4.3 Fuel Transfers........................ 322 , 16.5.5 Fissionable Material Storage. . . . . . . . . . . . . . . . . . . . 322 16.6 Adminis t ra tive Controls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 23 16.6.1 Organization.................................... 323 O 16.6.1.1 Structure............................. 323 16.6.1.2 Re s ponsibility. . . . . . . . . . . . . . . . . . . . . . . . 323 16.6.1.3 Staffing.............................. 325 16.6.1.4 Selection and Training of Personnel... 325 16.6.1.5 Audit and Review...................... 326 16.6.2 Procedure....................................... 328 16.6.3 Experiments Review and Approval. . . . . . . . . . . . . . . . . 329 16.6.4 Required Actions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 30 16.6.5 Re p o r t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 3 1 16.6.5.1 Ope rating Repor ta . . . . . . . . . . . . . . . . . . . . . 331 16.6.5.2 Special Reports....................... 332 16.6.6 Records......................................... 333 16.6.6.1 Records to be Maintained for a Period of at leas t Five Ye ars . . . . . . . . . . . . . . . . 3 33 16.6.6.2 Records to be Retained for at Least One Training Cycle........................ 334 16.6.6.3 Records to be Retained for the Life of the Facility.......................... 334 Chapter 17 QUALITY ASSURANCE..................................... 336 O
\_/. ~ References ................................................ 337
- v
(('j) LIST OF FIGCAE3 Figure Title Page 1-1 Co re Ove r vie w. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2-1 Virginia Radiological Emergency Response Plan............ 14 2-2 Ge ne ral Si te Loca tion. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 2-3 Ca mp us Lay ou t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 2-4 Restriction Area......................................... 17 2-5 Magisterial District Populations......................... 19 2-6 A-F Wind Direction and Velocities............................ 25 2-7 Seismic Acceleration Probability Map..................... 39 2-8 Seismic Risk Map for Conterminous United States.......... 40 3-1 Ground and First Floor P1ans............................. 43 3-2 Second and Third Floor Plans............................. 44 3-3 Plan of 24" Thick Concrete Shield in Floor Over Reactor Room................................................ 45 3 Basement Plan............................................ 46 4-1 Reactor Cross Section.................................... 51 4-2 Top View of Reactor Room................................. 52 4-3 Experimental Facilities.................................. 53 4-4 Top View of Corebox...................................... 58 4-5 Is o me t ric View of Co re Tank. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 9 4-6 Fuel Plate and Cell Geome try. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 0 4-7 Plan View of Fuel Element . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 61 4-8 Vertical Flux Profile in SW Quadrant..................... 63 O. 4-9 Horizontal Flux Profile.................................. 64 4-10 Vertical Flux Profile Along Central Stinger of Internal Reflector........................................... 65 4-11 The rmal Flux at Face of Duct in Shield Tank. . . .. . . .. . .. . . 66 4-11 Flux Profile Across Element W-3 (E to W)................. 67 4-13 Ve rtical Flux Profile in NE S t ringer. . . . . . . . . . . . . . . . . . . . . 6 8 4-14 Measured Reactivity Transient. . . . . . . . . . . . . . . . . . . . . . . . . . . . 69 4-15 Regulating Rod Calibration. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 70 4-16 Xe non Transients (100 kW and 500 kW). . . . . . . . . . . . . . . . . . . . . 71 4-17 Kenon Transients ( 500 kW). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 72 4-18 General Schematic of the Cooling Sys tem. . . . . . . . . . . . . . . . . . 7 5 4-19 Fuel Surface Temperature at.Several Power Levels......... 80 4-20 Operating Limits for Reactor............................. 80 4-21 Regulating Rod Reactivity Measurement. . . . . . . . . . . . . . . . . . . . 84 5-1 Schematic of Reactor Cooling System...................... 87 5-2 100 Kilowatt Nuclear Reactor Piping Diagram. . . . . . . . . . . . . . 88 5-3 Flow Pathways in the Heat Exchanger...................... 91 5-4 Cooling Tower Location (Plan) . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 5 5-5 Cooling Tower (Elevation)................................ 96 5-6 Layout of Cooling Tower and Secondary Piping............. 97 5-7 Coolant System Isometric................................. 98 6-1 VTAR ECCS System........................................ 107 7-1 Co ns o le Mimi c Bus . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 115 7-2 VTAR Console............................................ 117 7-3 Airborne Particulate Fission Product Monitor............ 119 0
. xvi f
1 l O ria re '1ete '--se-7-4 APFPM Vent S ta ck Spectrum (Typical) . . . . . . . . . . . . . . . . . . . . . 120 7-5 Area Radiation Monitoring Sys tem. . . . . . . . . . . . . . . . . . . . . . . . 121 7-6 Functional Diagram / Rad. Monitor Control Circuits........ 122 l 7-7 VTAR Nuclear Ins t rumenta tion. . . . . . . . . . . . . . . . . . . . . . . . . . . . 125 7-8 Reactor Protective System Block Diagram................. 128 7-9 Co nt ro l Rod Drive . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 131 7-10 VTAR Startup/ Interlock Bus.............................. 134 7-11 Simplified Block Diagram of the VTAR Primary Flowmeter.. 142 7-12 Secondary Coolant Flowmeter Functional Block Diagram.... 144 8-1 Elect rical Dis tribution. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 146 8-2 Power Panel Locations (Ground Floor). . . . . . . . . . . . . . . . . . . . 14 7 8-3 Power Panel Locations (First Floor)..................... 148 8-4 Uninterruptible Power Supply............................ 151 9-1 AMF Master / Slave Manipulator............................ 154 9-2 VTAR Ventilation Diagram. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 161 9-3 Schamatic Diagram of Purge Points....................... 162 9-4 VTAR Ve n t S t a ck . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 3 9-5 VTAR Pneumatic dabbit System Functional Diagram......... 165 9-6 central Stringer Irradiation Block...................... 168 12-1 Administrative Organizational Chart for the Readiation Safety Program at Virginia Tech.................... 178 12-2 Radiation Survey Points, VTAR............................ 186 12-3 Gamma Isodose Curve (mrer/hr.) O 12-4 October 1978, VIAR, 100 kW......................... 187 Fast Neutron Isodose Curves (mrem /hr.) October 1978, VTAR, 100 kW......................... 188 12-5 Thermal Neutron Isodose Curves (mrem /hr.) October 1978, VTAR, 100 kW......................... 189 12-6 Location of Environmental TLD's, VTAR................... 193 12-7 Survey Points, Unrestricted Areas, VTAR................. 195 13 I VTAR Organizational Structure........................... 200 15-1 Fuel Temp e ra ture LOFA AR0. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 24 15-2 Fuel Temperature LOFA ARI/L0CA.......................... 225 15-3 Fuel Temperature 10CA................................... 226 15-4 Fuel Temperature LOFA ARI............................... 227 15-5 2 Hour Whole Body Dose 100 and 500 kW Severe Conditions. 235 15-6 24 Hour Whole Body Dose 100 and 500 kW Severe Conditions.236 15-7 30 Day Whole Body Dose 100 and 500 kW Severe Conditions. 237 15-8 2 Hour Thyroid Dos e 500 kW Local Me teorology. . . . . . . . . . . . 238 15-9 2 Hour Thyroid Do se 100 kW Local Me teo rology. . . . . . . . . . . . 239 15-10 24 Hour Thyroid Dose 100 kW Local Meteorology. ..... . .. .. 240 15-11 24 Hour Thyroid Dose 500 kW Local Me teorology. . . . . . . . . . . 241 15-12 30 Day Thyroid Dose 100 and 500 kW. . . . . . . . . . . . . . . . . . . . . . 242 16-1 Generalized Reactor Saf ety System. . . . . . . . . . . . . . . . . . . . . . . 273 16-2 VTAR Organizational Structure........................... 324 I xvii
g LIST OF TABLES Table Title Page 1-1 Compa ris on of VTAR, UFTR, UWNR. . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1-2 Facility History.......................................... 9 2-1 Population Distribution Around the VTAR Site............. 21 2-2 Probability of Late or Early Freezes..................... 31 2-3 Climatological Summary................................... 32 3-1 VTAR Reactor Room Penetrations........................... 48 3-2 VTAR Facility Fire Extinguishers......................... 49 4-1 Nominal VTAR Characteristics............................. 55 5A-1 Primary Coolant Pump Design Specifications. . . . . . . . . . . . . . 100 SA-2 Primary Flow Monitor Design Specifications.............. 100 SA-3 Heat Exchanger Design Sp ecifications. . . . . . . . . . . . . . . . . . . . 101 SA-4 Design Specifications for the Temperature Monitors...... 101 SA-5 Design Specifications for the Radiation Monitors........ 102 SA-6 Level Monitor Design Specifications. . . . . . . . . . . . . . . . . . . . . 102 5A-7 Design Specifications f or the Dump Valve. . . . . . . . . . . . . . . . 103 5A-8 Design Specifications f or the Demineralizer. . . . . . . . . . . . . 103 5A-9 Secondary Coolant Pump Design Sr teifications........ .... 103 5A-10 Design Specifications for the Throttle Valve............ 104 SA-11 Cooling Towe r Design Specifications . . . . . . . . . . . . . . . . . . . . . 104 5A-12 Design Specifications for the Secondary Flowmeter....... 105 6A-1 ECCS Pump Design Specifications . . . . . . . . . . . . . . . . . . . . . . . . . 110 6A-2 ECCS Flow Monitor Design Specifications................. 110 lll 6A-3 7-1 ECCS Heat Exchanger Design Specifications. . . . . . . . . . . . . . . 111 VTAR Protective System Set Points....................... 129 7-2 Safety Rods 1 and 2 CRDM Performance Characteristics....132 7-3 Shim Rod CRDt1 Perf ormance Characteristics. . . . . . . . . . . . . . . 132 7-4 Regulating Rod CRDM Performance Characteristics.........133 7-5 Startup/ Interlock Bus Set Points........................ 136 7-6 VTAR Annuncia to r Se t Points . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 0 7-7 VTAR Rod Position Indicator Circuits . . . . . . . . . . . . . . . . . . . . 141 8-1 On-S i te B reake r Pane l Load s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 9 9-1 VTAR Overhead Crane..................................... 156 9-2 VTAR Ventilation Characteristics........................ 160 11-1 Solid Waste Disposal, VTAR.............................. 170 11-2 Liquid Waste Released from VTAR to Sanitary Sewerage System.............................................171 11-3 Ar-41 S t ack Dis cha rge Ra t e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17:' 11-4 Ar-41 Stack Discharge Yearly Totals..................... 173 12-1 Occupational Whole Body Gamma Doses , VTAR. . . . . . . . . . . . . . . 183 12-2 Non-Occupational Whole Body Gamma Dose s. . . . . . . . . . . . . . . . . 183 12-3 Radia t ion Leve ls , VTAR. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 5 12-4 Contamination Limits, Restricted Areas, VTAR............ 190 12-5 Environme n tal TLD Re sults . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 196 12-6 Radiation Levels , l'nres t ricted Are a. . . . . . . . . . . . . . . . . . . . . 197 12-7 Contamination Limits , Unres tricted Areas. . . . . . . . . . . . . . . . 197 13-1 Topics for Requalification.............................. 205 15-1 Equilibriun Gaseous Radionuclide Concentrations......... 231 15-2 Radionuclide Decay Constants and Dose Conversion Factors.234 9 M'. xviii
l Table Title Page (al 15-3 Activity and Dose Equivalents........................... 247 15-4 Aar.umptions and Notes for VTAR Fuel Handling Accident... 248 ' 15-5 Parameters for Multiplication Calculations ( 12 pla t e ele men ts ). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 50 i 15A-1 Parameters for Multiplication Calculations ( 13 pla te ele me n ts ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 264 16-1 Nuclear Ins trument Capabili ties . . . . . . . . . . . . . . . . . . . . . . . . . 284 16-2 Reacto r Protection Sys tem. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 285 16-3 Bypass Criteria......................................... 287 16-4 Major Coolant System Components......................... 289 16-5 Permanent Radiation Monitors............................ 300 16-6 Channel Calibration Requirements . . . . . . . . . . . . . . . . . . . . . . . . 313 O O ixx
(} LIST OF ABBREVIATIONS ALARA - As Low As Reasonably Achievable ANS - American Nuclear Society ANSI - American National Standards Institute AP - Administrative Procedure APFPM - Airborne Particulate Fission Product Monitor ARI - All Rods In ARO - All Rods Out. BIT - Boron Injection Tank cfm - cubic feet per minute CFR - Code of Federal Regulations Ci - Curie CIC - compensated ion chamber cps - counts per second CRDM - Control Rod Drive Mechanism CWA - Cold Water Accident dps - disintegrations per second DOT - Department of Transportation ECCS - Emergency Core Cooling System EP - Emergency Procedure EPA - Environmental Protection Agency FSAR - Final Safety Analysis Report GM - Geiger-Muller gpm - gallons per minute () IHX IPC Intermediate heat exchanger Input Power Converter IR - Intermediate Range kW - kilowatt kWt or kWth - kilowatt thermal LCO - Limiting Condition for Operation LED - Light Emitting Diode LMTD - Logarithmic-Mean Temperature Difference LOCA ' Loss of Coolant Accident LOFA - Loss of Flow Accident LSSS - Limited Safety System Setting MDA - Minimum Detectable Activity aho - inverse ohn MP - Maintenance Procedure MTR - Materials Test Reactor NAA - Neutron Activation Analysis NRC - Nuclear Regulatory Commission NRL - Nuclear Reactor Laboratory
' 0IC -
Operator in Charge 10P -- Operating. Procedure ppa - parts per million PSAR - Preliminary Safety Analysis Report RC - Radioisotope Committee rem - Roentgen Equivalent Man RO - Reactor Operator Radiation Procedure O RP v XX
l l
./ List of Abbreviations (continued)
RPI - Rod Position Indicator
-RPS -
Reactor Protection System, same as Reactor Safety System RRSO - Reactor Radiation Safety Officer RSC - Reactor Safety Committee RSS - Reactor Safety System, same as Reactor Protection System RTD - Resistance Temperature Detector SAR - Safety Analysis Report SHP - (Office of) Safety and Health Programs SL - Safety Limit SOP -- Standard Operating Procedure SP - Security Procedure SRO - Senior Reactor Operator SSR - solid-state relay TLD - Thermo-Luminescent Dosimeter TTL - Transistor-Transistor Logic UCIC - Uncompensated Ion Chamber UPS - Uninterruptable Power Supply URSC - University Radiation Safety Committee UTR - University Training Reactor VAC - Volts AC (alternating current) VDC - Volts DC (direct current) - () VPI or VPI&SU - VTAR - Virginia Polytechnic Institute and State University Virginia Tech Argonaut Reactor w/o - weight percent 4 i XXi
I s q 1. GENERAL DESCRIPTION OF FACILITY V < 1.1 Introduction The Virginia Tech Argonaut Reactor (VTAR) is located in Blacksburg, Virginia. The University is a state-owned, land grant institution located in Montgomery County. The reactor ts an Argonaut type reactor, graphite- and water- ' moderated, graphite.-reflected, and light-water cooled. Authorized power level is 100 kilowatts thermal (kWt). Original op'eration in December 1959 was at 10 kW thermal. In March 1969, license R-62 was amended to increase power to the current 100 kWt rating. Information, analyses, and references presented in this Safety Analysis Report are to show that the VTAR can continue to be operated at i 100 kWt, the current maximum authorized power, without undue risk to the . safety and health of the public. This report will also show that reactor components and systems will be capable of functioning properly for the period of the license renewal. 1.2 General Description The VPI&SU main campus is located in the town of Blacksburg with the facility site located on the northwest edge of the campus. Significant growth has occurred in all areas of the University in the 1970's. Enrollment has risen so rapidly that funding and space availability will limit future growth to no more than three per cent per year over the coming decade. Enrollment in the past five years is as follows: Year Enrollment 1978 20,261 1979 20,780 1980 21,069 1981 21,584 1982 21,510 Additionally, there are some 5,500 faculty and staff members currently employed. These figures should be compatible with the aforementioned growth rates for the next five to ten years. The reactor facility is housed in Robeson Hall. Also located in this building are Physics faculty, staf f and graduate student offices, ' other research laboratories, shops and class rooms which are primarily used for Physics. The building itself has free access; however, entrance to the reactor facility is strictly controlled in conformance with the Physical Security Plan. See Chapter 2 for an overview of the town of Blacksburg, the campus, major access routes, and the facility layout. O v 1
T The VPI&SU campus geological characteristics were well documented in the original construction permit. Principal constituents of the soil ! include an underlaying of shales and dolomites with an intimate l cataclastic mixture of these two rocks, generated by a geologically inactive cross fault. This region cpproaches Robeson Hall at a minimum distance of 400 feet. { The VTAR. is a heterogeneous, enriched-uranium, Argonaut type reactor. - A standard uranium-aluminum type MTR fuel enriched to approxi-7 mately 93 per cent is utilized. Pure light water (H2 O) serves as the j primary coolant and moderator. Additional moderation and reflection is provided by the graphite stack which surrounds the core tanks. Primary control of the reactor is accomplished by 4 boral safety and control rods. Movement is accomplished by a drive motor. and worm gear coupled
~ ; to the control rod shafts through an electromagnetic clutch. Figure 1-1
- shows an overview of reactor closurcs.
I
- ,The reactor core consists of graphite blocks (44" x 48" x 56") in
! which are embedded the two aluminum tanks containing the fuel. Each
- j. aluminum tank contains 6 fuel elements with _12 plates per element. See i Table 1-1 for a listing of VTAR characteristics, including a comparison with similar facilities.
1, Reactor operations are supported by the following systems: (
. (1) Reactor Instrumentation (includes the Reactor Safety System) 4 (2) Primary Coolant System (3) Secondary Coolant System <
l (4) Reactor Ventilation System i j (5) Radiation Monitoring System ' l (6) Aux 1111ary Systems-includes such components as Shield Tank, , j water purification and make-up systems, etc. 1 I An Emergency Cor3 Cooling System and associated instrumentation s (Note:
- are available for testing and training purposes only and are not con- l 2
sidered necessary for reactor operation.) The primary coolant flow path is from the dump tank to the suction l of the primary pump, out the pump discharge and through the secondary ' heat ' exchanger, to ' the dump valve (closed j during operation) into the ! bottom of each core tank up through the fuel elements and out to the , dump tank via an overflow line located near the top of the tanks and connecting back to ' the dump tank. Heat removal is performed by a secondary cooling tower located on the roof of the building and a high l capacity secondary pump in the reactor room.
.j 2 .
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O . O O Table 1-1 Comparison of VTAR, UFTR, UWNR , Parameters VTAR UFTR UWNR General: Modified Argonaut. light water Modified Argonaut light water Reactor type Modified Argonaut, light water moderated and cooled, graphite moderated and cooled, graphite moderated and cooled, graphite moderated & reflected, plate moderated &' reflected, plate moderated & reflected, plate fuel. fuel. fuel. 100 kw, 1.3 x 1012
~
Maximum Power and a . 100 kw, 2 x 10 12 n 100 kw, 1.4 x 10 12 thermal flux (peak)' 2 2 2 Reactor Yessel Stacked block & slab Stacked block and elab Configuration and 22.08 x 17.33 x 9.22 Stacked block and slab 20.31 ft x 15.48 x 11.91< 24.33 ft x 19.3 x 12.0 dimensions - length x width x height (ft) *
, Minimum Critical Moss 6.97 lb (3.17 kg.) 6.77 lb (3.07 kg.) not provided 4
I Containment - type, Confinemgntgoom ! volume, & meterial 3.2 x 10 ft -reinforced confinemgntpilding, Confingmengroom 5.2 x 10 ft , reinforced 1 x 10 ft , concrete t concrete & limestone. concrete Reactor Control: Safety Rods - number, 3 Boron-Aluminum alloy, 3, cadmium, 0.050 A/k 3, Cadmium 0.065 Ak/k , , meterial type, 0.02 Ak/k (approximate totsi reactivity) Safety Rods - shape, Square solid "windowshade", Vane. 0.13 A/k/ min. Vane, .013 g/k/ min. normal reactivity .003 Ak/k/ min. addition rate (avg.) Safety Rods - scram < 0.8 second, gravity drop and < 1 second, gravity drop upon < .6 seconds, gravity drop upon time and mechanism " spring assist" upon angsetic clutch de-energisation. magnetic clutch de-energization magnetic clutch de energization. I Regulating Rod - Boron aluminum alloy, square- Cadmium, Vane, 270/ min (450 Cadmium, Vane 300/ min (450 meterial shape, spad, solid, 27 in/ min, .0015 A/k total), .0098 &/k total), .0078 Ak/k and total reactivity worth (avg.)
O - O 0; } Table 1-1 (continued) Parameters VTAR UFTR lAntt Core and Fuel Datat Core dimensions & 2 rectangular tanks 1.87 ft x 1.74 ft x 4.33 ft., 2 ft x 1.67 ft x 4.0 ft, ' lattice configuration .(es. 1.67 ft x 0.5 ft x 4.83 ft) rectangular grid, 3 fuel boxes rectangular grid, 3 fuel boxes 2 gride; I element wide by 6 wide by 2 fuel boxes long wide by 2 fuel boxes long elements long (es. box ' 4 elements) (es. box 4 elements) Fuel assembly, 12 elements with 12 plates / 24 elements with 11 plates / 24 elemnts with 11 plates / meterial, and element. 93% enriched U-235 element, 93% enriched U-235 element, 93% enriched U-235 enrichment in U ,&l alloy in U,Al alloy in U ,&L alloy 4 Element dimensions 3.0 in. x 5.5 in. x 29 7/8 in. 2.84 in. x 2.14 in. x 25.63 in. 2.8 in. x . 2.4 in. x 25.63 in. l (width x thickness x . 4 length)- . La LPlate dimensions 3.0 in. x 0.08 in. x 26.0'in. 2.3 in. x 0.04 in. x 23.86 in. 2.0 in. x 0.04 in. x 24.0 in. , (width x thickness x length) Plate cladding .020 in., Al .015 in. Al .015 in. Al thickness, material Water channel spacing 0.40 in. .14 in. .14 in. Fuel element heat 116 ft 2 160 ft 2 86 f t 2 ; , transfer area (f < , plates / pins x active plate / pin surface.in ! contact with coolant) r Startup sources IgiPu-Be 1 Ci Pu-Be Pu-Be 2 1.6 x 100 n/cm2-sec i and flux (th) 10 n/cm -see Cf-252 fluxes not provided 25 Ci Sb-Be (flux not provided)
- Reactivity Values:
4 Maximum Excess 0.006 Ak/k 0.023 ak/k 0.013 ak/k , I
. _ . , . - . . ... - ~ . _ - . _. .--. .---- -. .. - _-
O O O Table 1-1 (continued)
~
Parameters VTAR UFTR UWNR Reactivity Values (cont.) Shutdown Margin - > 0. 30 ak/k , > .005 ak /k 0.05 ak/k, 0.027 ak/k 0.06 ak/k, 0.14 Ak/k water & rods in. (with most reactive safety rod (with most reactive rod out) water out & rods in and regulating rod out) , Moderator - -3.78 x 10-5 ak/k/*F, -3.0 x 10-5 ak/k/*F -6.0 x 10-5 Ak/k/*r Void -0.00184 Ak/k/% void -0.0021 ak/k/Z void not providad Experiments 0.003 ak/k 0.023 ak/k 0.011 Ak/k Coolant: e Type, minimum lightwager(demineralized) lightgater(demineralized) lightwater (demineralized) resistivity (steady 0.7 x 10 ohn-cm 5 x 10 ohn cm not provided
- state)
Flow (at 100 kw 40 gym, 160F 40 gym, 160F 25 gym, 270F , AT nominal (at 100 kw) Feak operating fuel 1700, unknown 199.40F, not provided 1450F, 1590F plate / pin temperatures (inside fuel meat) without hot channel factors, with hot channel factors Secondary heat Cooling tower & Pump via Deep rock well (city water City water via primary to removal type primary to secondary heat back up) via primary to sec- secondary heat exchanger exchanger (city water backup) condary heat exchanger
l Automatic shutdown causes the insertion of three control rods and the opening of the dump valve, which drains the water from the core tanks into the dump tank. The following conditions will initiate the above events: 4 (1) Power range high neutron flux (2) Intermediate range period scram (3) High moderator level in core tanks (4) Earthquake
- (5) Primary coolant loss of flow, and
-(6) Additionally, manual scram capability is always available.
Four other design features contribute to make the VTAR inherently safe. They are: (1) A long neutron lifetime (2) Very -low built-in excess reactivity (3) Negative-temperature coefficent of reactivity, and
'O c4) ne.ative veid coef ficient of reactivier.
^ All these factors combine to produce a self-limiting effect in the event of a power excursion as demonstrated empirically and experimentally in previous years. The VTAR is provided with multiple experimental facilities (Chapter i 4). These are: 1
- 1. Two 6 in. (' nominal) beam ports on the north and south sides of the graphite reflector.
- 2. A 4 f t x 5 f t x 5 f t thermal column with 15 graphite stringers.
i
- 3. Five graphite stringers (a 3 3/4 in x 3 3/4 in x 48 in central and 4 - 13/16 in x 3 3/16 in x 48 in vertical) grouped in 'the reflector region between the core tanks, and
- 4. A 2 ft 6 in x 3 ft 5 1/2 in graphite duct from the core to a 5 ft x 6 ft x 11 ft 6 in shield water tank.
Reactor operation, experiments, etc. are supported by the Of fice of Safety and Health Programs (SHP) of the University. The Reactor Radi-ation Safety Officer- (RRS0) serves as that office's safety O-7
\
.3 !
l l
- representative and inspector. Further information on the RRSO's duties,
. (_ responsibilities, and programs are included in Chapter 12.
The VTAR is a versatile research, teaching and training reactor with few environmental effects. Operations pose a minimal hazard to the i safety and health of the public. i 1 1.3 Comparison Tables i l 1 1.3.1 Comparison with Similar Facility Designs The VTAR has been operational since initial criticality in December 1959. The R-62 license was amended in March 1969 authorizing an increase to the current power level of 100 kWt. There are several - Argonaut reactors operating in the U.S. at this time including those at the University of California .at Los Angeles (UCLA), University of
- Florida, Osinesville (UFTR), University . of Washington, Seattle (UWNR),
! and at Iowa State University. Outside the U.S. , Argonaut reactors are I 4 located in Scotland, Japan, Australia and the Netherlands. fi Since the VTAR has several unique characteristics, it will be compared with the UFTR and UWNR (Table 1-1).
] ! Some of the -more notably unique items of the VTAR (with respect to . Argonaut type reactors) are the secondary cooling system (Chapter 5),
i fuel lattice configuration and dimensions, and control rod type. ! Other earlier reactors which also possessed some similar character-(. istics are the original American Radiator and Standard Sanitary Corp., i University Training Reactor prototype (UTR-10), Iowa State Univeraity Reactor (UTR-10), Argonne Low Power Research Reactor (Juggernaut), and several of the SPERT and BORAX cores. e i 1.3.2 Comparison of Final and Preliminary Information i i' l The VTAR has experienced numerous modifications during its history. j The original Hazards Analysis' was insufficient for renewal and a Pre-
! liminary Safety Analysis Report (PSAR) was submitted in' November,1979.
1 However, due to significant changes in facility equipment and design parameters, this document has been significantly rewritten, including the use of a ' standard format. Due to the large number of changes -from j the PSAR, these changes are addressed throughout the SAR. ) For purposes of clarity, a liht of previous R-62 license
, amendments, modifications,'and facility changes are listed in Table 1-2,
[ " Facility History". Is i-i
-O ,
8
O O O
. Table 1-2 " Facility History" Item
- Approval Date Amendmect Application (s) Major changes A. Original application: February 5,1959 A-1 April 1, 1959 A-1. Application for SNN
- for construction ifcense, applicant's technical I Fermit No. CPER-43 qualifications, elaboration on -i equip. & facility descriptions,
; additional procedures to address potential radiological &
i operational mishaps. A-2 July 2, 1959 A-2. Elaboration of experi-i mental programs, facilities, & the hazards associated with them. um Additional analysis of operational
& radiological accidents.
A-3 November 20, 1959 A-3. Further analysis & elaboration of all facets of the i . facility. i A-4 Decemb'er 14, 1959 A-4. Designation of startup responsibilities, VTAR reactor operator training, and ventilation system operating requirements with a limitation on amt. of integrated ;
. power operations.
For A through B: B. Facility License December 18, 1959 B. 10 kw maximum power with f No. E-62 Docket (to empire November 16, 1969) stated requirements for ventila- ; 1 50-124 isoned tion system, reactor startup, i operator licensing, and integrated power limitation. i
~ . . . ____ . . _ _ . _ . . _ . . _ . . . . . . . . _ . . _ . __. __ _ _ _ . . . __ ..
O O - O Table 1-2 (cont.) Item Approval Date Amendment Application (s) Major changes C. Proposal for Amend- December 7, 1965 C. Increase in power to 100 kw meet No. I to (application only) (th), increase excess License R-62 reactivity, use of a se'cond core for low power experiments and incorporate technical specifications. , , C-1 March 5, 1966 C-1. . Power specified for low power core, storage & use of cores, plate temperatures for 100 kw scram, ventilation monitoring, & increase in amount of SNM authorized. For C through C-1: c5 D. Amendment No. I to June 24, 1966 D. Following major License R-62 issued (to expire hovember 16, 1969) modifications made: (100 kw operation) (1) coolant return line , enlarged from 2" to 4" (2) Argon duct installed to center of core. (3) Cadmium shielding of Nuclear Instrument detectors ! i (4) Additional shielding added throughout the Reactor Room (5) Construction of a hot cell (6) Installation of a new venti-lation system E. Proposal for Amend- February 18, 1969 E. Corrections & the addition ment No. 2 to License of Sections on Definitions and R-62 Adatistrative Requirements to Technical Specifications. i E-1 March 11, 1969 E-1. Furt'her corrections and changes to Technical
. Specifications; specification of
{ renewal duration. .
O O O . Table 1-2 (cont.)
- Item Approval Date Amendments Major changes F. Amendment No. 2 to March 20, 1969 F-1 July 20, 1970 F-1. Console relocated to Room
! 108, replacement of tube Nuclear Instruments with Transistorized. F-2 May 23, 1973 F-2. Key actuated reset switch on radiation alarms, appt. of additional members to Rad. Safety Committee, facility admin, organization change, installation ' of additional pneumatic rabbit ! systees. 7 1 S
4 O' 1.4 Identifiestion of Agents and Contractors As discussed earlier many changes were performed on facility equip-ment and syst:.as since submission of the original PSAR. Due to budget constraints, virtually all work was accomplished "in-house" by the Reactor Staff. Design of the new cooling tower was conducted by Carneal and Johnson of Richmond, Virginia. Future retention of agents and contractors is dependent on reestablishment of neutron activation analysis and training services. 1.5 Requirements-for Further Technical Information
- f. This Safety Analysis Raport (SAR) will serve as the final report I
for purposes of satisfying the requirement for a Final Safety Analysis Report (FSAR). All proposed facility changes are included in this l' report since they encompass a wide area of facility equipment and systems. 1.6 Material In.:orporated by Reference i A list of the material incorporated by reference is given in l Appendix 4-1. f l.7 Instrumentation and Control Drawings VTAR Instrumentation and Control drawings (I&C) are shown in Chapter 7. Basic block diagrams only are used. Further. detail'is given in the applicable chapters. l 1 O .
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- 2. SITE CHARACTERISTICS 2.1 Geography and Demography 2.1.1 Site location and Description 2.1.1.1.
Specification of Location The VTAR is located on the campus of the Virginia Polytechnic Institute and State University.(VPI&SU) in Montgomery County. Figure 2-1 shows the'gecgraphic location of Montgomery County in the southwestern I portion of The Commonwealth -of Virginia. Figures 2-2 and 2-3 show the - site location .and the VPI&SU campus- within the town of Blacksburg. The town of Blacksburg is located in the north central part of Montgomery County which covers 395 square miles. The VPI&SU campus is near the center of . the town of Blacksburg adjoining and just west of the main business area. The town of.Blacksburg has a population of 30,638 (1980 d Censu=) which includes the student population at VPI&SU of.about 23,000.
. The total population of the New River Planning District (Montgomery, Floyd, Giles and Pulaski counties) is 141,343.
The VTAR is located in the Ph7 sics Building (Robeson Hall), as i shown in Figure 2-4. Concentric circles around the VTAR with radii of ! 500 ft and 2500 f t ' are . also shown in Figure 2-3. Within the 500 ft. circle, there are several classrooms and offices including those in O Rebeson natt. as idsen aal1 cchemister). rasPtin Hatt, wi111 ass na11 Derring Hall (Biology and Geology) and Cowgill Hall. 2.1.1.2 Site Area Map _ The site map shown in Figure 2-2 indicates the area surrounding the VPI&SU . campus. These boundaries comprise the ~ site boundary lines. An i isometric view of the campus is shown in Figure 2-3 where Robeson Hal1, the-building housing the VTAR, is identified.
. Since the VTAR is a low power research and training reactor, the
~
1 exclusion area (during an emergency) for the facility (as defined in 10 CFR Part : 100) consists of Robeson Hall. In the event of a' reactor
, emergency, the entire building is evacuated and re-entry to the building is controlled by the reactor staff and the Virginia Tech Campus Police.
l 2.1.1.3 Boundaries for Establishing Effluent' Release Limits ; I According to the- regulations of 10 CFR 100, a res'tricted area is
~
defined for the purpose of controlling access to protect individuals from exposure to radiation and radioactive materials.- The restricted area for the VTAR is shown'in Figure 2-4 and comprises the reactor cell , (Rm 10) and the associated. laboratories and electronics shop (Rm 106,7,
.& 8). In addition, the roof of Robeson . Hall is _ part of the. restricted O
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, emergency support center . ,
n===== First Floor Plan
. area since the cooling tower and the upper part of the exhaust stack
. from the reactor cell are located oa the roof. The access door to the roof is kept locked except during maintenance operations and the key is stored in the reactor area.
Two double-doors provide access to the reactor cell from the uncon-trolled area, one from the west side loading ramp and one from the east side leading to a corridor. These doors remain locked during reactor operation and at all other times except when access is required for maintenance or for the transfer of equipment or fuel. A third door from the reactor cell is located on the south side and opens into Room 6 which is part of the restricted area. During reactor operations this door is also locked and access to the reactor cell is monitored by a closed-circuit television camera. On request, this door can be opened for authorized personnel during reactor operations by a solenoid-activa *.ed lock controlled at the reactor console. On the first floor level, doors to Rooms 106 and 108 provide access from the corridor to the restricted area. These doors are always locked. The inner door between Room 108 and 106 is not locked, allowing access to the Neutron Activation Analysis (NAA) Lab. Access to the restricted area is controlled according to the VTAR Security plan. Further details on the reactor room, access procedures, and construction are withheld from public disclosure in accordance with 10CFR2.790(d).
.O 2.1.2 Population Distribution Table 2-1 shows the population distribution by sector (22.5 0 ) out to a radius of 1 mile. Demographic data are based on the 1980 U.S.
Census. 2.1.2.1 Population Within 10 Miles The major permanent population center within 10 miles of the VTAR is the Town of Blacksburg (see Figure 2-5). The town has a total popu-lation of 30,638 with most of the population to the south and east of the reactor site. 2.1.2.2 Population Between 10 and 50 Miles The major population centers between 10 and 50 miles are shown in
. Figure 2-1. In addition to the. Salem-Roanoke metropolitan area (popu-lation 100,220),- small towns within 20 miles include _Shawsville-(1,601),
Elliston (2,291), Price's - Fork (836) and Radford (13,456). - 18
O Figure 7-5 MONTGOMERY COUNTY POPULATION O < j g$ 'O N g ** \ x .et Oo gSO y [e # 10 mile'd amete.
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2.1.2.3 Transient Population Population variations in the Town of Blacksburg and the surrounding urban region are largely due to the VPI&SU students residing in the greater Blacksburg area. The VPI&SU population is mostly transient in its occupation of the campus buildings shown in Figure 2-3. Most of the approximately 23,000 students, together with about 5500 faculty and staff, occupy the campus in varying numbers during the normal work week (8:00 am - 5:00 pm, Monday through Friday). These numbers are fairly constant during the academic year (Sept. 15-June 15) and decrease sig-
! nificantly ~ during the summer months. About 8,950 students occupy on-campus housing; the remainder reside in off-campus housing, mainly in large apartment complexes.- Faculty and staff personnel are considered to be permanent, year-round residents.
2.1.2.4 Low Population Zone The low population zone, as defined in 10 CFR Part 100.3(b), I includes the VPI&SU campus which occupies most of the area within 2000 l feet of the reactor site. The only significant permanent population concentrations within the low population zones are the on campus dormitories (see Figure 2-3). The nearest dormitories are Slusher Hall (population 640); 1500 feet, Hillcrest Hall (population 96), 1400 feet, and Ambler-Johnston (population 1324), 2000 feet, Femoyer (217), Major Williams (337), and Brodie -(334) Halls, approximately 2000 feet. O 2.1.2.5 Population Center
! The nearest population center as defined in 10 CFR 100.3(a) is the Town of Blacksburg. The densely populated portion of Blacksburg is located about 1.0 mile to the south and soutbeast of the VTAR. This 4
distance exceeds the required one and one-half times the distance to the outer boundary of the low population zone as required by 10 CFR 100.ll(a). 2.1.2.6 Population Density Around the VTAR Site t Detailed information on population density out to 3r tiles from the reactor is not considered to be necessary since the VTAR is a small, i self-limiting reactor currently licensed to operate' at a power of 100 kWt. The nearest metropolitan area (Roanoke-Salem) is about 30 miles east of Blacksburg. The next largest neighboring town is.Christiansburg (population 10,345) located 8 miles ' south of Blacksburg. Figure 2-5 shows the population of the various towns and cities around Blacksburg together with circles drawn at 2.5 mile intervals. The specific population distribution around the 'VTAR for dose assessment calculations was obtained f rom 1980 census data. Table 2-1 shows the population'distributim at'a radius of 1 mile and within 22.50 secter bearings. This distribution was used as the basis for defining O i u l l 20 -
k _the " urban boundary" as used in Chapter 15 to evaluate hypothetical radiation doses following the design basis accident. Iable 2-1 Population Distribution Around the VTAR Site Radius One Mile Radius 500 feet Sector Population + Bearing Distance (ft) Population (Hall) (0 -22.500)' 949 00 0 650 (Robeson) 22.5 -45 2025 45 260 600 (Pamplin) 45 -67.5 843 113 240 700 (Williams) 67.50-90 0 1388 158 170 650 (Davidson) 90 -112.5 1067 315 170 1000 (Derring) 112.50-135 69 1350 -157.5 2441 1157.50-180 0 2551 1800 -202.5 0 2 202.50-225 0 0 2250 -247.50 0 247.50-2700 4 2700 -292.50. 215 O 292.5 -315e 3150 -337.50 1128 947 337.50-3600 1012
- Estimated daytime population in classroom building.
+ Based on peruanent residence 2.2 Nearby Industrial, Transportation and Military Facilities ,, 2.2.1 Locations and Routes Several industrial plants are located in the Blacksburg area; these include Federal-Mogul, Inc., a manufacturer of. automobile components and Electro-Tec and Poly-Scientific, both of which produce slip rings. The location of- these plants as well as Wolverine and Corning plants is I shcwn on Figure 2-2. The major highway access to Blackrburg is via U.S.
460. Local air transportation services are provided by the Virginia Tech airport. The nearest commercial airline service is located at Woodrum Field in Roanoke, Va. l l .O
'21
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~
2.2.2 Descriptions 2.2.2.1 Description of Tra nsportation Roates U.S. 460 highway carries the vast majority of vehicular traffic to and from Blacksburg. This highway connects Blacksburg to Christiansburg 8 miles to the south; I-81 is a major route through Christiansburg which runs approximately northeast (to Roanoke) and southwest (to Bristol). Travel to the west is via U.S. 460 which connects with North-South I-77 - at Princeton, West Virginia. 2.2.2.2 Description of the Virginia Tech Airport The major function of the Virginia Tech Airport is to serve the needs of the University. Thus flight training to ROTC students is a principal activity. The university, as well as several corporations and individuals, maintain aircraft for business and personal travel. During the period 1 July 1982 through 30 June 1983, approximately 18,000 operations (one take-off or one landing) were conducted. 2.2.2.3 Projections of Industrial Growth At this time no major increase is projected for industrial develop-ment within the next decade. 2.2.2.4 Description of Military Facilities The Radford Army Ammunition Depot is located about 14 miles from Blacksburg. This facility, positioned in a hollow surrounded by bluffs, produces propellants for weapons systems. 2.3 Metentninav and climatology 2.3.1 Regional Climatology VPI&SU is located in Blacksburg, Virginia which is situated in north central Montgomery County, in the southwestern part of the Great Valley of Virginia. The topography within the county is gently rolling except on the mountains which are located in the northern,. eastern, and southern areas of the county. Elevations vary from 1300 feet'to 3700 feet above sea level but most of the county is around 2000 feet in elevation. The New River flowing northward drains the western area of the county 'and the Roanoke River flowing eastward drains the eastern areas. 3lacksburg has a humid continental type of climate but modified by elevation. Temperatures are approximately 50 colder . than at Roanoke which is_at a lower elevation. The winters are' moderately cold and the summers are relatively cool. The nearby mountains produce various steering, blocking, and modifying ~ effects on storms and air masses. The o 22
county is thus somewhat protected from the weather extremes of winter and summer. Mean annual temperatures at Blacksburg vary slightly from year to year but average about 520, which is near the average of other nearby stations of similar elevation. Rather cool similar temperatures prevail
. 'during the cold season and rather warm similar temperatures prevail
. during the warm season with the spring and fall temperatures showing the upward and downward trends. May and September are relatively warm, even though temperatures below freezing have been observed during both months. Daytime highs during the cold season are in the middle 40's with night-time lows in the middle 20's. Maximum temperatures are in the 70's and minimum temperatures as low as -120 are the extremes during . the winter season. Daytime highs during summer are usually in the low 80's and night-time lows in the upper 50's. . A maximum temperature of 1000 and a minimum temperature of 41 0 are the extremes during July. The number of days with ' the temperature greater than or equal to 900 has ranged from none in several years up to 26 days in 1953. The maximum temperature is below freezing on an average of ,17 days each year. The temperature falls below freezing about 25 days a month during the winter months and has reached zero enough to give an average of 2 4 days a year. The growing season, defined as the period between the average date of the last freezing temperature (April 30) to that of the first freez-O ing temperature in fall (October 8), is 161 days. Freezing temperatures in spring have occurred as late as May 24 and as early in fall as September 18. The growing season is long enough to allow proper matur-icy of a large. variety of crops although not as long as some eastern low 4 altitude locations. Since knowledge of critical temperatures near and below freezing is needed by various users, a separate table (Table 2-2) has been computed showing the probabilities of. occurrence for dif ferent temperatures. Caution should be used in applying these data to other locations in the county as the location of this station may not be representative of other locations in the county. Elevation, air drain-age, soil characteristics. , night radiation, and type of air mass are some of the factors controlling the minimum temperature, sometimes causing large' differences in short distances. Precipitation is well distributed throughout the year with the maximum in July and the . minimum in November. During - this 30-year period, monthly amounts varied. f rom less than 0.5 inch up to 10.29 ; inches during September ' 1957. The highest daily total of over three ! inches occurred during September 1959. Rainfall in summer is due mainly to showers and thundershowers. About 40 days each year have thunder-storm activity which *.s about average for the state. In winter, some of the precipitation usually occurs as snow.- The average is 20 inches a
. year but yearly amounts are extremely variable ranging from two inches in several seasons to 52 inches in 1965-66.
23
The prevailing winds are generally westerly with a more northerly
-component in winter and a more southerly component in summer. The
- topography also affects' the winds with the air tending to flow parallel to the mountain ridges which are orientated mainly northeast .to south-west. Relative humidity varies inversely with temperatures being high !
! in the morning and low in the afternoon. Average humidity values during ' summer are in the 80's early in the morning, dropping into the 50's in ' the afternoon. Cloudiness is least in fall averaging around five-tenths F coverage, and greatest 'in winter, with over six-tenths coverage. Partly cloudy days are most frequent in summer with close to 40% of the days in this category. 2.3.2 Local Meteorology _ 2.3.2.1 Normal and Extreme Values of Meteorological Parameters t A summary of means and extremes of temperature and precipitation are given in Table 2-3, from reference 2.9. j 2.3.2.2 Potential Influence of - the VTAR and its Facilities on . Local Meteorology Since the thermal power generated by the VTAR is currently limited to 100 kWe, no adverse effects on local meteorology are anticipated. During winter months some condensation may occur from the cooling tower which will result in a modest (but visible) plume. 2.3.2.3 Local Meteorological Conditions for Design and Operating Bases Detailed wind d ata are available from the National Weather Service atation located at E-drum Field, Roanoke. This station is located 40 miles northeast of Blacksburg. Wind roses depicting average wind direction and velocity are shown in Figure 2-5 A-F. These figures show wind data for February, April, June, August, October, and December and ' illustrate important seasonal trends. Each wind rose shows the average percent occurrence for direction and velocity grouped into 22.50 sectors. i O r 24 l . m
.-.= y- -
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- 14 0 '
WND DIRECTION FEBRUARY CALM: 7,9 % Figure 2-6A 25
O - 320' 330* 340' 350' O* 10 ' 20* 30* 40' WOOORUM FELD 3'O . 50' h 3,6 % 300 . GO' 8,9 % 4,3 % 8% 7,9 % 'i,6 * . 70* 290 280, ,6 *' 80* O '
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- I40' WIND DIRECTION AFRIL CALM: 9,0 %
Figure 2-6B O
O 320' 330* 340* 350' O* 10
- 20' 30' 40*
WOOORNM FELD 3iO' 50* 300* 60* 3,1
- 7,7 ,4 4%
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WIND DIRECTIO!! JUNE CALM: 17,8 % Figure 2-6C O 27
.O 320* 330* 340' 350' O* 10* 20* 30' 40' WOODRUM FIELO M*
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- WND DIRECTION AUGUST CALM: 15,5 %
Figure 2-6D 1 O i 28 4
- m__-
1 0 320' 330' 340' 350' O' 10
- 20' 30' 40' WOOCRUM FIELD 50*
310 ' o 300* 3,1 % 4 */ 76 9,8 %N 290' 11,0% 70* 5,3 % g 280 2,1% 80"
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- 16 0
- 15 0
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WIND DIRECTION OCTOBER CALM : 16,2 %
. Figure 2-6E i
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- WOOO9bM FIELD 31 0 ' 50*
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- 14 0' WIND DIRECTION DECEMBER CALM 5 13.7 % l 1
Figure 2-6F ! O 30 v i l l
4 Table 2-2 PROBABILITY OF FREEZES OCCURRING AS LATE IN THE SPRING OR AS EARLY IN THE FALL AS DATES SHOWN IN THE FOLLOWING TABLE , i Percent Chance Temperature Levels of Later Date in Spring 160 200 240 280 320 360 90 FEB 7 FEB 22 MAR 17 APR 2 APR 14 APR 20 70 FEB 22 MAR 9 MAR 27 APR 12 APR 23 MAY 2 50 MAR 3 MAR 19 APR 3 APR 13 APR 30 MAY 10 30 MAR 12 MAR 29 APR 10 APR 26 MAY 6 MAY 18 10 MAR 25 APR 12 APR 19 MAY 6 MAY 16 MAY 30 Percent Chance of Earlier Date 1 in Fall i l 10 NOV 7 OCT 30 OCT 20 OCT 3 SEPT 22 SEPI 11 30 NOV 19 NOV 9 OCT 28 OCT 13 OCT 1 SEPT 21 50 NOV 26 NOV 16 NOV 3 OCT 19 OCT 9 SEPT 28 70 DEC 4 NOV 22 NOV 9 OCT 26 OCT 15 OCT 5 90 DEC 16 DEC 2 NOV 17 NOV 4 OCT 24 .OCT 15 2.4 Hydrologic Engineering $ There is free and open circulation and flow of ground water under-neath the Virginia Tech Campus. Some wells drilled on campus yielded polluted surface water and had to be abandoned. There are two under-above ground streams which pass near Robeson-Hall. One passes under . the drillfield about .1,000 feet south of the VTAR flowing in a north-westerly direction. It surfaces 500 feet south-west of Robeson Hall and drains into the Virginia Tech Duck Pond 600~ feet west of the building. The'other stream flows under the campus in a south-westerly direction about 200 feet northmtest of Robeson Hall and surfaces about 300 feet northwest.of the building where it flows into the Virginia Tech Duck Pond.
-At the western end of the pond, there is a six-foot dam. which controls the flow of watcr out of the pond and into a stream which flows west and south . to the- New River, approximately seven miles to the west. The New River generally flows in a. north-south direction in the.
vicinity of Blacksburg. O 31
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i' - f 2.4.1 Flooding
.2.4.1.1 Flood History The effects of heavy rains depend almost entirely on the level of the water table prior to such rainfall and on the permeability of the soil at the time.
. There has only been one significant flooding - type incident in recent years in the Blacksburg area. In August 1969, Hurricane Camille struck the Blac!:.sburg area accompanied by nearly two weeks of almost constant rain. Water levels in the area rose substantially and flooding
- occurred . in localized low lying areas. There was property damage and loss of life in parts of southwest Virginia.
Because Robeson - Hall is constrected on a hill surrounded by high ground, flooding did not effect reactor operation or safety. Howeveri some time af ter the rain period, the West Fuel Storage Pit was apened and it appeared that sore water had entered the pit. There was no danger of a nuclear criticality as the pits are ' designed to contain
- . nuclear fuel suberitical even when flooded completely. The East Fuel
! Storage. Pit showed no evidence of flooding. f There was no standing water in the area which would flood the Reactor Room. It is believed that wat r from the saturated ground
! O seeped into the West Fuel Storage Pit.
seepage is expected, any fuel that is in the West Storage Pit will be Therefore, in the future, if removed and returned to the Reactor Core or transferred to the East
~
j Storage Pit. a 2.4.1.2 Flood Design Because of the location of the VTAR and the low probability of flooding, no special design considerations .were necessary in 4 construction. Water wauld have to be at least four feet deep in the reactor room before there would be any danger of flood waters entering the core. . Criticality would then be unlikely . because i of the fully inserted control rods. In addition, extensive ~ wetting of the graphite moderator / reflector will cause complete shutdown from 100 per cent power operation even if all rods remain out. _ This was proved in 1976 when the core tanks over-flowed during fell power operation, resulting in immediate self-shutdown ! -of the reactor without operator intervention or automatic scram taking place. Changes were subsequently made in the core tank high level scram , . circuitry to preclude recurrence of this event. Since internal water systems are the most' obvious sources of i reactor room flooding, it should be noted that the largest single, water l- ' source in the reactor room, not including the city water supply lines, j o Q 33 -,
. _ _ _ _ . _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ , _ _ _ _ . _ _ _ _ _ _ _ . ~ , . , , , g..... _, ,4_ _, ,.y
s is the secondary cooling system. This system contains about 3800 gallons of water. The process pit can contain up to 2900 gallons. The remaining 900 gallons would then flood the reactor room to a depth of 1 1/4 in.
- 2.4.1.3 Effects of local Intense Precipitation
.During the nearly .two weeks of record rains associated with Hurricane Camille in August 1969, no precipitation effects were noted except for the evidence of small amounts of water inside the West Fuel Storage Pit which was described above. The fuel removal procedure noted deals with any effects from heavy precipitation.
2.4.2 Probable Maximum Flood of Streams or Rivers There is no record of flooding on any of the streams near the VTAR. The New River (410 feet lower in altitude), and the Roanoke River 1 (980 feet lower in altitude), which are at least 5 miles from the VTAR have their own river valleys and flood plains and present no hazard to the reactor. 2.4.3 Potential Dam Failures, Seismically Induced The nearest dam to the VTAR is a six-foot check-dam on tha western end of the Virginia Tech Duck Pond, nearly 1000 feet west of Robeson Hall. Any failure of the dam would result in the water in the pond O flowing out of the pond and in a westerly direction into a nearby stream which . flows to the west, away from Robeson Hall. Such a dam failure presents no conceivable threat to the VTAR. There are several dams on the New River and Roanoke River; however, because of the elevation of
. Robeson Hall and the distance to the rivers, failure of these dams wculd have no effect on the VTAR. .
2.4.4 Probable Maximum Surge and Seiche Flooding Because of the location of the VTAR (over 200 miles from the Coastal Plain and Tidewater regions) this type of flooding would not affect safety or operation of the VTAR. 2.4.5 Probable Maximum Tsunami Flooding The VTAR is over 200 miles from, and over 2,000 feet above, the Tidewater Area. Therefore, no conceivable threat of tidal wave flooding exists in this area. 2.4.6 Ice Effects J , All of the core-related and primary cooling system components of-the VTAR are in the Reactor Room. Some of the secondary cooling system components (cooling tower and support plunbing) are outdoors and subject O
'34. -_.2 - . . - .
to.the co'id winter weather. If the outdoor components contain water and are threatened with freezing, the water will be drained from them. If a , prolonged power failure exists during a period of extremely cold weather, and there is danger of the primary coolant water freezing, the
-dump tank will be checked for contamination and the water released.
Flooding from internal sources was addressed in Section 2.4.2.2. 2.4.7 Cooling Water Canals and Reservoirs There are no cooling water canals or reservoirs associated with the VTAR. 2.4.8 Channel Diversions This is'not applicable to the VTAR. ? 2.4.9- Flooding Protection Requirements Because of the small likelihood of a flood in Blacksburg and the elevation of the VTAR, no such protection is required. 3 2.4.10 Low Water Considerations l The VTAR has a self-contained coolant system which is replenished by the Blacksburg Municipal Water Supply. If that. supply is interrupted so that reactor operations are af fected, the VTAR will be shut down. 2.4.11 Dispersion, Dilution, and Travel of Accidental Releases of Liquid Effluents in Surface Waters The VTAR Primary Coolant System contains 220 gallons of demineral-ized, highly purified water (0.7 megohm resistance) and - is ~ the only liquid which contacts the fuel elements. Should this water become 4 contaminated and escape from the cooling system, it would collect in the sump. Any radioactive solutions would drain into the Virginia Tech Duck Pond which has a check-dam at its western end. The pond- could then be monitored and the dam used to control the - release of the water to the environment. In this way, any radioactive liquid releases could be monitored, controlled and decontaminated, either by being captured in the soil and undergoing ion exchange with the minerals in the clay, or by isolation in the Virgiaia Tech Duck Pond. 2.4.12 Ground Water e Because of the thickness of the dense and impervious clay layer (approximately: 12 feet), it is highly unlikely -that any contaminated liquid, _ in the amounts used by the VTAR, could reach ground water and cause a hazard to the public. However, if this event should occur, it appears likely that the drain would be to one of the underground streams flowing near Robeson Hall from there into the Duck Pond, where it could then be monitored. 35
-,,- , - ,- - ..--,...-r .,- ,
' 2.4.13 Technaal Snecifications and Emergency Operations Require-i ments Because of the size of the VTAR, and the characteristics of its location, no additional procedures or requirements are necessary other than those already specified.
2.5 Geology Seismology, and Geotechnical Engineering 2.5.1 Basic Ceology and Seismic Information
.The VTAR is located geologically on mudstones of the Rome Formation of the Cambrian Age. The Rome as well as adjacent carbonates and braccias of the slightly younger Elbrook Formation are all part of the
, Pulaski Thrust Sheet which overlies the Saltville block. The Price and i McCrady formations of .the Mississippian Age are the youngest units of the Saltville block that were overriddan . by the Pulaski Thrust ' sheet. These Mississippian rocks are exposed along the leading edge of the thrust and within the Price Mountain Window, north and south of the , site, respectively. The Pulaski Thrust Sheet was emplaced and folded during the A11eghenian origenic event in the late Paleozoic. Total movement on the Pulaski fault exceeded 35 miles. Available geologie data show no evidence of movement since the Alleghenian event.
- 2.5.2 Site Geology, The VTAR and Robeson Hall which houses it is underlain by shales
, and dolomites and an intimate cataclastic mixture of these two rocks, which has been generated clong a cross fault which crops out in several j places on the Virginia Tech campus. The bedrock formations are the Rome shale of Cambrian age and the Elbrook formation which is composed of shales, limestones and dolomites. The cross fault is geologically inactive, and no earthquake within historic time has been related either to this f ault or its parent break, the Pulaski Thrust. The rocks close L to the fault are intimately fractured and broken, and. locally the. frag-ments of rock are rolled -.out into a relatively thick body of crash conglomerate. Dissolution of carbonate bedrock has resulted in the development of cavernous conditions. In two nearby buildings, cavities in. the bedrock necessitated some minor changes in. foundation construction. In many places where similar geological conditions prevail, minor cave-ins have occurred as the result of surges in ground-water levels af ter prolonged rains. The rising ground water wets the clay cap, which absorbs water until it-is past the liquid limit, whereupon 'the- surface ground then
- cavec in.
t When Robeson Hall was constructed, -special care was made to ensure that.the foundation was on solid ground. The. reinforced concrete load-10 36 u ____ ______. __ _ _ _. , _ - . . - . . - _ ., _ _ _ . . _ . _ , _ , -- ._. __ ~ _i
l l d 1 i 3
' O ri== w 11- - re a a ever se c 9sx er es re a cie r stiff compact clay which' eventually rests upon a bedrock whose upper surface is a series of pinnacles and shallow depressions. . Only one place in the building area was found'to possess weak ground and that was excavated and filled with concrete. The thick reinforced concrete walls have a bridging effect which would counteract the effect of any minor cave-ins beneath the foundation. Since construction of Rcheson Hall, no -
cave-ins or major settling has ever been observed. 2.5.3 Vibratory Ground Motion With respect to her sister states, Virginia has an intermediate
- level of earthquake risk. Virginia is not as aseismic as South Dakota or seismic.as California. Virginia has experienced. moderate (magnitude 5.8) earthquakes within her borders in the historic past as well as l being affected by larger earthquakes in South Carolina (magnitude 6.6 -
6.9). The State's earthquake activity during the past few decades has , been low, but. persistent. During a recent decade (1968-1978), Virginia residents have felt the vibrations from' 15 small (magnitude < 3.5) shocks. This continuing seismic activity, plus the record from the past, indicates- the possibility that at least a small amount of vibra-tory motion could be encountered.
, During the period 1774-1970, some 128 Virginia earthquakes have l been cataloged. A thorough search of archival data sources (j ournals ,
diaries, newspapers, etc.) resulted in a detailed listing of the effects i O at specific Virginia lo'calities for 100 of these shocks. earthquake known to have occurred in Virginia was in Giles County on 31 The largest May 1897.
~
That shock was felt in portions of 12 states over an area of at least 280,000 square miles. Its maximum magnitude has recently been , estimated to have been 5. 8. Effects on the epicentral area-(30 miles 1 north of Blacksburg) . were structural damage to some brick buldings, rock-slides and . landslides and muddying of springs and creeks. The most recent earthquake to cause damage in Virginia was .on 19 November .1969, .and centered just across the state line at Glen Lyn near Elgood, a small West Virginia community about 60 miles northwest of I l' Blacksburg. Its magnitude was 4.6 and the felt area was 126,000 square miles. The effects at Glen Lyn were: many windows, including display windows, cracked and broken, fallen plaster and minor landslides. There is no record of any damage occurring in the Virginia Tech vicinity. Study of the historical database plus seismographic investigations l of recent shocks have led to the delineation of three seismic zones in l which the majority of earthquakes have - occurred in . Virginia. Faults associated with the earthquakes are thought' to occur at depths of three to ten miles . and their association, if any, with faults mapped by geologists on the surface, is L unknown in Virginia or, for that matter, in the eastern United States. The zones have been named: The Central Virginia Seismic Zone, the Northern l Virginia-Maryland Seismic Zone and I
. . l Os I 37 lt
the Southern Appalachian Seismic Zone, which contains Virginia Tech. These zones are located where earthquakes are most likely to happen. However, the effects from out-of-state earthquakes must also be con-sidered in any risk evalulation. The major example of this type is the 1886 Charlestou, South Carolina earthquake.' The 1886 Charleston, South Carolina earthquake is the largest event known to have occurred in the southeastern United States. It was felt over . 2,000,000 square miles, caused $5,000,000 in damage and approxi-mately 60 deaths. The magnitude has been estimated between 6.6 and
- 6. 9. Large areas of Virginia were affected and widespread minor damage resulted.
A seismic acceleration probability map is shown in Fig. 2-7. According to earthquake risk predictions, moderate damage could occur in the Blacksburg area, with ground motion ranging from 0.018g to 0.14g. This motion corresponds to an intensity of VII on the Modified Mercalli scale. Such motion could cause some cracks in masonry and small landslides. See Fig. 2-8. Robeson Hall, which houses the VTAR is constructed of reinforced concrete and is capable of withstanding an earthquake of magnitude j 6.5. This would result in considerable architectural damage, but structurally the building would remain sound. The intensity of such an earthquake would be almost 10 times greater than the largest earthquake recorded anywhere in Virginia. In addition, the VTAR has an earthquake ! sensor that scrams the reactor when vibrations of 3.0 intensity or accelerations of 0.14g are sensed. The Virginia Tech Geology Department also has at its disposal a standard seismic monitoring station to warn 4 of impending shocks. 2.5.4 Surface Paulting The closest surface fault to the VTAR and Robeson Hall is approxi-i mately 400 feet NE of Robeson Hall and runs northwest to southeast. j Randolph Hall (Mechanical Engineering) and the Newm:" f. brary have been erected on the fault and no_ effects have been observed (for the approxi- + mately 30 years the buildings have stood. The fault itself has shown no , signs of activity within historic time. Also, in this region of the country, no connection has been made between surface faulting and seismic activity. i 1 2.5.5 Stability of Subsurface' Materials and Foundations ! The VTAR and Robeson Hall, which encloses it, are built on a layer of hard, compacted clay which covers the bed rock. - The only area of l soft ground was excavated and filled with concrete. Robeson Hall was 4 built to house the VTAR, and is therefore made of reinforced concrete. O 38'
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Figure 2-7 Preliminary Map of . Horizontal Acceleration (Expressed as Percent of Gravity)_In Rock With 90 Percenc Probability of Not Being Exeeded In 50 Years 39 ____ _
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2.5.6 Stability of Slopes
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l I Robeson Hall and the VTAR are located at the top of a small hill on campus. There are no land areas in the immediate vicinity higher than the building, so slippage of land down onto the reactor is not a pro-blem. The hill upon which the VTAR sits is gectly sloped and covered with concrete and vegetation. A 6 inch soil layer is underlain by hard clay. The hill slopes are essentially stable and no sliding or other acvement has been cbserved. 2.5.7 Embankments and Dams i i' There are no embankments of any consequence around the immediate area of the VTAR. The condition of any of the small embankments around the. reactor would have no effect on safety or operation. The nearest dam to the VTAR is a six-foot concrete dam about 1,000 ft west of Robeson Hall which forms the west end of a small pond. The failure of the dam ' would have no effect on the safety or operation of , the VTAR as the water released would drain into a stream bed and would then flow in a westerly direction, away from the VTAR. 4 O 1 v 4 11
x
- 3. DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT AND SYSTEMS I
, This chapter identifies, describes and discusses the principle )
architecturai . and design features of the building housing the VTAR ; facility. The building and the integral structural systems are the only i features considered in this chapter, while all the systems dealing . directly with the reactor are covered in other chapters of the SAR. l 3.1. Structural Design The reactor is situated in Robeson Hall in the northwest corner of 3 the main campus. The building has a total of four stories and is "L" ) shaped with major dimensions of 220 f t by 50 ft and minor dimensions of 150 ft by 65 ft. The reactor cell is 40 f t by 40 f t by 22 f t high and is located in the northwest end of the building. The rest of the building is used for laboratories, faculty offices, and classrooms. The floor plans for all levels, ground through 3rd floor, are shown in figures 3-1 and 3-2. The building is constructed of concrete columns and beams with l hollow cement block curtain walls and metal sash windows. Tha cement block walls on the inside are painted with enamel while the outside walls are covered with locally obtained limestone blocks 6 in thick to match the rest of the buildings on campus. The floors and roof are poured concrete slabs covered with either asphalt or vinyl tile. O The reactor room, while an integral part of the building, is of slightly different construction. The floor is a minimum of 12 in thick overall and 18 in thick under the reactor. The walls are constructed of
- - 12 in thick re-enforced high-density concrete all around. The outside valls have 6 in of limestone bloch covering 6 in of concrete block, and 1 ft of reinforced high density concrete for a total thickness of 2 ft. The inside walls have 6 in. of concrete block covering and I ft of re-anforced high density concrete for a total thickness of 1 1/2 ft.
The ceiling is 18 in concrete with additional thickness placed over the center of the reactor. There is a total of 24. in of concrete in this area. See Figure 3-3. The reactor control console is located on the first floor in Room ) 108. A 3 ' 10" by l'10" plate glass window is provided overlooking the '
-reactor room from the south end. The personnel entrance to the reactor
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I toom and any area not visible from the observation window are monitored by remote cameras at the control console. The reactor room has three entrance / exits (see Figure 3-4). A large double door on the west end leading _to the outside of the' building is used for transporting large equipment and can be opened only_from the inside. A smaller double door is available on' the east end leading to the hallway outside the reactor room. This access is . also used - for O 1 a 42 I
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pv large equipment. Similarly, these doors cannot ha opened from the outside and are kept locked at all times. The last access is the personnel entrance / exit which leads to an anteroom below the control . room. This. access is monitored by a remote camera and is equipped with + a remote door latch controlled by the reactor operator during operation. 1 The large freight doors, which are 14 ft wide by 11 ft high, are i i maintained closed and locked at all times during reactor operation and l are opened only during the actual transfer of equipment or special maintenance activities. This also applies to the smaller, 6 ft wide by . 7 f t high, freight doors on the east end of the reactor room. An air handler / heater platform with dimensions of 8 f t by 10 f t is located in.the southwest corner of the reactor room. It is built 10 ft from the floor to provide headroom for equipment and personnel 5 A 10-ton bridge crane is provided.for handling heavy equipment, the fuel transfer cask, and shield blocks. Adequate clearance is available to permit the use of equipment necessary for fuel transfer operations and for installation or movement of experimental equipment. Fuel storage pits are available for both irradiated and new fuel. These pits are located in the northeast and northwest corners of the reactor room. Each pit consists of sixteen-6 in (diameter) X 6 ft steel lined holes embedded in the floor. A concrete plug fits into each hole and each pit has a 1/4" thick locked steel cover. Convenience outlets, both 115V single phase and 220V, three phase, ,~ . are located throughout the reactor room. Tap water is available in the vicinity of the process (equipment) pit. Penetrations through the walls of the reactor room have- been kept to a minimum. With the exception of the doors, all are non-movable, relatively small, and are either fully embedded in concrete or are sealed with mastic filler or a metal cover. Table 3-1 lists all pene-trations into the reactor room. The cooling tower is mounted on the far end of Robeson hall above the old accelerator lab. The cooling tower was mounted in this location due to roof loading considerations; there are several heavy beams that can easily support the cooling tower and its water inventory. 3.2 Utilities and Services 3.2.1 Ventilation i The reactor rcom is kept at a pressure slightly lower than atmos-
- pheric through the use' of a centrifugal blower fan and booster fan All duct work is on the suction side. of the ventilation
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~ fans to contain any leaks that may ' occur. Refer to Chapter 9 for a complete description of the ventilation system.
3.2.2. Fire Protection-The Virginia Tech Safety and Health Programs office maintains an - active, ongoing fire protection plan, including regular inspections. i l 47 l
l l O Table 3-1 VTAR REACTOR ROOM PENETRATIONS Penetration Inside location from floor Outside Access double doors east wall, south end floor level hallway 3" hole (viring) east wall, center 9'7" hallway, above ceiling 2" pipe (water main) ' east wall, center 9'7" hallway, above ceiling 1" pipe (water main) east wall, process pit 6" below floor below hallway 3" sewer tape east wall, process pit l'8" below floor below hallway 1 1/4" pipe (drain) south wa'.1, process pie l'8" below floor to sewer 6" secondary pipe east wall, north and 8'9" hallway, above ceiling 6" secondary pipe east wall, north end 8'9" hallway, above ceiling 1 1/2 conduit east wall, north end 9'5" hallway, above ceiling 3/4" conduit east wall, north end 9'5" hallway, above ceiling 1/2" plastic pipe north wall, west end 4'9" ti room old ventilation duct west wall, north end 13' outside old exhaust duct west wall, north end 13' outside freight doors west wall, center floor level outside 1" conduit west wall, south end at floor, on I beam old console room, 1" conduit east wall, south end at floor, or I beam old console room 2" pipe (heating) south wall, west end 14' shop (room 6) (supply) 2" pipe (heating) south wall, west end 14' shop (room 6) (return) 3/8" control line south wall, west end 13' shop (room e) ( 1/2" conduit south wall, vest end 13' control room (room 18) stack duct (18" dia.) south wall, west end 13' control rom (room 18) door south wall, west end floor level shoo (room 6) contain pressure switch for reactor room vent. 4' window south wall, center 16 1/2' control room (108) window south wall, center 4' old console room (room 6) 3/4" conduit, lighting south wall, center 15' control room (room 108) 2-3/4" conduit, control south wall, center 15' control room (room 108) 2-3/4" conduit, south wall, center 15' control room (room 108) control 2-3/4", signal south wall, center 15' control room (room 108) 1/2" conduit south wall, center 17' control room (room 108) rabbit tuber hole south wall, center 12' old control room (room
- 6) ,
1/2" conduit, reactor south wall, center 12' old control room (room I 6) on light floor trough south wall, east end floor level old control room (roon 6) entrance door south wall, east an.d floor level ante room (room 6) 3/4" conduit south wall, east end 4 1/2* ante room (room 6) panel supply conduit south wall, east end floor level switch room (A) sealed I" conduit south wall, east end floor level old control room (room 6) N.I. trough south wall, center 12' control room (108) 3/4" conduit south wall, center 12' control room (108) j a 48
l l Due to the materials used in construction of the reactor building (virtually all concrete), a serious fire in the reactor room is con-sidered unlikely (Chapter 15). The possibility of an internal fire in the core region cannot be ruled out entirely. This could be prompted by an experimental failure (Chapter 15), however, this possibility is considered remote due to the administrative controls and Technical Specifications employed which address sample and experiment limitations (Chapter 16). Fire : extinguishers are maintained in various rooms throughout the facility (Table 3-2). TABLE 3-2 VTAR Facility Fire Extinguishers Application Location Type Weight (1bs.) Rating (Class) Rm 108, Console CO 2 10 108:C B,C i Rm 106, North Wall CO 2 15 10B:C B,C Rm 106 CO 2 10 108:C B,C Reactor Room South Wall Dry Chemical 10 4A:608:C B,C East Wall CO 2 10 108:C B,C North Wall Dry Chemical 10 2A: 30B:C A,B,C O North Wall West Wall CO2 Water 15 N/A 108:C 2A B,C A Top of Shield Water N/A 2A A These extinguishers are checked on a regular basis. Additional fire extinguishers are located throughout Robeson Hall. The Virginia Tech campus has maintained a longstanding agreement with local organizations for fire response. This is addressed in the VTAR Emergency Response Plan. The VTAR Emergency Plan also includes an overview of fire response, organization, emergency action levels, and training for facility and local organizations. l 3.2.3. Flood Protection Due to the VTAR location and accumulated weather data (Chapter 2), the probability of flooding is virtually nonexistent. A continuous rainfall for 14 days did result -in leakage into the west fual storage pit (Chapter 2). in the event seepage seems likely, fuel in the wert storage pit will be transferred to the east storage pit or returned to the reactor core. No criticality would result in the event the storage pits are flooded (Chapter 15). O The VTAR Emergency Plan also addresses require-ments for Severe natural phenomena. s' 49
+
- 4. REACTOR 4.1 Summary Description 4.1.1 General Reactor System Description The VTAR is an Argonaut type research and' training reactor originally designed and installed by American Standard Nuclear Division.
The reactor has been in operation since June 1959. Originally licensed
- for'a maximum power level of 10 kW(th), the reactor was modified and the license (R-62) amended to allow a maximum power of 100 kWt in March 1969.
2 The VTAR core is heterogeneous in design, using 93 per cent enriched uranium-aluminum matrix fuel plates. Figures 4-1 and 4-2 show cutaway views of the shielding and the. core area. Figure 4-3 shows an overall view of the experimental facilities available around the reactor, as well as a general layout of the reactor cell. Thermal power output of the VTAR is currently limited to 100 kWt with water used as a coolant and as part of the moderator; operation at 500 kWt is anticipated in the future. The remainder of the moderator
- consists of graphite blocks which curround the boxes containing the fuel
, and the water moderator. The fuel is contained in MTR-type plates assembled in elements. Each element is composed of 12 (or 13 at a power level of 500 kWt) fuel plates, each of which is a sandwich of aluminum clad over a uranium-aluminum " meat". The reactor core has a two-slab geometry and is presently composed of 12 fuel elements arranged in two water-filled aluminum boxes which dre surrounded by reactor grade graphite. The primary coolant, demineralized water, is pumped upward over the i fuel elements, then fed by gravity through the overflow orifices to the heat exchanger, where reactor heat is transferred from the primary coolant to the secondary coolant. Heated secondary coolant is pumped from the heat exchanger to a cooling tower, located on the roof of Robeson Hall, where the heat is then dissipated to the atmosphere. The reactor is equipped with four " window shade" type control l blades, each consisting of a boral blade attached to a flat spring. As the rod is withdrawn, the spring is wound on a drum by the drive motor. . The motor is attached to the control blade drive assembly by a magnetic clutch. When the clutch is de-ene rgized for any reason, the control blade is forced back into the core by the combined gravity and spring
- forces. Drives are connected to the blade assemblies by long shaf ts.
This connection allows each drive to be mounted outside the shield for accessibility. The maximum reactivity addition rate of the safety and control blades is limited ~ to 0.02% ak/k per second by system design to prevent 50 ji . . - - _ - - -. . -- . . . .. -
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O sudden 1erse reactivitr increase . maximum overall excess reactivity of 0.60% ak/k ensures that there can This 11mitatten. ceuPt ed with a be no chance of prompt criticality. The nuclear design of the core also ensures that the combined response of all reactivity coefficients and an increase in reactor power yields a net decrease in reactivity. This asset is explained more fully in Chapter 15. Operation of the reactor is monitored and controlled from a two-tiu.d desk-type console. The console displays such parameters as power level, flux, period, flows, temperatures and all other parameters neces-sary for safe operation of the reactor. The control console has six channels of nuclear instrumentation. It also contains the reactor protection system, startup interlock system, and a schematic representation of the cooling system with con-trol and indication at key points (mimic bus). A detailed description of the instrumentation and control system may be found in Chapter 7. Experimental facilities available in the VTAR are: (1) Shield tank on east face of core (2) Two (2) six-inch beam ports on the north and south faces of the reactor (3) One central penetration into the center of the core area. ' This penetration also contains a graphite sample block for up to 49 3/5-dram vials (4) Three top penetrations into the core area (1 in x 2 in each) (5) 4 in x 4 in removable blocks in the thermal column. (6) Pneumatic " rabbit" receiver in central region of thermal columns (7) Pneumatic " rabbit" receiver in central region of core. Shield plugs are normally inserted into these facilities .inless the facilities are in use. 4.1.2 Design and Performance Characteristics The general design and performance characteristics for the VTAR are outlined in Table 4-1. The VTAR self-limits the maximum power release in a power excursion by means of a negative temperature coefficent and a negative void coefficent. These features are inherent and therefore effective even if all safety systems fail. The worst situation occurs if a large amount of reactivity is added suddenly. The maximum excess 54
y O Table 4-1. Nominal VTAR Characteristics General Features Reacto r Type. . . . . . . . . . . . . . . . . . . . . . . . . . . . . He te rogeneous , The rmal Maximum Power Leve1. . . . . . . . . . . . . . . . . . . . . 100 kw Maximum Thermal Neutron Flux in Central Irradiation Stringer.............c....... 1.3 x 10 12 N/cm2 see Excess Reactivity (@ 750F ) . . . . . . . . . . . . . . 0. 60% ak/k maximum Clean , Cold Critical Mass. . . . . . . . . . . . . . . . 3031.45Ggams Prompt Neu:ron Lifetime.................. 1.3 x 10- see Void Coefficient......................... -0.184% ak/k/1% Void Temperature Coef ficient. . . . . . . . . . . . . . . . . . -0.004% ak/k/ F Reflector................................ Graphite Moderator................................ Pure Water, Graphite Startup Source........................... 1 Curie Pu-Be Fuel Fue1.................................... 93% Enriched, U-Al Fuel Loading. . . . . . . . . . . . . . . . . . . . . . . . . . . . . Approximately 3. 00 Kilograms Plate Thickness......................... 0.80 in Plate Width............................. 3.0 in Plate Le ng th. . . . . . . . . . . . . . . . . . . . . . . . . . . . 26. 0 in Plate Separation........................ 0.40 in Cladding Thickness. . . . . . . . . . . . . . . . . . . . . . 0. 020 in () Fuel Composition......................... 14.0 w/o U 235 Primary Coolant Typ e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . De mi ne ra lf z e d Lig ht Wa t e r Flow @ 100 kw. . . . . . . . . . . . . . . . . . . . . . . . . . . 42 G PM US (Nominal) Equilibrium Core Inlet Temp. (@l00 kw).. 840F Equilibrium Core Outlet Temp.(@l00 kw)... 1000F Secondary Coolant Ty p e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . H20, Po ta b le , Tr e a te d w/ Potassium Chromate Flow 0 100 kw........................... 400 GPM US Equilibrium Heat Exchanger Inlet Temp... 70.00F Equilibrium Heat Exchanger Outlet Temp.. 71.70F Control Blades Ty p e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Bo ra l , Wind owshad e Numb e r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 S a f e ty , i Regu l a t ing Material................................. Boral Shim & Saf e ty Rod Dimensions. . . . . . . . . . . . 7" x 7" x 0.125" Regulating Rod Dimensions. . . . . . . . . . . . . . . 2" x 2" x 0. 025" Shim & Safety Rods Worths. .. . . .. . ..... ... Approximately 0.65% ok/k Re gula ting Rod Wo rth. . . . . . . . . . . . . . . . . . . . . Approximately 0.12% Ak/k Maximum Reactivity Insertion Rate....... 0.02% ak/k/see reactivity for the VTAR is limited to 0.6% ak/k. Calculations by Tuley and SPERT data indicate that an uncontrolled reactor period of 3.2 msee is required to produce the heat necessary to melt the fuel. A step ; insertion of 0.6% ak/k will result in a period of 200.0 msec at the 55 1 y mn mm, m---
I - l
- VTAR. There is therefore no danger of a fission product release or l
4 damage to the structural integrity of the reactor due to a large reac- , , tivity addition. Reactivity accidents .are discussed further in Chapter 15. Reactivity control is provided by two safety blades, one shim blade and one regulating blade. Table 4-1 shows the worth of each blade as i veil as the maximum allowable reactivity addition rate for the VTAR. The shutdown margin available with the most reactive blade out is approximately 1.46% Ak/k. The control blades are " fail safe" in that they will drop into the core by gravity and spring force in the event of a loss of electrical power. The reactor protection system provides a series of control blade interlocks and reactor scram conditions which both prevent occurrence of situations which endanger the integrity of the reactor system and assure safe operation. The protection system is discussed in greater detail in Chapter 7. Because changes were made in the heat removal system (i.e.: chang-ing from city water cooling to use of a cooling tower), an in-depth analysis has been performed for cases ranging - from 50 kW to 700 kW. This analysis is discussed in section 4.4 Thermal and Hydraulic Design. l l 4.1.3 Shielding ! Biological shielding is provided around the VTAR to provide levels which are as low as reascnably achievable. The biological shielding is t O aae of c c-t viace 6 Ft - c cret- te while assuring its effectiveness. 4 c <* e< ra11 twic* c - The shielding consists of the following: 5
- (1) 6 ft 8 in cast-in place barytes concrete on north and south sides.
J (2) 6 ft 10 in cast in place barytes concrete on east and west sides (augmented by shield tank on east side). i (3) 7 f t 1 in removable barytes concrete blocks on top. j (4) 2 in lead gamma curtain on west side between core and thermal column. (5) 2 in. stacked lead block on east side of shield tank. ! Snielding at various points around the reactor is augmented by removable 2 ft x 4 ft x 8 ft thick standard concrete blocks as shown in Fig. 4-2. l Access to the west end and to the top of the reactor is provided by 1 , removable concrete blocks which were cast to fit these openings. These l
- blocks, which weigh ' up to 11,500 lbs each, are equipped with pick up j luga so they may be handled with the overhead crane. Bleck and shield
!_ arrangements are depteced in Fig. 4-2. . -O ! 56 1
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4.1.4 Experimental Facilities The -irradiation facilities and instrumentation ports are shown in Fig. 4 -3. There are four 2 in x 1 1/2 in and one 6 in x 6 in vertical experi- c ment holes centrally located between the core tanks where the flux is highest. One of the 2 in x 1 1/2 in ports is permanently utilized for the central pneumatic rabbit system. Access to these ports is accon-plished by removal of the top shield blocks. The 6 in x 6 in port is equipped with a graphite-sample irradiation block. Access to this block and sample port is accomplished by use of a small hand crane to lift the five-foot steel and borated parafin plug which is normally in the port. A thermal column is provide.d with one 6 in x 6 in port and fourteen 4 in x 4 in removable stringers. Three of these stringers are currently utilized for nuclear instrumentation. The 6 in x 6 in port is used for a second pneumatic rabbit system. The thermal coluan itself is 48 in x 60 in x 60 in. Two other horizontal openings, 6 in diameter, are located on the central plane of the reactor, one on the north side and one on the south side. These horizontal ports are particularly useful where a beam of neutrons is required or for irradiation of large samples. O A water tank is placed on the east face of the reactor opposite the thermal column. This shield tank is constructed with high density concrete and is 5 f t x 6 f t x 12 f t high. It is usetul for large, bulky experiments requiring well thermalized neutrons. 4.2 Fuel System Design The reactor core has a two-slab geometry, composed of 12 fuel elements arranged in two aluminum water-filled boxes, as shown in Fig. 4-4. The arrangement of the fuel elements in each core tank is shown in an isometric view of a single core tank in Fig. 4-5. The coolant / moderator inlet and outlet are also shown in Fig. 4-5. Each fuel ele-ment consists of 12 plates, manufactured from 93% enriched uranium-aluminum alloy. The fuel is of the MTR type, in which each plate is a sandwich of aluminum clad over the uranium-aluminum alloy. Figure 4-6 is an illustration of plate construction and fuel geometry. A sheet of 0.020 in thick aluminum provides the cladding. Each fuel plate is 26 in long by 3 in wide with a' total thickness of 0.080 in and each plate contains 22 grams (nominal) U-235. Each fuel element is composed ' of 12 plates as shown in Fig. 4-7. l The plates are separated by a gap of . 0.40 in through which the coolant / , moderator flows. Coolant / moderator is pumped upward from the bottom of I the core tanks, flows through the elements, removing heat, and is then 4 g gravity-fed through the outlets on the top of the core tanks. V ! t- ! 57 )
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O . O 26" 4 O _ _
.m,_,
_ m
, , y ,_,_. .p 3
Figure 4-7 Plan View of Fuel Element 61
.t - ,
/
i-i l 4.3 Nuclear Design 4.3.1 -Flux Distributions The principal nuclear parameters for the VTAR are listed in Table 4-1 in Section 4.1.2. The experimentally obtained (in 1961) neutron flux distribution for the VTAR during 100 watt (thermal) operation is shown in Fig. 4-8 through 4-11 as a function of relative flux. Since relative flux is a percentage of-absolute flux, the absolute flux may be , found by l (ABS (*) " l x (1.28 x 10' 2 ) em - see where a relative flux of 100 is equal to the absolute flux in the cen-tral stringer. Figure ~4-12 shows the neutron flux distribution in a
- representative fuel element in
- a positon adjacent to the center of a
- , core tank. Figure 4-13 shows the neutron flux distribution along the vertical (Z) axis along the northeast stringer of the internal reflec-
}' cor. Distributions shown in both Fig. 4-12 and 4-13 are considered to be representative of flux distributions throughout the reactor. j In a study of the performance characteristics of the VTAR, Stam noted that the flux is not symmetrical about the midplane of the reactor. This condition may be easily noted in Fig. 4-9, which shows O the erase fiu 1 the e e 1ah to he c e 14erab17 tower tha that 1-the west slab. According to Stam, the reasons for the approximately 15% flux tilt are the presence of the half-withdrawn shim rod and greater leakage at the east side of the core. Though it is felt that there may be other contributing factors as well as those noted by Stam, this tilt seems to be common among Argonaut type reactors. It is felt that this anomaly should be investigated thoroughly prior to 500 kWt operations as a significant temperature difference could occur'between the two tanks. 4.3.2 Xenon Transients The xenon transient during a ten-hour run at 100 kWt was observed
- beginning four hours af ter start-up. The observed power coefficient is shown in Fig. 4-14 where Ap is plotted as a function of time, t. Since i the observed transient reflects both the xenon transient and the graphite heating effect, a correction for the latter was applied. With
- this correction, the curve labelled " Pure Xenon Transient" on Fig 4-14 i was obtained. The calibration curve for the regulating rod used in the
- experiment'is shown in Fig. 4-15.
- The xenon transient expected for operation at 500 kW
- : was calcu- -
laced with the XETRAN computer module and was generated by increasing i the flux level used in the experimentally-fit (100 kWt) program by a factor of five. Results are shown in- Figs. 4-16 and 4-17. The peak in negative reactivity during a standard week of 500 kWe operation (seven i . hours per day for five days at 500 kWt and seventeen hours shutdown) was l 62 I
. I
=
t ! L ! $ ! .' , : , O . s. 2 I 1 _ 0 11 18 3 . w I6
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O I 24 . . Total
. Epithermal 20 - o Thermal l
16 12 8
~
4 - E b f0 -- - - - - - U -4 . E B -8 - 5
-12 -
-16 -
-20 -
-24 f f f f I t I f f i g 30 40 50 60 70 80 90 100 0 10 20 Relative Flux Figure 4-10 Vertical flux profile along central stringer of internal reflector.
O 65
l O r I Position of Shim Rod 110 - o39.5% Withdrawn
*50.It Withdrawn 100 -
I x 3 90 - a m k 80 - Oi y 70 - I O e er ce raverse 60 - I j 50 - a f f I f f ! f f e i 3
-12.5 7.5 -5 -2.5 0 2.5 5 7.5 10 12.5 Distance from Center - in.
Thermal Flux at Face of Duct in Shield Tank FIGURE 4-11 1 O . 66 4
-n, --
, n - -
, , , ----,w -
----w,---
O 100 - Midplane l 90 -
- 1 Below 80 -
Sh" Above .. . 8
- 70 -
u m O c e t 3 60 - 1 E I l l 50 - I { 11" Below i 11" Above 'I j '
~
i i e s e e e i I 8 l ! ! !
' ! ! ! l ! !
30 -2 -1 0 1 2 3 4 5
-5 -4 -3 l Water Gap Number Figure 4- ' Flux profile across element W-3 (E to M) .
> .O ) 67
J O , 18 - 15 - 12 - 7 i
$9 -
o 3 4 6 - w 3 E 3 t o O 2 E 0 2 E g -3 B w m j -6 o o 0w
-9 o
-12
-15 -
-18 l t t t t t 0 20 40 60 80 100 Relative Flux ($rd}
/T (j Figure 4-13 Vertical flux profile in NE stringer.
l 68
O O O
~
l l 1 i
.040 -
y/ j
/ '
I
'030 - !
i -
/
I: 4
/
xs '020 - / Y [ l, 8 -
.010 -
l t i i I I I I i 4 5 6 7 8 9 10 (CM M ) - -PLRE XENON TRANSENT OBSERVED TRANSIENT } Fiqure 4-14 Measured Reactivity Transient 4
O O O e i
.13 -
i
.12 -
i , . .i t - i ^
' .10 -
x N y .O9 - ! <3
! .Os -
i u. .N. . O w z .07 i-1 m .Os -
! o 3 .05 - . o i O .04 -
i E j .03 - i
.02 --
i i .01 - 1
/ -
I I I_ I i i 1 ] 2 4 6 8 10 12 14 16 1 INCHES WITHDRAWN i' l Figure 4-15 Regulating it.nl Caillisation f 1
~
O O O o XENON REACTMTY TRANSIENT OPERATE OPERATE OPERATE OPEHATE OPERATE 7 hts, , 7hr s , 7 hts , ,7 hrs SHUTDOO4, SHUTDOWN SHUTDOWN SHUTDOWN SHUTDOWN (WEEKEND) 17hre 17has 17 hrs 17 hre
.50 -
Q A5 - SOOKw OPERATION
- N M .40 -
<l g .35 -
y ~
> .30 -
, p 5 .25 - 6 o .20 -
<t W
(C
.l 5 -
' * ^I
.lO -
i'
.O5 I i l i i t I i i i l iI i i i I e i e e e e 6 12 18 24 30 36 42 48 54 60 66 72 78 84 90 96 102 108 114 120 126 132138 TIME (hours) 9 .
Figure 4-16 Xenon Transients (100 kW and 500 kW) b
O O O XENON REACTIVITY TRANSIENT 500 Kw OPERATION ! .90 ' - 27 hr OPERATION
, ' . - -~ ~ ~ - [
( .80 - x .
's,
's,
<1 .70 -
b .60 - 14 hr OPERATION ' y F _- .50
> .40 -
7 hr OPERATION l 4 i D
<t w
.30 -
/- -
* .20 -
.10 -
1 I I I I I I I I I I I I i 1 l l l l 2 4 6 8 10 12 14 16 18 20 22 24 26 28 30 32 34 36 38 40 ) TIME (hours) i j Figure 4-17 Xenon Transients (500 KW) i i i
4 s found to be 0.47% Ak/k. If longer periods of operation are needed, it can be seen from Fig. 4-17 that 14 hours of operation will result in a reactivity loss of approximately 0.6% ak/k.
-4.3.3 Other Reactivity Considerations
! Measurements of the temperature coefficients in the VTAR have been made during extensive operations at 100 kWt. The combined coolant and fue 10 g temperature coefficient ( g) was found to be -(5.1
- 0.5) .c Ak/k/ F The graphite coefficient (a ) was measured to be +(2.3
- O.2) x 10-7% ak/k/*F.
With .the highly-enriched uranium (HEU) contained in the U-Al alloy in the present fuel elements, the fuel tuaperature coefficient of react-ivity (Doppler coefficient) is .much less than the moderator / coolant coefficient. For the present analysis. the fuel temperature coefficent can be assumed to be zero. The modified cooling system (for eventual 500 kWt operation) is designed to maintain the temperature profile in the reactor core very near the existing 100 kWt profile. Thus an estimated net positive reactivity of approximately 0.016% ak/k will result for operation at 100 ! kWt. 4.4 Thermal Hydraulics
$1=c- t. <*- v'^ 8 :) O element configuration, as noted previously.
e t et = tea <* 12 at =- <t For this reason, such . parameters as heat transfer coefficients, heat balances, rod worths, flux levels, etc., have been well known quantities for many years. I The only difference between the 12 plate and the 13 plate elements is the width of the cooling channels. The coolin'g channels are 0.40 in x 3.0 in in the 12 plate configuration and 0.36 in x 3.0 in in the 13 ! plate configuration. The resultinh change in Reynolds Number is well I within the uncertainties of the calculations, primarily because the flow is in the boundary between laminar and turbulent flow. In this flow I regime, the effects of locally induced turbulence or a return from turbulent to laminar flow can create uncertainties greater than the j differences between the two above cases. j Operational experience since 1969 with the present 12 plate i configuration and the small dif ferences between 12 and 13 plate thermal hydraulics lead to the _ decision to include the 13 plate thermal hydraulics study for the VTAR. A second reason for this court.e of action is the higher fuel loading for the 13 plate case. For these reasons it was felt that the 13 plate configuration produced more conservative results in accident analysis. The 13 plate study was used in the following sections. l . O I ! 73 l
\
s.-, , _ _ _ . _ _ - _. . _ - . . _ _ _ - . . . _ . . , . , . . . . _ _ _ . . ,.
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4.4.1 Primary Coolant System The major modification to the primary coolant system was the installation of a larger primary coolant pump. The primary coolant flow rate was increased to a nominal 40 gpm with the capability of reaching
>90 gpa. The resulting temperature rise across the core will be approx-imately 16?F at a power of 100 kWt.
4.4.2 Intermedia e Heat Exchanger and Secondary Coolant System The secondary cooling water in the pre-existing facility was town water. Af ter a single pass through the secondary side of the heat exchanger, the water was discharged into the local sewer line. The original heat exchanger has been replaced with a water-to-water parallel plate type heat exchanger. The plate material is type 304 stainless steel. The primary coolant makes two passes (5 coolant chan-i nels per pass) through the heat exchanger and the secondary coolant makes a single pass (11 coolant channels per pass) through the heat exchanger. A secondary coolant pump has been installed to provide a secondary flow rate of up to 685 gpm. 4.4.3 Cooling Tower The largest single component in the upgraded system is the cooling tower. The cooling tower is a mechanical draf t type capable of dissi-pating up to 1 MWt to the atmosphere. 4.4.4 Emergency Core Cooling System An emergency core cooling system (ECCS) is one of the engineered safeguards included in the modified facility. The ECCS is able to provide emergency core cooling in the event of certain classes of loss of coolant accidents (LOCA) and loss of flow accidents (LOFA) during 500
- kW operation. The previously existing cooling system was retained as the ECCS. As an added backup, the shield tank adjacent to the core is available to provide additional emergency cooling water. A boron injection tank has also been installed. This tank is connected to the suction side of the emergency primary pump. The borated water will not only provide added emergency cooling, but will also provide an added reactivity shutdown margin in the event of an accident. Potential accidents will be discussed in Chapter 15.
A general schematic of the cooling system is given in Figure 4-
- 18. The notation used tu define the coolant temperatures at various points in the cooling system is as follows:
T c
= primary cold leg temperature.
Th " primary hot leg temperature, t e
= secondary cold leg temperature, and i b
m ! l 74
O Tg i i
]
'~ ~~
CCOLING CORE TANKS Te l T g c b " I I DUMP TAT. 4 O , PRI .' ( 4 HEAT I EXCH. h c G " SEC. PUMP PRI. PUMP Figure 4-18 General Schematic of the Cooling System O 75 1
9 1:
- 0 t = secondary hot leg temperature.
h 4.4.5 Thermal Hydraulic Limiting Conditions 4 From a thermal hydraulic standpoint, there are three limiting conditions that were imposed for safe operation of the reactor. They include: (1) bulk or saturated boiling of the primary coolant must not occur, 4 (2) subcooled or surface boiling of the primary coolant met not occur, a,nd
. (3) fuel melting mst not occur in the case of a LOCA or LOFA.
The steady-state thermal hydraulic analysis examines whether condi-tions (1) and (2) can be raet while the transient thermal hydraulic analysis examines whether condition (3) can be met with the reactor ! operating at' a maxima rated thermal power level of 500 kWt. 4.4.6 Steady-state Thermal Hydraulic Analysis 4.4.6.1 Core Heat Balance } All the heat generated in the core mst be removed by the primary coolant; hence, a heat balance may be described by P = MppC h(Tc-T )' 4'1 where P = core thermal power (watts), Mp = total primary coolant mass flow rate (kg/sec),
- Cp = primary coolant specific heat (J/kg-K),
Th = primary hot leg temperature (OC), and i
" primary cold leg temperature (OC).
T c Rearranging equation (1).gives, for the primary hot leg temperature, T 3 = T , + y'C . 4.2 PP l 0 76
O 4.4.6.2 1ntermediate seat exchan er seae na1ance All the heat generated in the primary loop must be transferred, via the intermediate heat exchanger (IHX), to the secondary loop. In per-forming the IHX heat balance, the primary coolant pump heat and insula-tion losses in the piping were ignored . A heat balance across the secondary side of the IHX gives th"Ec *MC ap where i . M, = total secondary coolant mass flow rate (kg/sec),
- . t h
= secondary hot leg temperature (OC), and ,
t e
= secondary cold leg temperature (OC).
To relate the primary and secondary temperatures, the heat transfer across the IHX can be expressed by: P = AU E , 4.4 where i A = effective heat transfer area of the IHX(a ), U = overall heat transfer coefficient for IHX(J/sec-m 2), 2 and AT = average ef fective temperature difference for the entire IHX ( c) For a parallel plate heat exchanger, E is called the p garithmic - mean temperature difference. (LMTD). Since both U and AT ' depend on both primary and secondary temperatures, they must be determined by iteration. 4.4.6.3 Hot Channel Heat Balance i The coolant channel of most concern is the hot channel. The hot channel heat balance may be written as app C (thhe -T )=f, 4.5 , I where up = . average' mass flow rate per channel (kg/sec), 77 L s
..,-.m,7 .-. v -.,.y4 %v ,
e-- w , ,rw--.~ --y y - .
.w.eer ,- ..,eur e.w2
t Thhc " outlet temPetature of the hot channel (cC), F = hot channel factor, and N = number of coolant channels in the core. 4.4.6.4 Bulk Coolant Temperature Profile in Hot Channel The bulk coolant temperature profile in the hot channel is given by: Tbhc(*} " c +M PP 0 9hc(*'} *' '
- where Tbhc(z) = bulk coolant temperature, at axial position z, in the hot channel (OC),
w = width of the coolant channel (m), and l q"hc(z) = axial heat flux distribution in hot channel (J/m 2-sec). The hot for the reactor has been channel axial heag*f ux determined experimentally distribution to be: i 2 q"hc(z)=0.0946P(0.515+0.485 sin (f)]kw/m, 4,7 i 0 where : 1 ]. P = core thermal power (kW) and z/L = fractional s.xial position along hot channel (z=0 is at chan-
- nel inlet).
4.4.6.5 Fuel Clad Surface Temperature Profile u The fuel clad surface (wall) temperature profile in the hot channel l is given by l i q" (z) Twhc(*) " bhc(*) + h ' 'O 4 where i Twhc(z) = fuel clad surface (vall) temperature, at axial position s, in the hot channel (OC)-and h = hot channel heat transfer coefficent (kW/m2 -K). 4 78
= _ _ . - . . .
\
l
, 4.4.6.6 Results of Steady-State Thermal Hydraulic Analysis
. WL.,a the reactor -power level and primary coolant flow rate deter-eine the tesperature rise in the core, it is the secondary cold leg temperature, tc, which is the main factor influencing the pdaary cold leg temperature, T,. This, in turn, affects all other temperatures in the core. The secondary cold leg temperature is determined by ambient j environmental conditions. All calculations of temperature were per-4 formed for several secondary cold leg temperatures and power levels . o as to cover the entire anticipated range of operating conditions. Tho anticipated range of secondary cold leg temperatures is 16-380C (60-1000F) and the anticipated power range is 0-500 kWt. The primary cold leg temperature will range from 2400 (750F) to 440C (1120F) at 500 kWt with a ma::imes anticipated secondary cold leg temperature of 380C (1000F). The corresponding primary hot leg temper-atures will range.from 240C (750F) to 660C (1500F).
- The primary concern during steady-state operation is ensuring that
- no boiling occurs in the core. This can be guaranteed only if the fuel clad hot spot temperature can be kept below the saturation temperature of the primary coolant. The reactor facility is located at an altitude i of 610 n,(2,000 ft.). The corresponding coolant saturation temperature is 980C (2080F). For conservatism, the operational restriction that the i fuel clad hot spot temperature must never exceed 9300 (2000F) was imposed. From Figure 4-19, it can be seen that the reactor can be i
O operated at a power level of 500 kWt for secondary cold leg temperatures below 290C (850F) while meeting the above restriction. For higher secondary cold leg temperatures, the reactor must be operated at a power
! level less than 500 kwt.
Ii Under these limiting conditions, the corresponding primary hot leg
; temperature is 570C (1350F). During 500 kW operation, a high temper-i ature scram set point will be set at 570C (1350F) while the high temper-i ature annunciator is set at 540C (1300F). Hence, for continuous opera-tion at 500 kWe, the primary hot leg temperature mast not exceed 540C l (1300F).
j Based on these results, a steady-state " operating regime" has been
- determined for 500 kWth operations, i.e., the maximum power level at j which the reactor may be operated, for given secondary cold leg temper-atures, so as to ensure that no boiling will occer anywhere in the core. This is shown in Figure 4-20. The solid curve represents the power level at which the primary hot leg temperatures will be 540C (1300F). This temperatura must never be exceeded. As the secondary I cold leg temperature increases, the power aust be decremented when this j temperature reaches 16.60C (800F). The solid line in Figure 4-20 is the actual operating curve; however, to be able to operate along this curve,
; an automatic control system would be required to control the power level
! as a function of the secondary cold leg temperature. Since an automatic control system is not yet implemented, it is necessary for the operator. I to manually decrement power. From an operational viewpoint, the operat-i ing curve is the dashed curve in Figure A-20. I 79
5 t 71-ure 4-19 Fuel Surface Temperature At Several Fower Levels n o 300 -
- O -
n A e 250 - - o L - - 3 200 - - - - - - - - - - - - - - - - - - - - ------------ e i g _ c. 7 E i i e 150 - t-- - e _ 0 - e to 100 - ca - E j 50 - x e E O ' ' ' ' ] 60 70 Ah d0 IDO Secondary Cold Leg Temperature, t e M i-O Figure 4-20 Operating Limits For 2eactor Soo ........_..E i l n I, l - 3 450 - '------ 4 1 i - I e t
> I e
l i 4 400 - '-------h I 4 e ' 2 I o I A I 350 - --------- ! c H : .l- I
.o. I n t 4 w" 300 - '-------
a 8 i % l c '
- A a 250 - '---
o - a
.e4
' C x y ~ ohl , . . . . o 60 70 80 90 100 Secondary Cold Leg Temperature, te ( F) 80
. - - - - -. .. . .- - _ . - - . . - = .. --- . . - -_- - - . -
3 Z-t 4.4.7 Transient Thermal Hydraulic Analysis i 4.4.7.1 Types of Accidents Studied i [ Several classes of LOCA's and LOFA's were studied. The causes of
- these accidents can be broadly categorized as purely accidental (e.g.,
equipment failure) or due to operator error. l
- f. 4.4.7.2 Loss of Flow Accidents l It is desirable to maintain full primary and secondary flow at all times. There are, however, several credible situations in which primary and/or secondary flow any be lost. Tha flow may be lost by pump motor burnout, pump shaft shear, loss of building power, operator error 4 (accidentally tripping pump (s)), etc.
! If primary flow is lost, the coolant will becanc essentially stag-
- nant and virtually no heat will be transferred acroas to the secondary I- system. To maximize the potential severity of the accident, it was assumed that no heat is transferred across to the secondary side when performing the analysis. The worst-case loss of primary flow accident that can occur is one where all the control rods remain stuck out in i their critical positions with the reactor at full power when attempting i to scram after the accident. Present operating procedure requires that i the primary coolant be dumped from the core as part of the scras. - This
] provides a reactivity shutdown margin of about -30% ak/k. However, this
- procedure should not be routinely used with the reactor operating at 500
- kWt instead of the present power level of 100 kWt. The decay heat is j sufficient to cause fuel melting if there is no air flow through the i core afcar the coolant is dumped. Calculations have shown that l in ~2-4 hours, fuel melting could occur if no heat were lost f rom the j fuel. This is supported by similar results in the Safety Evaluation Report for the UCLA research reactor.
l With the control rods stuck out and a loss of primary coolant flow, a new procedure for shutting down the reactor was adopted. As soon as - l the loss of flow is detected, the system will ' automatically attempt to i j scras. When the operator notices, by means of the rod position indi-i cators, that the rods have remained stuck out, the first course of ! action is to check indicated power level and rod top / bottom indicators. 3 The operator would then dump the coolant from the core to provide a safe' j shutdown margin. An attempt must be made to manually insert the rods. ' l If this is successful, the dump valve should be closed and emergency ! = cooling initiated. Should the rods ressin stuck out, boron injection + should be initiated, followed by closing the dump valve and initiating i emergency core cooling. Refer tw "f.cre 1-lC fu an overviw of the l cooling system. t [ '4.4.7.3 Loss of Coolant Accidents t
- These say occur as 'a result of a rupturn of the primary piping, I rupture of the dump tank, any other form of ' leakage or operator error 81
- - - , - . _ . _ - - _ . . . , - - - - _ - - . . - - . - . , - , , ~ . _ _ . - . _ _ . ~ . . . . . - , - _ . . . - . - - . _ . . - - _ .
b . , , \
\
f - (accidental opening of the dump valve). Primary flow will fall below the minimus permissible flow rate and the reactor will scras. As mentioned above, if sufficient air flow is not established through the j core, fuel melting may occur. To prevent this from occurring, the i immediate operator action is to verify that all control rods are fully } inserted and to initiate emergency core cooling within 110 minutes. ; 4.4.8 Conclusions i j It has been shown that it is feasible to upgrade the VPI&SU reactor
- facility from the present maximum power level of 100 kWt to 500 kWe from j
- a thermal hydraulic standpoint. However, at 500 kWe, operation of the j reactor will be more restrictive than is presently.the case.
4 The steady-stace theresi hydraulic analysis has shown that, depend-ing on external environmental conditions, operation at 500.kWt will not always be possible. The limiting condition for steady-state operation ]~
- is the prevention of coolant boiling anywhere in the core. This will
- necessitate operating below 500 kWe if the secondary cold leg temper-j ature (environmentally determined) exceeds 26.6*C (804).
! The transient analysis has shown that, under certain assumed acci-l dent conditions, fuel melting may occur at 500 kW, a condition of i primary concern in the event of IDFA or LOCA. This has necessitated the
- need for installing an emergency core cooling system and a boron i p injection system. Also, prior to 500 kW operation a means of manually j v opening and closing the dump valve will be provided.
i j 4.5 Materials } All original materials and equipment in the reactor core area and j shielding were provided and installed by American Standard or its sub-j contractors. Though equipment such as detectors, wiring, and shielding j asterials have been replaced. or added during prior license periods and during the recent systems upgrade, the materials used were of equal or } higher quality. As an example, the original 2 in' diameter core tank ' 1 overflow pipe was replaced with a 4 in diameter pipe, but the material ! used in the replacement was 1100 aluminum in place of the original 6061- , , T6 tempered aluminum because of the lesser activation problem with the l i 1100 grade. - Stailarly, the control rod drive shaf ts were replaced with l 1100 aluminum, ' though present plans are to return these shaf ts to the } 6061-T6 original specification because of the improvement in' mechanical l ) properties over the present 1100 grade shafts. l l Appendix 4-1 gives a list of asterials and applications. Details in the form of complete materials lists and drawings are maintained on file at.' the ' f acility. i f. l 4.6 Reactivity control Systems j Reactivity control of the VTAR is provided by four . control blades O (2 < cr 1 1 =1 =. t *i-)- a
=
- ea 1 6 c -
82
,,-..~,--,-me -ve,-- ,.--n, n n,- w.rk- s - n e - c .e -s m ,w g w, . g -- - r-e,.g. .-w,-- ~me,n--- -,m-,---------s-ne-.-,- .-- vm-,,,,--mm,,-r-,w,"
7 .- p11shed -by removing the coolant / moderator from the core by the opening of the primary coolant system dump valve. The dump valve is fail-safe in that it must be held in . position by an electric motor and clutch against spring tension. Interruption of clutch power allows the spring to open the valve, which then drains vatar from the core and primary piping directly to the dump tank. For additional detail and description, see Chapter 5. The control blades are of the " window shade" type. The functional section of each is a boron-aluminum material, 7 in x 7 in x 1/8 in in safety and shim blades and 2 in x 2 in x 1/8 in in the regulating blade, ! operating - in an aluminum shroud or raceway. The most recent blade calibration curve is shown in Fig. 4-21. Blade withdrawai is limited to a minimum time of 160 seconds per blade, with safety blade withdrawai interlocked to operate in series. Therefore, for example, safety blade #2 cannot be energized before safety blade #1 is completely withdrawn. Insertion trip time (blade drop) is measured to be less than 0.8 seconds. Additional contro'L blade detail is provided in Chapter 7. O + t 4 1 4 e f O 83
- 4. , .
. . . x .., - . , , , -......,-..,-y,, , - , , _ . .- . . , , _ - , - - , . , - -
' t O O O
.13 -
1
. 12 -
I i
.30 - '
i 9 09 - i l .. ou - m safety != peo X 5.153 e .01 - Safety 2= peg X 4.069 ., - 3 shin 3= nug X 5.046
.06 -
~
Total Worthos (uK/El i .05 - Regulatine .131 2/21/83 , $ safety 1 . 675 2/21/43 i _ g Safety 2 .533 2/21/83
- - d .04 -
Shim .661 3/16/83 i
.03
.02
.01 -
l , O n . e . . e a e e e a e a e a e 2 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 i Inchus Hod Withdrawal , Figure 4-21 Regulating Rod Reactivity Measurement 2/21/83 g s
. 4
1 () Appendix 4-I Materials Incorporated in VTAR Component Materials Used A. Beam Port and Plugs Carbon Steel Mild Steel Ordinary Concrete B. Concrete Shield Aluminum Alloy G2 42-A Aluminum 6061-T6 Concrete-class C 4211 Rubber Coated Packing Grout-Embeco Premixed Mild Steel Wood 1 C. Core Support Plate Aluminum 6061-T6 Reactor Grade Graphite i i D. Core Support Plate Grout-Embeco Premixed i Hardened Steel { Mild Steel E. Core Tank Aluminum Alloy G-5 11A-T6 Mild Steel F. Gamma Curtain Aluminum 1100-F Aluminum 6061-T6 Lead j Mild Steel G. Graphite Shield Tank Duct Aluminum 6061-T6 Grout-Embeco Premixed Mild Steel Nuclear Grade Graphite i H. Graphite Stringer Aluminum 6061 ! Aluminum 6061-T6 Reactor Grade Graphite I. Thermal Column Door Boral ' Grease High Density Concrete t Lead Mild Steel J. Thermal Column Support Plate Aluminum 6061-T6 Grout-Embeco Premixed-(). Mild Steel i l f 85 ,
, . l l
- 5. COOLING SYSTEM 4
5.1 General Description l This chapter describes the primary cooling system and its various components. The reactor cooling system serves the dual functions of removing heat generated by fission and of moderating the fast fission neutrons in the Virginia Tech Argonaut Reactor (VTAR). 4 The cooling system consists of a primary system and a secondary system. Also included is an Emergency Core Cooling System (ECCS) and a Boron Injection Tank. Those auxiliary systems are described in Chapter i 6. The primary system transfers the heat from the reactor to the heat exchanger; this heat is then removed by the secondary system to the i cooling tower with no mixing of water between the two systems. A functional diagram of the cooling system is shown in Figure 5-1. 5.2 Primary Coolant System i 5.2.1 System overview 4 i l The primary coolant system of the VTAR is shown schematically in l Figure 5-2. The primary coolant is stored in the dump cank which has a } capacity of 220 gallons of water. Make-up water is supplied to the i primary system by desineralizing city water and using a connection to the dump tank. The primary circulating pump takes its suction from the dump- tank through 'a 2 in gate valve. The pump discharge is directed
; toward a plate-type heat exchanger and the primary conductivity cell.
The outlet of the heat exchanger branches to either the 6 in inlet / dump line through the primary flow monitor and inlet temperature monitors or via a shut-off valve and local flow meter to the ion exchanger. The outlet of the ion exchanger (normally adjusted to 1.5 gpa) returns to
,i the dump tank through a mechanical f11ter and an isolation valve. The primary flow monitor will transmit a reactor scraa signal to the control console if the primary flow rate drops below 25 gpa. This signal is
- part of the reactor safety system, preventing operation or scramming when the primary flow is insufficient for heat removal. ] The water pumped into the 6 in inlet / dump liu will either return to the dump tank (if the 6'in dump valve is open) or will flow into the ; ! two parallel-connected core tanks. Water near the top of the core tanks then flows into the 4 in coolant return line and returns ' to the dump i i tank passing the outlet temperature monitors. The information from the outlet temperature monitors is supplied to the reactor safety systes l with an annunciator set point at 165'F. ' This safety measure alerts the Reactor Operator to conditions such as restriction or reduction of , 4 primary coolant ' flow, reduction or restriction of secondary coolant , flow, a malfunction of the heat exchanger or excessive reactor power. A i portion of the returning water is directed to the N-16 delay tank which i is connected to the primary coolant radiation monitor. This monitor is t . - designed to detect the radiation level of any fission products which t. 86 1 l _ _ _ _ _ _ _ _ . _ _ _ _ _ _ __. _. ._ . ~ _ _ _ ._.D
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i mightl be released to the primary coolant. If the readings f rom the monitor are greater than 5 times normal, an annunciator signal is trans-mitted to the control panel. This provides an indication of possible ] fuel element - f ailure. The outlet from the delay / counting chamber
- returns to the dump tank.
1 A sight glass located adjacent to the process pit allows a visual ! [ check of the~ reactor core tank level. A float switch in the sight gless ! is wired to the reactor scram. system actuating a reactor scram if the 4 water level is within one inch of the top of the core tanks. 5.2.2 Primary Coolant Pump The primary coolant pump is a Crane Series G Chempump designed to circulate 40 gpa of water with a maximum capacity of 110 gpa. The design specifications for the primary pump are listed in Table SA-1, Appendix 5-I. 5.2.3 Flow Monitor ! The flow monitor system is composed of a paddlewheel flow sensor j and a digital averaging display flowmeter. The sensor is capable of j detecting flow velocities from 1 to 50 ft/sec. A scram signal is trans-mitted to the control console if the primary flow drops below 25 gym. The design specifications of the flow monitor are summarized in Table SA-2, Appendix 5-1. I j 5.2.4 Heat Exchanger The heat exchanger is a type 304 stainless steel water-to water i plate-type unit designed to have 90 gpa of reactor coolant water circu- l laced through the primary side and 685 gpa of cooling tower water circu- , laced through the secondary side for removal . of 500 kWe heat load. There are 2 passes (5 channels per pass) on the primary side and one pass (11 channels per pass) on the. secondary side. The flow patterns in i the heat exchanger are illustrated in Figure 5-3. t . The design specifications for the heat exchanger are summarized in 4 , Table SA-3, Appendix 5-I. ' . 5.2.5 Inlet and outlet Temperature Monitors i Both the inlet and outlet temperature monitors are resistanes temperature detectors (RTD), augmented by local dial thermometers, used to monitor the mean bulk temperature of the coolant. . An annunciator signal w111- be transmitted to the control panel if the outlet temper-L store exceeds 165 F.
-The design specifications for the temperature monitors are O ri 4 i= ' 61 5^-' ^aa di 5-t-90
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5.2.6' Primary Coolant Radiation Monitor The radiation monitor is a Victoreen Model 855 series monitor system consisting oft
- 1. An 857-20 detector module containing a gamma-ray sensor (GM tube) and associated counting circuitry.
- 2. An 856-20 read-out module which gives the level of radiation on a panel meter with output connecting points for input to a computer and/or recorder; it also includes an alarm circuit for warning of excessive radiation and a fail-safe circuit which continually monitors the operation of the system.
The design specifications for the radiation monitor are listed in Table 5A-5, Appendix 5-I. 5.2.7 shield Tank / Core Tank hvel Monitors All the level monitors are pressure-actuated diaphrag:n micro-switches. The shield tank low level monitor transaits an annunciator signal to the control console if the water level is lower than 2 f t. below the top of the tank. The design specifications for the level monitors are suonarized in q Table SA-6, Appendix 5-I. D 5.2.8 Dump Valve The dump valve is a 6-inch aluminum-bodied angle valve with stain-less steel seat and disk. The valve is closed during normal operation of the reactor. In the event of scram or reactor shutdown, the valve is opened automatically and the moderator drains from the core tank. The design specifications for the dump valve are listed in Table 5A-7, Appendix 5-I. 5.2.9 Domineralizer_ Two nuclear grade ion exchange resin beds (capacity 1400 grains expressed as CACO 3 ) are combined to form a 19" sonobed domineraliser which removes all anionic and cationic constituents. The quality of water produced is at the level of triple-distilled water, i.e., 15-18 segohns-cm resistivity. The design specifications for the dominera11:er are summarized in Table 5A-8, Appendix 5-1. 92
f f 5.3 Secondary Coolant Systes , 5.3.1 Systes Overview h secondary coolant system of the VTAR is shown schematically in Figure 5-2. h secondary cooling system is a closed-loop systee util-ising water as coolant, a plate-type heat eschanger previously described and a cooling tower located on the roof of the reactor building (Robeson Hall). N nominal secondary flow rate is 200 sps with a maximus capac-ity of 750 spe. The flow rate through the ' secondary coolant system is controlled by a motor operated throttle valve located on the output side 1 of the pump to allow for operation of the reactor at lower power levels , where the rates of heat removal are proportionally smaller. l A floweeter located in the pump outlet piping transmits the flow i indication and an alarm signal to the control console if the secondary ! flow rate drops below 125 gym. Water is made up to the secondary system by utilising city water to compensate for evaporation loss. In order to minimise rapid temperature changen in the primary system and to prevent ! the secondary coolant from freesing, the temperature in the secondary t system will be controlled to stay above 55 F. For this purpose a 10 kW ; heater is installed in the sump of the cooling tower. i l 5.3.2 Secondary Coolant Pues . The secondary pump in an Allis-Chalmers , Model F-4 C80 pump d i= 4 t =- **5 => e ti= i=* < -
-O i capacity of 750 spe. A motor-driven throttle valve is located on the output side of the pump to control the coolant flow rate.
The design specifications for the secondary pump are summarised in Table 5A-9. 5.3.3 Throttle Valve The throttle valve is a 6-inch motor-driven butterfly valve used to I control the secondary coolant flow rate. h design specifications for ! the throttle valve are listed in Table SA-10. ! i 5.3.4 Pressure Indicator The pressure indicator is a pressure gauge manufactured by the ; < Amstek U. S. Division. The full scale pressure'of the gauge is 100 pet. t
- 5. 3. 5 Cooliu Tower h most significant change in the reactor coolant system is the l l
. installation of the cooling tower. With the installation of the cooling l l tower. .the ~ secondary coolant system becomes a closed loop system. h !
l cooling tower is a mechanical draft type capable of dissipating up to 1 i NW of heat to the atmosphere. l l rO i l 93
-.----x_-- --_-____-__-____-__-_N_-___________________________,__________-___._________________________
( l i q V 5A-11. The design specifications for the cooling tower are listed in Table
- 5. 3. 6 Temperature Monitore The temperature monitors located at the inlet and outlet of the heat exchanger monitor the mean bulk tempetsture of the secondary cool-t ant. The design specifications for the secondary temperatura monitors '
l are the same as those for the primary temperature monitors. l l 5.3.7 Floweeter The flowmeter is a Mapco Model 9000 Nusonics Flowseter consisting of two transducer assemblies, a transmitter assembly, and a 6-inch flanged flowtube. The flowmeter utilizes the measurement of the speed i of sound through liquide for determining fluid velocity. An alarm ( signal is sent to the control console if the secondary flow rate drops below 125 gps. The design specifications for the flowmeter are Itsted in Table SA-12. 5.4 Functional Block Diagrama
- Additional drawings of piping layouts, cooling tower arrangement, block diagrams of inatruments, etc. are shown in figures S-4 through 5-7.
O V 5.5 Cold Water Acetdent Operational Congideratione l The need to address a Cold Water Accident (CWA) during all phases . of reactor operation has been thoroughly investigated and measures to l prevent this are discussed here. 1 i There are four levels or barriers to prevent an occurrence. Titice
- a year the secondary flow rates will be adjusted for optimum performance l with the reactor shutdown. Operational history will determine the tiow rate settings.
l The two manual butterfly valves located on top of the cooling tower l will be used to adjust the flow rate. These valves will then be secured l in their throttle position settings by the locking levers on each valve. This is the first level of protection. Providing a second harrier of protection is a locked key controlled hatch leading to the roof. A deliberate throttle valve setting change would require an individual to i scale ,the ws11e or break through a locked equipment room and locked roof hatch to gain access. Additional protection is afforded in the reactor room by lockwiring the secondary throttle valve handwheel (which can manually override the motor controller) and posting a red sign on the valve cover. A pues on/ valve shut alarm and interlock atfords the fourth level hs of protectton. This prevents starting the Secondary Cooling pump with 94
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the throttle valve open and sounds an alarm if an attempt is made to do so. This will also aid the operator. When first starting the pump the alarm is also present which serves as a reminder that cecondary cooling must be initiated slowly. In the event all these measures fail and a CWA occurs, a fuel element failure still would not occur (Chapter 15). J O 2 d 1 99 f
l
- o APPENDIX 5-I Cooling System Specifications Table SA-1 Primary Coolant Pung Design Specifications Manufacturer = Crane Co., Chempump Division
- Pump Model = GC-3K-152H-IS Capacity = 120 gpm with 60 ft head Power = 4.4 kw, 220 VAC, 3 phase Pumping Temp. = 50-400 F Operation / Control = pump on-off from the reactor console, and the local controller.. "0R" function for pung start, "AND" function for pump off.
Table 5A-2 O eri rv '1 " iter o t= sa ciric ete - Flow Sensor Manufacturer = Signetics, Inc. i Type = paddle wheel Model No. = C-5618-10 Flow Range = 1 ft/sec minimum, 50 ft/see uaximum i Flow Meter ) Manufacturer = Signetics , Inc. Type = digital averaging display (D.A.D.) Model No. = C-5622-35-VH Accuracy =
- 1% full scale O
100
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f~ . Table SA-3 Heat Exchanger Design Specifications Manufacturer - Alfa-Laval, Inc. Model No. = AM10HBM Primary Side Secondary Side
~
Flow Rate, gpa = 90 685 T-in, OF = 135 80 T-out , OF = 97 85 Effective Heat Transfer Area = 92.57 ft 2 Plate Thickness = 0.002 ft
- Distance Between
~ plates = 0.016 ft plate width = 1.92 ft i
j O operating pressure = 100 psi i
- operating temperature = 2400F l plate material = AISI 304
- No. of plates =
22 (20 effective) Heat Load = = 1,704,765.9 btu /hr
- 500 kW values Table 5A-4 Design Specifications for the Temperature Monitors Manufacturer = Omega Engineering, Inc.
j Type = RTD Model No. . = 50
-Accuracy =
- 1.0% full scale
{} Range = 0-2000F F 101
i + 0 Table SA-5 Design Specifications for the Radiation Monitor Detector Manufacturer = Victoreen, Inc. Model No. = 857-20 Radiation detected = Gamma rays Range'of radiation = 0.1 to 10 4mR/hr Readout Manufacturer = Victoreen, Inc. Model No. = 856-20 Accuracy =
- 20% of reading Range = 0.1 to 104mR/hr Table SA-6 Level Monitor Design Specifications Manufacturer = Barksdale valves Model = 420 Type = 347 Stainless Steel Diaphragm i
Proof Pressure = 10 psi Approximate Actuation Pressure = 0.15 * .07 psi i k l O 102
l I Table 5A-7 Design Specifications for the Dump Valve t Manufacturer = American Standard Design pressure = 20 psig Design temperature = 2000F Design pressure Across disk = $ psi Time to open = <0.5 see with 5 pai across the disk Maximum leakage rate = 0.5 gph with 5 psi across the disk Table SA-8 Design Specifications for the Demineralizer Manufacturer = Cole Palmer Instrument Co. I Model No. = K-1503-00 Height = 23-1/4" Diameter = 7" Maximum flow rate = 2 gpn Table 5A-9 fecondary Coolant Pump Design Specifications Manufacturer = Carotek Inc. Allis-Chalmers Industrial Pump Div. Model No. = F-4 Type = CSO, centrifugal Capacity = 750 gpm with 100 f t head Power = 30 kw, 460 VAC, 3-4 Operation / control = Pump on-o'f from the control console and the j local controller. "0R" ~ function for both pu pm ' start and pump off. 103
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Table SA-10 Design Specifications for the Throttle Valve Manufacturer = Keystone International, Inc., Keystone Valve Div.
- Type = Butterfly Size = 6 in Body = Carbon steel Dise and Stem = AISI 316 Stainless Steel Operation / control = Positional control from control console, 0-100%, local manual control with override capa-
, bility Table SA-ll i Cooling Tower Design Specifications Manufacturer = Marley Co. O Model No. = 8805 A Flow Rate = 375,000 lbm/hr I T-in OF = 850F i T out OF = 800F , l 4 Fluid Circulation = treated city water l I Specific heat = 1,0 Btu /lba OF I Plenum reserve = 1600 gallons (minimum) Heat exchanged = 1,875,000 Beu/hr
. Operating pressure = Atmospheric i
i (:) 104
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- Table SA-12 Design Specifications for the Secondary Flowmeter
, i Manufacturer = Mapco Controls Co. ! Model = 9000 Nusonics i Serial No. = 34095-2xxB Range = 100 gpm - 1000 gpm Accuracy =
- 2.0% full scale 0
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- 6. ENGINEERED SAFETY SYSTEMS i 6.1 Emergency Core Cooling System (ECCS)
In the event of a Loss of Flow Accident (LOFA) or a loss of Coolant Accident (LOCA) after extensive operation at 500 kW, a significant amount of decay heat will have to be removed. An ECCS has been installed that is entirely . independent of the primary coolant loop. At the present time operations are limited to a maximum power of 100 kW; however, there are plans to upgrade the power to 500 kW. With this in mind - the ECCS was installed to facilitate operator training on the i system and to establish a safe and reliable system prior to the actual power upgrade. 6.2 System overview i . The Emergency Core Cooling System schematic is shown in Fig. 6-1. The ECCS has a. nominal flow rate of 20 gpm. The emergency cooling pump takes its suction from the dump tank through a 1 in. gate valve. In the event of a pipe rupture or break in the primary cooling system, it alternately can take its suction from the shield tank. The shield tank will serve as a > 2200 gallon reserve of cooling water should a coolant pipe break occur and not allow water to flow back to the dump tank. The O gem, ai charse i- airectea tewara 1oo ** heti ana tehe heat exchanger, a flow transducer, a gate valve, a check valve and then into the core. , 6.3 System Operation Initiation of the ECCS is inherently simple due to the nature of its use. The gate valves on the suction and discharge serve only as isolation . valves and are open during reactor operation. To initi ate , emergency cooling, the actions necessary are to energiz'e the ECCS rump ' and actuate the switch initiating -city water cooling to the f. eat exchanger. The check valve on the primary coolant loop core inlet 1Line will shut due to back-flow from the emergency coolant pump and flow will be routed into the core. The reverse . is true during normal operation. i Flow from the primary loop will keep a check valve in the emergency cooling loop . core inlec line closed 'and hence there will be no back-flow into the emergency lines. 6.3.1. Emergency Coolant Pump The emergency cooling system uses a Chempump series S sealess type pump designed . to circulate 20 gpa with a maximum capacity of . 25 gps. The-design specifications for.the emergency pump are listed in Table 6A-1, Appendix 6-I.
- O 106
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-
- y Dump Valve O Dume Tank (220 gal. cap.)
Heat Boron A j , Exch. Injection 100 kW Tank ( 10 gal. cap. ) Shield Tank (2200 gal. cap.) y O .
" PEinary c Coolant City System Water ECCS Pumo Out (20 gpm)'
Figure 6-1 VTAR ECCS System 107
L I 6.3.2 Flow Monitor i The flow monitor sys am is composed of a differential pressure transmitter and an analog display mounted on the control console. Design specifications may be found in Table 6A-2, Appendix 6-I. 6.3.3 ECCS Heat-Exchanger 4 The heat exchanger is a shell and tube type unit manuf actured by the Harrison Radiator Division .of General Motors. - The unit *s designed to have 20 gym of reactor coolant water circulated through the primary side and 150 gpa of municipal water through . the secondary side for removal of a 100 kW thermal heat load. The design specifications of the heat exchanger are summarized in Table 6A-3,' Appendix O-I. I 6.4 Boron Injection System In the results of the transient thermal hydraulic safety analysis for 500 kW, it was indicated that, 'under certain accident conditions, excessive fuel temperatures may occur. t
- The accident of primary concern is the loss of primary flow l' accident (LOPFA) at full power with All Rods Out (ARO). .In the transient thermal hydraulics section, a procedure is suggested for i handling this accident. The procedure requires the water to be dumped i
from the core to provide a safe shut-down (S/D) margin (-307, ak/k). The operator must then force the rods in by whatever means possible, shut j the dump valve, and then initiate emergency cooling to prevent excesaive
; fuel temperatures from occurring.
t Under the worst possible conditions, if all or some of the rods cannot be inserted, even by force, some other method must be used to provide an adequate shutdown margin to prevent criticality from-l occurring when emergency cooling is initiated. The simplest method is j to inject a concentrated solution of boric acid (H3B0 3) 1"C* th' j coolant. 6.4.1 -System overview I The barua lajection system i.vanists of a Borou Injection Tank (BIT) I with ' a 10 gallon capacity mounted above- the process pit. The tank is l valved into the suction line of the emergency core cooling system. The i tank contains a minimum inventory of water and is kept at - ambient temperature with no Boron maintained in the tank.
-6.4.2 System operation 1 As indicated in the transient thermal hydraulics section, flow must
.be re-established within two -hours, under - worst possible conditions,
- before excessive fuel temperatures occur If the' operator cannot insert
.the rods even by force, then a preseasured amount of boron is dissolved l
-h- in the Boron Injection Tank, the tank outlet valve is opened, and the 108 i
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1 4 f emergency core cooling system is started. A calculation is included in 1 Appendix 6-II for the necessary boric acid inventory. This inventory is l 9 to be maintained in the vicinity of the Boron Injection Tank during operations above 100 kW. + t 4 t 1 f Y O i t [O 109 { , e n . s ,. , . . , , - - a - - + - , e ,.w,e. ., ---,--e-a,-- ,-,.--,~m-, , e . , , --v e ev, ,, ,e,-.e--a -
l i- () Appendix 6-I Tables of Specifications Table 6A-1 ECCS Pump Design Specifications Manufacturer = Chempump Corporation ? Pump Model = SF 3/4-3/4S Type = canned rotor, canned stator, sealess i Capacity = 25 gpa 8 40 ft TDH Pumping Temperature = 50-2000F Operational / Control = Pump on/off from console f Table 6A-2 5 0v ECCS Flow Monitor Design Specifications
- Flow Sensor / Transmitter Manufacturer = SWART 00T CO.
Type = Orifice / Differential Pressure-Transmitter Model No. = D2T Range = 0-742 PSI Construction Materiais = 316 Stainless Steel Accuracy =
- 1/2% of Span i
l i ()- l 110 l' . - _ . _ _ _ _ _ _ . _ _ _ . . .__ .-. - , --. .
O raste 6i-3 . ECCS Heat Exchanger Design Specification
- Manufacturer = Harrison Radiator Division, General Motors, Inc.
Model No. = 2124-360 Type = Shell and Tube Design Pressure = 165 PSI Materials = Carbon Steel , Flow Rate: Primary Side = 20 gpa 4 Flow Rat.e: Secondary Side = 150 gpa (Nominal)
- Heat Load = 400,000 ETU/Hr I
9 t I 1
- O 111
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t (. APPENDIX 6-11 BORIC ACID REQUIREMENT FOR THE BORON INJECTION SYSTEM The . boron injection , system should be able to provide a S/D capability equivalent to, at minimum, the total worth of all the control rods. This is approximately 2%&/k. For added protection, the boron injection system should be capable of providing a reactivity worth of 2.5% ak/k. The required concentration of boron, in PPM, is given by: p, x 10 where C = 1.92(1-fo) p, = Required reactivity worth in %Ak/k and
- f, = Thermal utilization factor for the reactor in the absence of boron for the 500 kW core, the thermal utilization factor is 0.9. To provide 4 a reactivity worth of 2.5% Ak/k, the boron concentration required is
C = 1.92 1 0. 9) =10.2 m i This is the concentration of boron, uniformly distributed throughout the } primary cooling system, necessary to keep the reactor suberitical when
- emergency cooling is initiated. The molecular weight of H 3B03 is 61.8 j and the atomic weight of natural boron is 10.8. Thus, the boric acid l concentration required is
)
(61.8/10.8) x 130 = 745g of H3 803 per 106 8 of water. j Assuming unit density water, the boric acid. concentration is O.745g/ liter. The total capacity of the dump tank is 220 gallons. The inventory of water which - stays in the primary system piping is 13 gallons. This i leaves an empty volume of 13 gallons in the dump tank when the system is
- filled (not including the core tanks). Therefore, to prevent overflow of the dump tank, the boron injection system inventory should be kept under 13 gallons.. Assuming an inventory of 10 gallons, the boric acid concentration in the boron injection tank mast be:
0.745 x 230/10 = 17.14 g/ liter of water. The solubility of boric acid at room temperature is 63.5 g/ liter, i
- hence there will be no problem in keeping the boric acid -in solution.
The total _ boric acid inventory required is: 112 _ _ _ _ . - . _ _ _ . . . . . _ _ - _ __. ,, , _,_ . ~ . _ _ _ . _ _ _. _ .
w O 17.14 g/ liter x 10 gallons x 3.785 liters / gallon
\
= 649g of H3 B03 in 10 gallons of water.
I l O O 113 l
i l f
- 7. VTAR INSTRUMENTATION AND CONTROLS 7.1 Introduction The VTAR instrumentation and control circuitry monitors and controls virtually all important nuclear and process functions. It also ensures the initiation of a shutdown signal or application of an
- interlock in the event of a potentially unsafe condition.
7.2 Identification of Safety Related Systems The safety related systems for VTAR include the following: (1) Mimic Bus (includas associated control and indicating circuits) (2)- Radiation Monitoring Systems (3) Nuclear Instrumentation (4) Reactor Protection System ] (5) Regulating Rod Control System L (6) Startup/ Interlock Bus O (7) iuu ici cer S7 e a l (8) Rod Position Indication System 7.2.1 Console Mimic Bus All functions required for VTAR normal operation can be controlled
. from.the console with_the exception of the startup source which must be l manually inserted.
5 The primary and secondary mimic bus is shown in ' Figure 7-1. The following indications and controls are available to -the operator from j this panel: l (1) A primary - mimic bus showing the major primary cooling system components, (2) A secondary mimic bus showing the maj or secondary cooling- ! + system components,
.'(3) Primary coolant inlet and outlet temperature indicators, (4) Secondary coolant inlet, outlet, cooling tower pan, and ambient
- (cooling tower) temperature indications, I
(5) Primary coolant' dump ' tank heater control and indication -- 4 manual and auto modes with a " heater on" light, 114
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- O (6) Primary coolant pump start /stop switch with " pump on" indication, (7) Primary Coolant Digital Flometer for primary flow indication, (8) Primary coolant conductivity indication, (9) Fission product " pump running" indication, 4
(10) Secondary coolant two-wire process indicator for digital flow rate indication, (11) Secondary coolant pump start /stop switch with " pump on" indication, (12) Secondary throttle valve motor control switch with " full j open/ full close" and " percent valve open" position indication, ! (13) Cooling tower fan start /stop switch with integral " fan on" indication, i ! (14) Cooling tower pan heater control and indication manual and l auto modes with a " heater on" light. ! (15) Primary coolant. pump overheat indicator (Stator temp. > 425 F). 7.2.2 Radiation Monitoring System j The radiation monitoring cabinet (Figure 7-2) houses the following .I circuits, controls, and indications: I i (1) Airborne Particulate Fission Product Monitor (APFPM) (2) Area' Radiation monitoring system and control circuit. (3) Reactor Ventilation stack and APFPM chart recorders. (4) Two Lambda variable DC power supplies for the radiation monitoring system control circuits. An additional item contained in the cabinet is a closed circuit TV monitor with the capability. of observing the reactor room entrance or the fuel' pits. This feature allows the Reactor ' Operator. to control access to the reactor while operating and serves as a . supplement to Physical Security Plan requirements. I
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7.2.2.1 Airborne Particulate Fission Product Monitor A functional block diagram of the APFPM is shown in Figure 7-3. An air sample is drawn from the reactor ventilation exhaust stack and , passed through filter tape. I l Radioactive particles deposited on the filter tape are detected l using a sodium-iodide scintillation crystal. The signal is amplified, discriminated, sent to- a ratemeter, strip chart recorder, and alarm bistable. Under normal operating conditions two short-lived fission products are detected: Ba-139 and Y-91m. The fission products are produced from U-235 trace contaminants produced from fabrication on the surface of the reactor fuel plates. See figure 7-4 for sample spectrum. In the event of an abnormal release of fission products an alarm will be received and the magnitude of the release recorded on the strip chart recorder. The system is calibrated periodically ut.lizing standard gamma sources. See VTAR Operating Procedures and Technical Specifications for further detail on system requirements. 7.2.2.2 Area-Radiation Monitoring System The permanently installed radiation monitoring system for the VTAR O- consists of a Victoreen 855 6-channel area monitor (with 857-20 series detector) and an Airborne Particulate Fission Product Monitor. Refer to figures 7-5 and 7-6 for an overview of the Victoreen i radiation monitoring system functions, controls, set points, etc. A brief explanation of the theory of operation will be given here. The detectors operate in the Geiger-Mueller region, essentially creating a short pulse output from the " avalanche" effect for each ionizing event. If the detector tube is saturated, additional circuitry is called upon to create an output current which is (approximately) proportional to radiation level. Detector output is sent to a log pump circuit. This circuit performs a logarithmic function. The output current is proportional to the log of the pulse repetition rate (f requency) _ at the input and, therefore proportional to the radiation level at the detector. The DC amplifier provides two functions: substantial, low noise amplification; and distribution of the signal to the various external controls and alarms which are available. The high radiation alarm circuitry also performs two functions. It provides an alarm condition if radiation levels exceed a pre-set value and also monitors itself such that if a circuitry failure occurs (in the alarm circuit), a high radiation alarm will occur. The low alarm (fail-safe) circuit monitors most of the remaining circuits so that in the avent of low signal levels, the green l l 118
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; B/s PWR.
RANGE DWR.) ! ANNUNCIATOR i I Figuro 7-3 Air Particulate Fission Product Monitor 4
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- Alarm Set Point asame Incation (or/hr) (or/hrI Tndication/Coatr01 toest h 11, Sound horno all floors nobeson Itall, Building Evac.
I I" "O AnerJaciators Trip menctor/ Facility ventilation, . heactor Room Disable Rabbit System Controller, Demote Indicator Cround Floor (900rth Entrance) , 3rd Floor, ! physics IJhrary 0.3 15 Scme as building ; Stack ' , (In Vent. Duct) Primary te-16 Delay 100 500 Radiation in Primary coctant Annunciator - N f East mil, Area Beactor Room 1.0 5.0 over tolerance area monitor annunciator (Above Process j Pit)
- *These values may be different from the maximun allowed by technical specifications.
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HCW .- l HMn MT. " K-1/HIGH RAD. LEVEL ' I s K-2/CKT. FAIL I K-3,4/DC CONTROL RELAYS I K-5/ BLDG. EVAC RELAY tigure 7-6 FUNCTIONAL DIAGRAM / RAD. MONITOR CONTROL CIRCUITS ! l i
q ., - ,. ._ __ . - _ . - , _ - _ _ _ _ . _ _ - - - - - - - s b 4
' fail safe panel light will go out. This serves as a warning that the j;
~
circuitry may not be capable of serving its basic function. ' , The Radiation Monitoring System requires little maintenance other ! t . than 6 month calibrations to verify accuracy and operability. functioning is also checked during quarterly building evacuation Proper i f drills. System operability is required at all times. Technical ! j, Specifications address system requirements in greater detail. ! The radiation control circuits for the building alarm are l j controlled by two modules - stack and building. During normal operation ! ! all relays are ' maintained energised. Any control relay failure will -, j result in the building alarm sounding. Horn power availability is ! annitored continously at the console. Two inputs froa each module are j received - one from the high radiation level circuit and one from the t ! circuit failure detector. Each controls a relay which opens a set of j contacts to interrupt power to K-1 and K DC control relays. ! The DC control relayst (1) de-ene'e giae K-5 (building evacuation relay), (2) initiate the building evacuation alare at the console, and (3) de-setivate the rabbit controllers by removing power to the , l contro11e'r solenoid valves. ! l ! i De4nergising K-5 accomplishes several functions: (1) removes j power to all facility ventilation controllers, (2) applies power to the 2 slara horn circuit, and (3) accomplishes " positive lock-out" by opening j - a contact is series with the DC control relays. !
+
i i Ther'efore de-energisation of any single control relay results in I the entire system initiation: The circuit can only be reset by a key * , operated reset . switch on the console which bypasses the K-5 contact (in ! the acknowledge position) and directly energises the DC control relays. ! j To Ieset- the building al' ara control circuit, the following ; i conditione sust,be asts. (1) power available to all circuits. (2) alarm i condition cleared,'and; O) two personnel available - an SRO and a health . physicist.. , i : i For circuit 3 reliability and flexibility 'all control relays are ! !- e supplied continously by the UFS. Additionally, the DC control relays j
- are supplied (via the <UPS) by a , pair of sectioneered +9VDC power -
! supplies. '
j > > I It should be noted thac the building detector (west well of reactor ! i room) also serves as a c'riticality accident monitor-during fuel handling ;
- r f and refueling operations. 6i i j {j 7.2.3 MuclearInstruisentatico [
w The VFAR offeraIs diversity .of nuclear ~ instruasnes for the ! [ 4 monftoring, control, and automatic shutdown of the reactor. i i 4
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10 Two types of source range detectors are utilized - a fisson chamber and a-BF3 proportional counter. Normally only one is used although both may be used. Currently. only one intermediate range using a CIC detector is used. This drawer also performs the important function of a period shutdown. Two power ranges (CIC's) are used with both having automatic high flux shutdown capability. 1 Additionally, two Keichley wide range picoammeters are available. Both are capable of monitoring the entire range of power with either used to supply the Regulating Rod Control circuit. The two instruments share. a high voltage power supply, detector, and outputs to the l Regulating Rod Control' circuit and chart recorder. Selection is accomplished via a switch.- All nuclear instrument histables are continously monitored by a test drawer. This drawer indicates the state of the bistable and if . a failure has occurred. See figure 7-7 for a block diagram of the VTAR Nuclear Instrumentation. 7.2.3.1 Source Range Instruments-Two source cange instruments are available at the console. They O are identical except for the high voltage power supplies and detector types. Each drawer uses a discriminator circuit and logarithmic integrators to amplify neutron induced impulses only and to provide an analog display of 1 to 10 counts per second. The ' rate signal is developed by means of a differentiator-integrator circuit using linear i feedback to a meter display of (-30 to + 3 second) reactor period.- The level circuit supplies s' bistable which serves as an interlock condition for rod withdrawal at values less than 3 counts per second. 7.2.3.2 Source Range Channel No. 2 l A backup alternate source -range channel is 'available which utilizes a BF3 - proportional counter. This channel is not required to be - in i service .for reactor operation. With the exception of the neutron detector, the circuitry for Source Range Channel 2 (SRC-2) is identical j
- to that for SRC-1. The BF3 proportional counter is 26 in.long and.2.1 l in in diameter and is made g .N. Wood Counter Labs (Model No. G 18022).- The counter contains BF3 gas at a pressure of 40 cm Hg and ~
has a nominal operating voltage of 2200V. The BF3 detector is located in the center vertical tube' which f 9 , mounted at the west face of ' the shield tank. The detector: can be moved ! vertically in' the tube and limit switches are provided which give ! -indication lights at the console for the "up" and "down" positions., 1 l During normal reactor start-ups the detector is ~ located in the.
- j. . down position. . When the console count-rate meter, reaches full scale,
- the detector 'is raised to the up position. A test'will be run'with the 124-a ~v - , n. ~ -
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- O reactor operating to determine whether the BF3 detector has any effect I
- . V. '
l [ on the reactivity or the response of the adj acent detectors which service the Intermediate Range CIC and the Keithly Picoammeter UCIC. j 7.2.3.3 Intermediate Range Instrument
. The intermediate range drawer uses a modulated, AC amplified, and demodulated - signy with gogarithmic feedback to provide analog level i indication of 10 to 10 av (i.e., flux). A period signal is also generated in the same fashion except linear feedback is used. The
- i. important period signal . supplies two bistables: a ten second period 4
anuniciator and a five second period scram signal. Rate display is identical to source range. Additionally, there is an annunciator in the
- event of.a high voltage power supply failure.
l 7.2.3.4 Power Range Instruments There are two identical power range' instruments utilizing i' uncompensated ion chambers. Flux levels are high enough such that a steady DC current in microamps can be used to be amplified by a DC amplifier. A 0 to 150% power is indicated via a selector switch. l Both instruments supply bistables, either - of which can initiate a scram signal at 120% of full power (125% by Technical Specifications). 7.2.3.5 Keithley Picoanneters l Two Keithley model picoammeters are available at the console. Both i operate in a similar manner using high gain DC amplifiers with negative feedback. Input is varied by means of inserting dif ferent values of fixed resistors. One model is a vacuum tube type and the other is solid I state. Both instruments are supplied from the same detector, use the same high voltage power supply, and provide an output to the Power recorder and Reg. Rod controller. Selection of the desired instrument is accomplished by a selector switch.
- 7.2.4 Reactor Protective Systems The reactor protective system (RPS) consists of the bistables, scram logic circuitry, power supplies, and associated input signals necessary to manually or automatically shut down the reactor in the event of an unsafe condition.
i. A 90 VDC signal is supplied to four electromechanical clutches:
- (a) two safety rods, (b) one shim rod, and (c) the dump valve. Receipt of a scram signal by 'any of the 7 functions will remove power to the rods' - disengaging the control rod drive mechanism (CRDM) from the rod
'. and allowing them to drop in and open the dump valve, draining the core tanks. The shutdown margin provided by this method is greater than 30% AK/K.
O 126 _ _. _ - , - - - _ _ - _ _ - _ _ . ~ . _ _ - - _ .
F
~
Rod drop times are less than 0.8 seconds; the dump valve opens in .l less than a second with resultant emptying of the core tanks taking only I l slightly longer. ! The protective circuitry consists of the bistables and their signal inputs, mixer-drivers, solid state relays, auxillary ralays and back-up scrams (Refer to. Fig. 7-8).
- The mixer-drivers are negative-OR gate discriminators which supply
- control voltage to the solid-state relays and aux 111ary relays. The i
negative OR logic is arranged so that any one of the seven scram signals { at or below 0 volts will cause a zero volt scram signal output. Output 4 is positive ("no scram") only when all inputs are positive. The two mixer-drivers are designed to operate in parallel in a "1 out of 2 for scram" manner. This ensures that no semiconductor failure in a. mixer-
- driver can lead to a "cannot scram" condition.
The solid-state relays are commercially available solid state . relays with maximum ratings of 240 VAC at 10 amps. They operate in the present system switching 110 VAC at less than 1 amp. A positive input from the mixar-driver acts as a trigger signal for the SSR, producing an . SSR output. Absence of a positive mixer-driver signal will cutoff the
; solid state relays, causing the output to the clutches to fall to zero
! volts. Indicator lights on the console are provided to show SSR operation. There are five levels of redundancy built into this system. The i first level is the SSR's. Should either of these receive a zero input, experience electronic failure, or be removed from the system for any reason, the ef fective ouptut drops to 0 volts, producing a scram. The
- auxillary relays are in series with the SSR's. Again, any scram signal
~ to the aux 111ary relays, relay failure, or relay removal causes a scram condition. In addition to the above four (2 SSR's and .2 aux relays), a back up scram provides a fif th level of redundancy. Again, electronic failure, any scram signal, or relay removal will cause a scram condition. This ensures that if both SSR's and both aux 111ary relays fail, the reactor will still scram. In addition to the above, another. important part of the reactor , protective circuitry is the console power key switch. When this switch . is off, power cannot be applied to the rod and dump valve morors and ! clutches. If this switch is turned off during reactor operation, a scena will result. l Reactor Safety System Setpoints are found in Table 7-1. i 7.3 Non-Nuclear Instrumentation and Control Circuits 4 Several -circuits are available for reactor controls and indication. Among these are the Control Rod Drive Mechanisms (CRDM's)
- and associated control circuitry,. Rod Position Indication ~ (RPI) circuitry, Startup/ Interlock Bus, and Power Controller.
s 127 i
---,-,--a
~_ . , _ , , , , . - - ,
O O O-110 VAC.(Clutch Power) -+ SSR ---7 Ilot l t t Mixer / Driver l Sram Signals
) : AUX ------------------------7 l
=
\
+ l i i To l l L- Clutch l :
l l I.Jower Supply i ! l Backup l Mixer / Driver - -_I , O , ( _ AUX - - - - + Scrams
, 1 110 VAC (Clutch Power) I
___(_____a V Neutral SSR Scram Signal Display Separate Neutral Line Roset
- - - - Separate llot Line
- -To Parel Display Piqure 7-8 Reactor Protective System Block Diagram
Table 7-1 VTak Protective System Setpoints Limiting Safety Limiting Condition
- Condition System Setpoint for Operation
, Power level (max) > 125% licensed power 100%
(2 channels) . l Primary Coolant flow (min) 20 gal per min 25 gal per min i j Intermediate Range Period > 5 seconds > 10 seconds (min) Core High Water Level (max) not higher than 1 in. Same j below the top of the
- core tanks Earthquake > 0.14g lateral Same acceleration l Manual operator discretion prior to exceeding
! any above limiting
! condition 7.3.1 Control Rod Drive Mechanisms Four control rods are provided for the VTAR: two safety rods, one shim rod, and one regulating rod.- Each rod assembly consists of four major subassemblies: 1) reel, 2) shaft, 3) shield labyrinth, and 4)
{ drive. The reel assemblies are essentially identical with the exception of the regulating rod, which has a ses11er Boral element. As the rod is withdrawn a stainless steel sheet flat spring (to which the Boral i element is riveted) is coiled around a drum. l Rupture of this " spring" or any drive component will cause the i insertion of the control element to bottom full travel by the uncoiling action of the stainless steel and element weight. Note that the elements are driven into the graphite reflector region directly on the i outside edges of the core tanks.
! The shaft assembly consists of the drive shaf t, coupling, pillow block bearings, and the position transmitter drive gear.
The shield labyrinth consists of borated polyethelene and lead with bearings on each end to support and align the drive shaf t. This shield arrangement virtually eliminates neutron and gamma streaming. , 1 O : , l 129
l The' drive mechanism will be described in detail in this section. This assembly and the entire elements themselves have undergone numerous modifications ' from their original configuration. These modifications are discussed in the following paragraphs. . In 1969, the drive shaf ts were lengthened new support plates (for F the CRDM's) f abricated, and the blade shrouding replaced with a heavier thickness aluminum. The shaf ts were lengthened because in the original design the CRDM's were recessed into the shield blocks and access for I maintenance was difficult. Covers were made to protect the mechanisms
! .from dust. Control blade shrouds were thickened from 0.02" to 0.032" due to binding experienced in the machined graphite alots at temper-j atures above - 1300F. -
4 In April, 1972, the bronze radial bearings on the interior of the reel assemblies -(for the shaft drum support) were replaced with stainless steel roller bearings. The original captive coupling method (shaft to real assembly connection) was replaced by a split shaft arrangement. This permits removal of the shaft without requiring a fuel
- _ transfer and core dismantling.
i I Early'in the facility history, two rod position indication i- potentiometers were added to provide safety rod console indication. 1 I
; Unfortunately, rod position indication potentiometer misalignments
! have resulted in several stuck rods over the facility history. l Accordingly, in 1982 a new method for mounting and adjusting the rod
; position indication and limit switches was designed. This method was
{ tested on Safety Rod One fcr several months and functioned extremely j well. Adjustment on all planes with respect to the position transmitter ; i drive gear was now possible with the added benefit of a significant 1 reduction in scram time (see Figure 7-9).
! Each safety rod is driven by a small split phase motor. Speed i
reduction is accomplished by an integral two-stage worm gear speed reducer. The drive speed is further reduced through a third worm gear
- on the drive shaft. Transmission of motion to the reel assembly is
- accomplished through an electromagnetic clutch, which sust be energized. For travel at extreme ends motion is limited by electrical 1 motor cutoff switches. In the event they fail, a mechanical-stop in the l reel assembly prevents damage and/or further travel. Design data are
{ given in Table 7,-2. The shim rod is virtually identical to the safety rods with 'the exception of the motor type, as this rod is used for coarse reactivity ' ?
, adjustments. The shim ~ motor is ' a three wire capacitor type, with instant reversal capabilities while running. Characteristics are described in Table 7-3, below.
l The regulating rod has a smaller surface area of boral and is .used-
- for-fine reactivity adjustments or automatic power control. Regulating r - Rod and drive characteristics are shown in Table 7-4 Its drive.
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Table 7-2 Safety Rods 1 and 2 CRDM Performance Characteristics Design Parameter Specification Scram Time < 0.8 seconds, from cutoff of current to clutch 1 Full Travel 16 inches; out direction only (except bypass) Drive Speed 6 in/ minute Full Travel Rotation 3060 Control Rod Material BORAL plate; 7 in x 8 in x 1/8 in Scram Torque 20'in - lbs Drive Motor Ratings 1/50 Ho, single phase 100 VAC, 60 HZ, split phase with 41 in-lbs torque @ 9.2 rpm j Position Transmitter he11 cal, 100 ohm potentiometer with 32630 travel Clutch 90 VDC, .110 amps, with 60 in 1bs j static torque Table 7-3 Shim Rod CRDK Performance Characteristics, Design Parameter Specifications Scram Time < 0.8 seconds, from cutof f of current to clutch Full Travel 16 in, In and Out directions t Drive Speed 6 in/ minute Full Travel Rotation 3060 Control Rod Materia 1 BORAL piate; 7" x 8" x 1/8" i ' ~ Scram Torque 20 in - lbs , Drive Motor Ratings 1/50 HP, single phase 110 VAC, 60 HZ, 3 wire reversible 5 ufd capacitor i l Position Transmitter helical, 100 ohm potentiometer with ! 32630 of travel j 0' ' C1utch 90 voc. static torque 110 ames. ith 60 in 1bs 132
l 1
' Table 7-4 Regulating Rod CRDN Performance Characteristics Design Parameter Specification Full Travel 16", In and Out directions Drive. Speed 30" minute In/27" minute Out Full Travel Rotation 306 0 1
Control Rod Material BORAL plate; 2"x2"x1/8" Drive Motor Ratings 1/100 HP, two phase 110 VAC, 60 HZ, low inertia servo motor with 3.5 oz - i in @ 1600 rpa Power Controller Feedback helical, 1000 ohm potentiometer with . 32630 trave 1~ Position Transmitter helical, 100 ohm potentiometer with 34270 travel i I mechanism is entirely different from that of the safety and shim rods. , It also does not have a clutch and must be manually inserted following receipt of a scram signal. The rod is driven by a small servo motor through a worm gear reducer to the shaf t ' of a two-unit potentiometer. This potentiometer transmits the force to a pinion gear which drives the spur gear and i rotates the drive shaft. i One unit of the potentiometer indicates rod position and the other is used as a feedback signal for the Regulating Rod. 7.3.2 Regulating Rod Control System The regulating rod control system is an automatic reactivity l control system used to maintain a pre-set desired power. Circuit use , i while at power is optional in that this system is. not required- for normal operation. This system is being redesigned. A description of the system will be forwarded as a revision when the system is completed. 7.3.3 Startup/ Interlock Bus The Startup/ Interlock Bus (Figure 7-10) requires that certain systems and/or conditions are in the necessary configuration prior to reactor startup. l The . original startup panel was replaced in 1969 and was very different from the original design both in function and appearance. In 1983, this panel was removed entirely and replaced. ;
)
133 1 __ ~ _- _
\
CT OP LEVEL s DUMP VALVE CLOSED
\
MIN COUNT RATE MIN SEC TEMP O , MIN PRI TEMP SR2 SR1
\ TOP TOP pg7 i PUMP -
ON
\
I VENT FANS s ON SCRAM SUS 3 COUNT y REG. SHIM y. SAFET y SAFETY R0D M R0D R0D 2 ROD 1
/s .
115 VAC (UPS) Figure 7-10 l VTAR STARTUP / INTERLOCK BUS 134
._ . _ . ~ - . _ _. .. . _-
1 h'
/' Functionally, the new startup/ interlock bus is basically the same 4
- b but ' differs in appearance from the original system. It closely resembles the original layout in a " christmas tree" type arrangement.
Operation of the control rod drive motors (CRDM) is a two-step F process. First, six conditions aust be met prior to closing the dump valve. These are: 1
- (1) Scram Bus Continuity - this ensures all main and backup scram signals are cleared.
(2) Reactor Ventilation Fan On - prevents operation if reactor
, ventilation is inoperable, which might lead to excessive levels ,
j of Argon-41. ' l (3). Primary Pump Running - prevents a startup without the primary pump. At low powar levels, flow is not required and this can j be bypassed with Renctor Safety Committee authorization. i (4). Minimum Primary Coolant Tamperature - this prevents operation if minimum primary temperature is outside operating limits. j (5) Minimum Secondary Coolant Temperature - two RTD's are used for signal input: secondary inlet and pan temperature. This does j not require the secondary pump to be running since (especially 1 during winter) initiation . of cooling prior to reaching the kW { range would result in a secondary temperature below allowable ;
! limits. The remainder of the piping is contained inside the building and water temperature is monitored- by the inlet RTD's.
l (6) Minimum Count Rate - this ' interlock prevents a " blind" startup 3 by requiring the source to be inserted into the core area. This completes the first step of the startup and allows ! the operator to perform the actual startup. At thi=> time the
- dump value is closed and within approximately one minute 15 j seconds (at 40 gpm), the core tanks are filled.
1 (7) Dump Valve Closed - this will actuate when the dump valve is fully closed. l (8) Core Tank Operating level - When the coolant / moderator in the core tanks reaches -a certain level above the fuel element, rod withdrawal will be permitted. The previous eight indicators constitute the " Christmas Tree" l l arrangement on the startup/ interlock bus. An additional indicator, core tank low level, is included for information only and provides no interlock functions. Power is now available to the Safety Rod One drive motor. When withdrawn to the fu1L1 travel, a contact is closed and Safety Rod Two can now be withdrawn. l l 135
-O
. When Safety Rod hto reaches full out, the final interlock is V engaged and power can now be applied to either the Regulating or Shim Rod Drive Motors.. The Startup/ Interlock Bus controls out motion only. The Shim and Regulating rods 'are the only two rods which can be inserted by other than a scraa signal. If an interlock condition occurs at any time, all rod out motion will be prevented. However, Shim and Regulating Rod in-motion capability is always available. The dump valve clutch requires a separate key switch to be energized.- This was dor.e to allow for greater flexibility and ease of maintenance regarding Technical Specification shutdown modes. 4 A list of the startup conditions and the setpoints is given in Table 7-5. These setpoints may be different from those required by Technical Specifications to allow for a safety margin. Table 7-5 Startup/ Interlock Bus Setpoints Condition Setpoint t Scras Bus Continuity N/A; requires that all scram signals be cleared i Reactor Ventilation Fans N/A; requires reactor ventilation O o= ce erett r ce 6 rat a i Minimum Primary Coolant > 75 F Temperature Primary Coolant Pump N/A; requires primary coolant Running pump controller to be energized 1 Minimum Secondary > 55 F Coolant Temperature l Minimum Count Rate > 3 counts per second on Source Range i Dump Valve Closed Dump valve fully closed Core Tank Operating Level Not less than the overflow pipe level in the core tanks Safety Rod One Fully withdrawn, ~ 16" Out 1 Safety Rod Two Fully withdrawn, ~ 16" Out O 136
. One last important item on this system is the bypass ' features required for testing. A Test and Bypass panel is located directly below the Startup/ Interlock Bus. It houses the bypass switches with
,' individual' light indicators and a locked plexiglass cover. Access is controlled by two dif ferently keyed locks required to i remove the - cover. Conditions which can be bypassed are addressed in detail in Technical Specifications. 7.3.4 Annunciator System J The VTAR annuciator system provides the operator with a visual and audible indication of a number of conditions which require prompt i corrective action. Setpoints are chosen such that time is available to correct the situation or, if required, shutdown the reactor prior to reaching a potentially hazardous situation. ( The system consists of the detectors, associated input circuits, < and two Panalarm units with the required switches for operation. Two Panalarm units are utilized. They are original equipment; i~ however, due to the simplicity of design, they continue to function extremely well. Each unit consists of an array 4 rows wide by 3 rows high with a . relay for each condition'and a control relay for the entire unit. O Operation is with 110 VAC and input alarm initiation is fail-safe. De-energization of the input sensor closes sets of contacts. This in turn energizes the respective relay and results in alarm initiation; a solid red nameplate with the parameter etched into it is lit by individual lamps and an audio signal is generated by an external 6-28 VDC "Sonalert" horn. i Depressing the alarm acknowledge switch results in silencing of the horn. TLe nameplate remains illuminated. When the condition is cleared, the lamp reset switch must be depressed - to clear the alarm 4 unit. In the event the annunciate condition has not cleared, the ! nameplate will remain illuminated. The following conditions are annunciated at the console: (1) Radiation In Primary Coolant - a Y detector is mounted in a well in the hold-up tank (for N-16 decay). The setpoint, which is 2.5 times the normal value, is 200 mR/hr and is used i to detect fuel element failure. (2) Low Level in Primary Shield Tank pressure switch in process pit. The setpoint is > 2 feet below the top of Shield Tank i and is used to detect tank leakage and thus prevent ! possibilities of abnormally high radiation levels. . 137
4. i q (3) Intermediate Range Period Under 10 seconds, - setpoint is 10
;V second period or less. This feature alerts the operator to a
- _ potentially unsafe rate of power increase and allows time for 1- reaction prior to automatic shutdown.
t (4) Building Evacuation - Radiation levels of 15mR/hr or greater in the west wall (building) or reactor room ventilation (stack) detectors initiate the sequence. This signal alerts the operator in the event of abnormal radiation levels or a release. (5) Automatic Cutout - Only in effect when Regulating Rod Control System is in Automatic. Setpoints are reserved pending final design and testing of the system. This feature also results in an automatic reversion to manual control. (6) Loss of Power Intermediate Chamber - Alerts operator to loss i of signal detecting capability to Intermediate Range (IP.) detector. Actuates when IR high voltage power supply output i drops to zero. (7) Loss of Power Keithley Chamber - Identical to (6) above except signal input is from Keithley high voltage power supply. 4 1 l (8) Overtolerance Area Monitor - Input from a y detector and
- circuitry. The detector is mounted on the east wall of the reactor room above the process pit. Set point is 5 mR/hr.
] (9) High Moderator Temperature - Platinum-wire RTD and ] circuitry. The RTD is in the primary outlet piping. Basis
- for setpoint (> 165*F) is to prevent excessive fuel element I
peak central temperatures and allow time for the operator to
- determine and correct the cause prior to exceeding the 4
Technical Specification limit. (10) Water in the Process Pit - A detector in the process pit sump actuates the alarm at 3/4 or less water level. Indicates leakage or rupture of primary and/or secondary piping. Also used to indicate core tank overflow and containment of all
- potentially radioactive fluids prior to sampling.
(11) Fission Products in -Stack - Alarms at 200 counts per second on 1 Air Particulate Fission Products Monitor. Alerts operator of l a possible loas of fuel element integrity. J
- (12) Power Recorder Of f-Scale _ - When the chart recorder reaches 5%
or 95% of scale readings, the operator is alerted. Typically, 4 this would require a range change on the Keithley instrument. ' j (13) Secondary Coolant Low Flow - Warns the operator that a failure has occurred in the secondary cooling ' system which could .! effect heat removal capabilities. i O i 138 i
4 (14) Low Negative Pressure in Reactor Room - If ventilation is lost or if a door is opened, an alarm will occur within 30 seconds. (15) Loss of Uninterruptible Power Supply (UPS) - When an internal circuit f ailure occurs in the UPS, this will be sensed and an automatic transfer . occurs whereby the power supply is bypassed. t . (16) Emergency Core Cooling System Initiated - This alarm is used to inform the operator that the primary side of the ECCS is operating. (17) Emergency Core' Cooling System Low Secondary Pressure - Following initiation of ECCS, city water must be used to cool '. A motor-operated valve aust be the ECCS heat exchanger. opened from the console or the ECCS secondary is not fully j operable. i 1 (18) High Secondary Coolant Temperature - When the outlet
, temperature of the primary to secondary temperature reaches a
- j. value of 1100F an alarm occurs.
(19) Pump On/ Valve Shut - An interlock prevents starting the i secondary coolant pump with the discharge throttle valve open. This prevents a sudden insertion of cold water through l the heat exchanger. When the pump is running with the discharge valve shut (as when first starting) or if an attempt
. is made to start the pump with the valve open, this alarm will occur.
This feature serves two functions: (1) to inform the ]' operator to slowly . initiate secondary cooling following a pump start and, (2) to inform him when a procedural error has been
; committed.
The approximate facility set points for actuation are given in Table 7-6.
+
1 7.3.5 Rod Position Indication System The rod position indication system (RPI) monitors and displays the relative positions of all control rods in the VTAR. Each rod has an individual position display in the form of an LED bar graph meter with top and bottom lights. A' digital voltmeter with a j selector switch is used to provide a 0-100% indication.
-O 139
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- - ,,m n -e , , , ,, ,m.s ~g ,..,.--nn a + -. .
l Table 7-6 VTAR Annuciator Setpoints Condition Setpoint Radiation in Primary Coolant 2.5K normal; 200 mR/hr Shield Tank Low Level > 2 ft below top of tank i Intermediate Range Period < 10 second period Building Evacuation > 15 mR/hr building or stack detectors Automatic Cutout Loss of Power, Intermediate 0 Volts detector power supply Loss of Power, Keichley 0 Volts detector power supply Overtolerance Area Monitor > 5 mR/hr area detector Iligh Moderator Temperature Primary outlet temperature
> 165 F Water in Process Pit Sump level < 3/4 full Fisson Products in Stack > 200 counts per second on APFPM Power Recorder Off-Scale 5 or 95% of recorder full travel Secondary kw Flow 150 gallons per minute Loss of Negative Press. Reactor No differential pressure Room between Reactor Room and any adjacent space Loss of UPS UPS internal circuit failure ECCS Initiated ECCS pump energized and ECCS on/off valve open ECCS Low Sec. Pressure < 20 psi in city water supply line High Secondary Coolant 1100F Temperature Pump On/ Valve Shut (1) Discharge throttle valve open when attempting to start pump.
! (2) Discharge- throttle valve shut with pump running. ! h 140 i i l l
l l f' The bar graph displays by an increasing red light (LED) display. x
- That is, when the rod is fully withdrawn, the display will be fully illuminated and the top light will be on.
-Insertion of the rod will result in a completely dark display with an illuminated indicator for the bottom limit switch. .
l A 5 volt DC power supply is used to develop the voltage across the ; } 100 ohn rod position indicator potentiometers. . f l Specifications for the Datel-Intersil DMB bar graphs and digital ; voltmeter are given in Table 7-7.. 5: 4 Table 7-7 VIAR Rod Position Indication Circuits Function Characteristics 4 Individual Display LED Bar Meter with 5% resolution (0-16 inches) i Group Display . LED Digital Voltmeter with l (via selector switch) percent of full travel 1 indication (0-100% corresponds j- to 0-16 in.) Accuracy
- 1%
i 7.3.6 Primary Coolant Flow Monitor l The primary coolant flow monitor is a microprocessor based instrument designed to indicate flow rate by measuring the frequency of the signal generated by the flow tr noducer in the primary system piping. The paddle wheel transducer works as a small AC generator. Permanent magnets mounted in the paddle wheel rotate past a coil, inducing a signal whose voltage and frequency are directly proportional to the fluid velocity in the pipe. The flow meter electronics convert the sine wave signal from the transducer -into a t'ransistor-transistor logic (TTL) signal . for use by the microprocessor and for external. equipment. Variations in flow are damped by internal averaging with a time constant of eight seconds, yielding a non-flickering readout which ! is updated every two seconds. The display is a three and one-half digit j liquid crystal display. 1-i A frequency-to-voltage converter converts the external TTL signal to a voltage and sends this voltage to a comparator circuit. The
- comparator drives a relay which sends a scraa signal under low . flow
i conditions. A block --diagram of the primary flow ~ monitor is given in figure 7-11. .) 7.3.7 Secondary Coolant Flow Monitor The secondary flow monitor is an ultrasonic instrument which j measures flow rate by comparing changes -in the frequency -of sound waves 4
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l INPUT .
' MODULE ANO i ; F Q CY SI C ND.
PLOWSENSOR I PRESCALE ( PA DDLt.- l WHEEL) I SQUARE WAVE l OUT = FLOWSENSOR V V, . FREQUENCY NOTES: (1) Vow-REFERENCE VOLTAGE
-2 F/F (2) f -FLONSENSOR FREQ.
(3) N -FREQ. PRESCALE SWITCII, (=0) ~ Figure 7-11 Block Diagram of the VTAR Primary Flowmeter
I l
\
q caused by the flowing liquid. Figure 7-12 is a functional block diagram V of the flow meter electronics. An upstream and a downstream transducer are mounted in the secondary piping and are connected to the flow meter electronics as shown in Fig. 7-12. Short duration pulses, produced alternately by two pulse generators, excite the transducers, causing them to vibrate at resonant frequency (in the megahertz range) and thus producing sound waves in the coolant channel. After transmitting, each transducer is switched to the receive mode. The transmitted sound waves travel through the water to the opposite transducer where they'are converted to an electrical signal and amplified in the receiver section. As soon as the pulse amplitude reaches a preset level, the leading edge detector triggers the appropriate pulse generator, via a timing circuit, and tha cycle is repeated. This establishes a pulse repetition rate which is dependent on the transit time of the sound waves through the water. Since the effect of the motion of water in the pipe is to slow the sound waves travling upstream and to speed up the sound waves traveling downstream, the transit time, and hence the frequency, of the upstream and downstream pulses is different. The difference between the two frequencies is a direct measurement of the rate of flow of the water in the pipe. The difference frequency is fed to a frequency-to-voltage converter where a voltage is developed which is proportional to the flow rate. This output is available for external use and is connected to a O) compensator circuit which gives an annunciation of a low flow D condition. This voltago is also sent to a voltage-to current converter which gives a 4 mA to 20 mA current loop for remote flow rate indication at the control console. l The entire flow meter is fail-safe in that if any part of the flow i meter system fails, the circuitry drives both voltage and current outputs to a low condition, producing a low flow anunciation. (vl l 143 )
O - O . O L AGC FDOK u
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i WA N ER$ RECEIVER > -
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) FAIL ;
e i SWAMP y SAFE _ l PHASE. Two WIREPROCESS .' LOCK l ladplcAToR 8 far l AfD
- FREO TO PULSE l N 4 qi20 ~
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CURRENT CONSOLE i r Figure 7-12 Secondary Coolant Flowineter Functional Block Diagram 4
- 8. ELECTRICAL DISTRIBUTION 8.1 System Overview
} T'ae Virginia Tec.h Argonaut Reactor (VTAR) is a 100 kWt Research Reactor and as such does not generate electric power. Electric power must, therefore, be supplied by an outside source. The Virginia Tech Electric Service supplies the normal electric power needed by the Nuclear Reactor Laboratory. An Uninterruptible Power Supply (UPS) supplies regulated AC power to vital console loads (Table 8-1) and, in the event of an extended loss of power (> 10 min), will provide power to
- monitor vital equipment after shutdown. See Figures 8-1 through 8-3.
0.2 off-site Power Distribution 8.2.1 440 VAC Distribution The 440 VAC three phase (34) power is supplied from a breaker panel ! in Room A-11 (Robeson Hall Transformer Room) to the secondary coolant pump and cooling tower pan heater. 8.2.2 110/220 VAC Distribution The 110/220 VAC power is supplied f rom the electrical load center, located in Room A-15 of Robeson Hall. Power is distributed from the O~ load center to Breaker Panel Distribution Centers in Rooms B-4, A-9, B-1 15 and on site in Room 10 (the Reactor Room). All power supplied to these panels is 220 VAC 3$ and. from there is supplied, as needed, to 110 VAC power.
; 8.3 On-Site Power Distribution
- 8.3.1 440 VAC Distribution i Two pieces of equipment are run from 440 VAC, 39 power suppliest the secondary coolant pump; and the cooling tower pan heater.
' The cooling tower pan heater is supplied from Room A-11 through a breaker disconnect and controller unit, both mounted on the roof of l Robeson Hall at the cooling tower. The secondary coolant pusp is fed l f rom Room A-11 through a knife switch disconnect and controller unit, both mounted on the east wall of Room 10 near the secondary coolant pump. 8.3.2 110/220 VAC Distribution Low voltage on site power distribution is supplied by the six off-site distribution centers previously mentioned. Breaker panels B-4A and ! B-48 in room B-4 supply lighting for rooms 106 and 108 of the facility I O V 145
TRANSFORMER ROOM A-ll l ROBESON HALL ROOF SWI'ICH i
' YARD tt
)(
J( n-a n CONTROL.
- % ; COOLING TWR.
1=
) ." PAN THERMOSTAT
- 440/220 VAC - I 416/440 VAC C PANEL 1A-ll
( """- REACTOR ROOM 10 SEC. DISCONNECT . ( COOL g
- COMROL.
BRR TANPL-1A-15 l n_ ^ )
- LOAD CENTER ROOM A-15
^ ^
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a. I b LOAD CENTER
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2- . 4 B-4B B-4A A-9A A-9G bh Rb kk ' i OUTLET} M AIR COND. RM 106 LIGHTING p -} ,OUTLETS , AND 4 LIGHTING Hmm , MGHTING l, RM.109 ' LIGHT
- I ' f i RM 106 '-
- CRANE 240 VAC i -
l 2*C I E LOADS *
"^" ~A 4
_, " LOADS *
" ' ' "~
PANEL 10-A j 6-A ROOM 6 ROOM 10 i M; - N :2- - l G --- LOADS * ,,; tOAoS .
- i. Ca ""~- U.P.S. .
a E' .l PANEL 10B-A PANEL 10-B UNREGULATED REGULATED ROOM 108 FOR LOADS SEE TABLE 8.1 . O ELECTRICAL DISTRIBUTION FIGURE 8-1 146 a
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7 7 l ** l se y . 5 p.w w =va Panel Locations u 4 6^* 1. Reactor room panel
, 3
- 2. 440V panel (secondary pump, cooling tower pan heater)
[a# * "',,**"' 3. Supply panel for panel in room 6 and room 108 building load center
- 4. Light panel, reactor room and lat' gu,o 5. Outlet panel, reactor room and lub t=n oer ,, 6. Equipment panel, reactor room and cooling tower i 'A 3 .,
Rl aa aa lsaa. 3' Disconnect locations qe 3 I a. Secondary and primary pumps disconnects
,,, b. Crane power disconnect
_ , g, j
- c. ECCS pump disconnect 4
I
- P ; GROUND FIDOR PLAN Piqure 8-2 Power Panel locations
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- 1. Outlet power panel for console room and NAA lab I 3 e 2. Light panel for console room and NAA lab z.
co 3. Console room power panel LAS 606 ggag3
, a os All panels. shown are 230 vac. 3 phase I .
i LAS g ,g , a. ' 5 g i i ll i PIRST PIDOR PLAN Pigure 8-3 Power Panel Incations 4
and also power for the air conditioning units in room 106. Breaker panels A-9A and A-9C, in Room A-9, supply reactor room lighting and reac*or room 110 VAC outlets, respectively. The load center in room Table 8-1 On-Side Breaker Panel Loads Panel 10A (Reactor Room)
- 1. Dump-tank heater 5. Process Pit Outlets
- 2. E.C.C.S. Pump 6. Sump-Pump, Process Pit
- 3. Main Ventilation Fan 7. Hot Cell Manipulator Arm
- 4. Booster Ventilation Fan Panel 6A (Room 6)
- 1. Heater Tape, Cooling Tower 4. Primary Coolant Pump ,
- 2. Heater Tape, Cooling Tower 5. Secondary Flow Monitor
- 3. Cooling Tower Fan Panel 108A (Room 108, Unregulated)
- 1. Air Conditioner #1 4. Console Distribution Bus
- 2. Air Conditioner #2 5. Uninterruptible Power Supply
- 3. Emergency Lighting 6. Building Evacuation Horns Pane 1088 (Room 108 Regulated)
[;
- 1. Console Distribution Bus Console Distribution Bus Regulated Loads (Panel 1085 Supply)
- 1. Nuclear Instrumentation 8. Primary Coolant Conductivity
- 2. Keithley Pico-Ammeters Monitor
- 3. Radiaton Monitors 9. Regulating Rod Drive
- 4. D.C. Power Supplies 10. Reactimeter
- 5. Air Particulate Monitor 11. Annunciators
- 6. Primary Temperature Monitors 12. Chart Recorders (Onega &
- 7. Primary Flow Monitor Rustrack)
- 13. Dump-Valve McLot Unregulated Loads (Panel 108A Supply)
- 1. Secondary Temperature Monitors 5. Close Circuit Television
- 2. Secondary Flow Monitor 6. Chart Recorder (Westronic)
- 3. Safety and Shim Rod Motors 7. Cooling Tower Temperature Monitor
- 4. Rabbit System Gas Solenoida 8. Ambient Temperature Monitor 149 l
l
._. ~ _ _ ___ . _ _ _ . - . . _ _ - _ _ _ . - _ . _ _ _ _ _ . _ . . _ . _ _ _ _ _
4 L i p 6
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i' i f A-15 supplies 220 VAC 3$ power through the crane knife switch disconnect I to'the crane coctroller and all'220 VAC 3( outlets in room 10. The load j center is also the power source for the breaker panel 10A on the reactor < , room south wall (load shown in Table 8-1). Bresker Panel IA-15 in room i A-15 is the power source for breaker panel 4 6-A in room 6' on the east vall and for breaker panel -108-A, the unreguisted console distribution i panel in room 104 (north wall). The loads are listed in Table 8-1. h j unregulated breaker panel 108A supplies normal power to the UPS which j regulates the 220 VAC input and supplies a 220 VAC 1$ regulated output i to the regulated console breaker panel 108-5. The UPS is located ! i directly behind the reactor console in roon 108; its operation is [ i covered in part 4 of this chapter. The regulated console breaker panel - supplies power to the vital components of the console and is located on
- the operating platform, room 108 (north wall). ;
8.4 Uninterruptible Power Supply , t j 8.4.1 Systes Description f ! The Uninterruptible Power Supply (UPS) is a voltage regulator / power l 1 supply and supplies the vital components of the reactor console with 110 VAC regulated power. Figure 8-4 is a functional block diagram of the : l
- Uts. !
i I j The UPS comprises three major components: the input power i converter / charger, the battery bank and the inverter circuit. The UPS t ! takes a 220 VAC 1$ input converts this to a DC voltage . to drive the ! I inverter, and also maintains a charge on the battery bank, and then . ! gives a regulated 110/220 VAC 14 output to the regulated console distribution panel. In the event of a loss of 220 VAC input power, the l UPS instantaneously converts to the battery bank back-up power source. l The transfer of power takes place without interruption of the regulated ! j output power. The back-up battery bank power source will provide full i r i load output power for a period in excess of 20 minutes giving sufficient !- time to ensure that vital reactor functions are safely monitored in the r j event of an extended power outage. 1 4 8.4.2 Desertation of Systes Ooerstion ! j 8.4.2.1 Input Power Converter \ j The input power converter (IPC) converts and conditions the input i AC power (110/220 V) for operating the ' inverter, the control circuits l and charging ' and/or maintaining the battery bank ' to a fully charged ,' condition at 44 VDC. t 8.4.2.2 Battery Bank , i The battery bank supplies on demand eastgency power in the event of l normal line power f ailure and consists of four 12 vult, deep-cycle ! l O-1 ) I s i i 150
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batteries connected in series to the output of the IPC. 8.4.2.3 Static Inverter The static inverter accepts a DC input voltage from the IPC or battery bank and provides a 110/220 VAC single phase output. The 's inverter is all solid state and uses power transistors to switch the DC input. This switched DC input waveform is filtered to provide the AC
? sine wave output. The static inverter is composed of eight component parts: the timing and control board, the driver board, the driver transformer, power transistor banks A and B, the current-sense transformer, the output transformer tank, and the voltage-sense transformer.
4 e i 2 6 n h. 5 EO cl 152
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- 9. AUXILIARY SYSTEMS 9.1 Fuel Storage and Handling 9.1.1 New Fuel New fuel-is kept in one of two locked, key controlled storage pits inside the reactor room. Only a single fuel element is allowed in each storage hole. Calculations were made to verify suberiticality even with the pits fully flooded (see Chapter 15). Any fuel handling is done by Senior Reactor Operators only.
Information regarding fuel access control or the physical 4 protection of stored fuel is withheld from public disclosure in accordance with 10CFR2.790(d). 9.1. 2 Spent Fue1 Fuel transfers are accomplished using a steel and lead transfer l cask. Since decay heat is not a problem, the irradiated fuel is stored in a storage pit until it decays to lower levels. Again, only Senior Operators are allowed to handle fuel and key inventories are done on a regular basis. Surveys are performed during fuel transfers and fuel transfers are performed in accordance with the VTAR operating procedures. An analysis has been performed on. expected releases in the event of a dropped fuel element (Chapter 15). Fuel is manipulated in the shield tank or, if levels are low enough, in the reactor hot cell. Remote handling and manipulator coola l' are used in the shield tank. An AM7 aaster/sisve manipulator (Fig. 9-1) can be used in the hot cell. The hot cell hae a lead glass window for 4 viewing purposes. . . 9.1.3 _Reacter Roon Oveyhead b y The VTAR facilit.y is equipped with a 10 ton overhead crane. The crane is used for moving ' shield blocks for the reactor and for fuel transfer between the reactor core and fuel pit.- The crane is also used
~
I l to transport various heavy loads. l The crane - is a Robin and Myers, F5 double girder, three moto r ,- overhead crane. Support for the crane is supplied - by - two 'I' beams, supported at 4 points - on each beam by two center posts and by the wall-on either end. :A railroad type rail is mounted on top of the beams for the crane to roll on. The bridge travels in a north - south direction, I l O 153
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Figure 9-1 AMF Master / Slave Manipulator 7
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I - at the rate of 30 feet per minute. The hoist is mounted on a movable trolley which hangs f rom the . lower part' of .the bridge. The trolley > travels in an east - west direction, also at 30 feet.per minute. The hoist can lif t or lower an item at 17 feet per minute. .The hoist is powered by a 12.5 HP motor, with a-hoist brake built into it. 4 . The t entire reactor room can be covered by the crane, with two exceptions: (1) The southwest corner of the reactor - room, where the heating system for the reactor room.is located. and (2) The crane- cannot lift an item that is next to a wall i (approximately 3 feet), due to the construction of the crane j itself. Crane power is switched from a disconnect mounted on the wall ' of the reactor room, in the southwest corner. Power for the disconnect is j provided from the switch room (A-15) in the basement of Robeson Hall. In case of an emergency (e.g., fire, etc.), crane power can be shut off
- from this location.
Power is transmitted to the crane via hard drawn copper wire strung . below the support rail, and mounted on porcelain insulators. A similar arrangement is used on- the bridge for power and control signals to the hoist and trolley motors. - Movement and lif ting / lowering of the crane is accomplished by a 6
- i. button control box, which hangs from the center of the south side of the bridge. The buttons are momentary contact, that is the buttons must be depressed to operate the crane. The buttons control relays mounted oo ,
the bridge trolley. The relays for a given control function are. mechanically interlocked, (e.g., up/down, left/right, forward / backward), so that if both buttons are pushed down, damage will not result to.the crane. t l Soveral safety features are incorporated .into' crane- hoist l' operations in the form of limit switches and a weight mechanism. i The limit switch mechanism functions as a stop. for. the . crane - hoist
~
hook at the maximum upper travel. point. This switch consists of a weight suspended from a lever by a small chain. When.the block reaches the maximum upper travel point ' it lifts the weight, allowing the lever l to rotate... The lever then actuates the limit switches. I l When this occurs, the limit switch interrupts' control current, ~ stopping the hoist ~ in : that position. In. the, event of motor brake - l l failure or excessive hoist hook "drif t", a second switch immediately - reverses the ~ motor . and lowers the hoist hook and load downward a few l l inches.- The hoist can then.be operated by depressing the "down" button.. I d' 155
- '- r~ -,- , , . . ,, ,, , , _ _ _ _ __, , _ . , .
'A third emergency stop switch kills the entire control circuit.
This switch is used only if the limit weight should break and fall to the bottom block in which case the entire control circuit is de-energized. This action is done _because without the weight, the other two limit switches are inoperable. ! Only qualified- Reactor Operators are Allowed to operate the crane. Maintenance . is performed on an annual basis which includes an inspection, lubrication, and oil changes (for the respective motors). No loads are transported over the core area when the reactor is operating. Protection against unauthorized crane use is withheld from
! public disclosure in accordance with 10CFR2.790(d).
'- Specifications for the crane are given in Table 9.1. Table 9-1 VTAR Overhead Crane , Component Description Crane Robbins & Myers Hoist Electric, Model #F5D1B, S/N 3064 Holst Motor 208 VAC, 3' phase high torque, 4 squirrel cage induction, 12.5 HP, I . S/N HIM 11668KP Trolley Motor 208 VAC, 3 phase squirrel cage induction, 1.5 HP, S/N HMM11592KP Bridge Motor 208 VAC, 3 phase squirrel cage
.-induction,-1 HP, S/N HMM11592Ko i
9.2 Water Systems I 9.2.1 Shield Tank Water Maintenance System I t. l The shield tank water maintenance system serves two purposes: (1) the filtratioa section filters the water of any floating or suspended debris, and (2) the demineralizer ' maintains the purity of the water to
- i. prevent corrosion and algae buildup.
The . filtration system consists of a curf ace ckimmer, a filter / pump j unit, and the connecting tubing. The surface skimmer is used to filter out any large debris that may-
=be drawn into the intake of the-filtering unit.
The filter unit consists of a prefilter before the pump and the ' l ' main filter after.the pump. !O 156 i
In operation, water is drawn in the surface skimmer, which collects any large debris that is in the water. The prefilter stops most of the smaller debris before it enters the pump. The main filter removes any small particulate matter, 'and the water is then returned to the shield l ! tank. The pump is capable of pumping up to 12.5 gpm. The main filter ] has a total surface area of 16 sq. f t. The demineralizer loop consists of a submersible pump and a demineralizer. The pump, which is submerged in the shield tank, draws water from the tank and casses it through the demineralizer. Then the water is returned to the chield tank. The demineralizer produces water that is at the level of triple-distilled, 15-18 megohms-cm resistivity. A pilot lamp in the demineralizer de-energizes when the resistivity of the water falls below 200 k ohms-ca. In operation, the demineralizer is run continously. The filter unit is run as needed. 9.2.2 Primary Make-up Water System The make-up water system is used to provide the water in the primary cooling system and in'the shield tank. The system draws its water from a 1 in cooling water supply pipe
. O through a 1,2 in ,1,e. There are two control valves in this line.
solenoid valve is used to control the water flow. A A timer can be used if needed (i.e., filling the dump tank), or bypassed if a continous flow is desired. A manual shutoff is used for servicing of the demineralizer circuit. Th2 water then passes through a 20 micron filter and into the demineralizer. As the water passes through the demineralizer, all anionic and cationic constituents are renc,ved. The quality of the water upon exiting the demineralizer is at the level of triple-distilled water (15 to 18 uegohns-cm resistivity). On the demineralizer there is ,a pilot light which de-energizes if the resistivity falls below 200 k ohms cm. The demineralized water is split into two 1/4 in lines, each of which has manual control valves in them. One goes to the dump tank in the protest, pit. This line is used to fill the dump tank, and for make- . up water as it ' is needed. The other line is - used to fill the shield tank. At the top of the shield tank is a solenoid valve, whi& through a float arrangement,' prevents exceeding a preset water level. To ' use the system, the manual inlet valve is opened to the city
- water supply. - Either of the two outlet valves is then opened, depending-on whether the dump tank- or the shield tank is to be filled. Depending on the filling condition, the inlet water can be timer controlled, or the timer ~can be bypassed.
i 157
. ~- ,_
) For shield tank filling, the float controlled valve on the shield
' tank must be switched on for water flow into the shield tank.
9.2.3 Secondary Make-up Water System The secondary make-up water system is used to provide water to the secondary system. The system takes its water from the main building supply line through a 1 1/2 inch pipe. This pipe runs to the roof of Robeson Hall and terminates at the cooling tower. The pipe contains 3 isolation valves, a pressure reducer /back-flow preventer and a drain down valve to empty the pipe for maintenance. Cooling tower fill and level control is achieved through a 1.1/2 inch float valve attached to the end of the pipe in the cooling tower. The float valve is adjusted to shut off at a level of approximately 4 inches below the tower overflow drain which is the normal operating
'Jevel during pump operation.
Also included with the float valve is a 3/4 inch hose tap which l facilitates periodic cleaning of the cooling tower. 9.3 Reactor Room Ventilation System 9.3.1 Description of the System
, The ventilation system for the the VTAR facility serves the following functions:
(1) To remove Ar-41 from the core region, process pit, and reactor room and to exhaust it into the atmosphere after dilution; (2) To provide negative pressure in the reactor room; if small arounts of contamination should occur, this feature helps to keep it inside the reactor room; (3) To provide an exhaust vent for the waste N2 gas from the , rabbit system; (4) To allow for the ' monitoring of the exhaust stack for any radiation release to the environment; and (5) To exhaust the chemical fume hood in room 6. The negative pressure in the ventilation system is provided by a centrifugal fan (ILG BC-182) mounted on the roof. The output from the fan is directed into the input of an axial fan (ILG U-PDE 369), which dilutes the stack exhaust and provides additional flow (Figure 9-4). 158
. . _ ~ ._. . .- .__ _ -- - -
l q U Intake from the ventilation fan is provided by an 18 in. diameter galvanized duct which runs- from the roof down to the intake duct in the reactor room. All seams on the duct are soldered to insure air
.. tightness. Radiation monitoring is provided in the stack in three locations. In room 306 radiation monitoring is provided by a Victoreen
- detector (Series 857-20) mounted in the stack. In room 206, sampling ports are provided (presently closed off). In room 108 (the ' control room) sampling ports are provided for the operation of an Air Particulate Fission Product Monitor in addition to TLD's mounted in the duct. 4 A connection is provided to exhaust the fume hood in room 6. The
- exhaust for the rabbit is also provided in this connection. A damper is provided to close off the fume hood if necessary.
The 18 in. stack passes through the wall behind the' control room (108) and into the intake ducting in the reactor room. There are five openings in the duct; three are for room ventilation. One is ducted into the core for Ar-41 and moisture removal, and the other is for Ar-41 and moisture removal from the process pit- by a booster fan running into
~
i i the intake duct. Fan motor control is accomplished by two controllers mounted in the reactor room, one controller for each f an. Ventilation is shut down, e until manually. restarted, for two conditions: O (1) Building alarm is triggered for either the stack or building i radiation monitors. The system cannot be restarted until the
- alarm is reset.
(2) The fans can be manually shut down f rom the fan ' controls in the reactor room and from the control room.
~ ! In all car.es, the ventilation system can be restarted only from the fan controls located in the reactor room.
1 A pressure switch.to indicate negative pressure in the reactor room
- operates an alarm en the console to indicate the losa of ventilation in the reactor room.
4 9.3.2 Operation of the Ventilation System The ventilation system for the reactor is normally operated et all times, unless the building alarm has _ been triggered. The ventilation cannot be run until the building alarm has ' been reset to -its normal. operating state'. The - reactor cannot be operated unless the ' ventilation system is e operational. The startup/ interlock bus prohibits dump valve closure
~unless reactor ventilation is operable. If ventilation fails ' during i
O 159. l
l l operation, it must be restored within 5 minutes or the reactor is shutdown by the Reactor Operator. In no case will more than two restarts of the system be attempted without reactor shutdown. All doors to the reactor room must be closed during operation. If a door is open while the reactor is operating, it must be closed within 30 seconds, _ or a loss ' of negative pressure annunciator on the console will alarm. If either the stack or building radiation monitors set off the building alarm, the ventilation system will be shut down automatically, precluding an uncontrolled release of high contamination levels, as would be expected with the associated alarms. When in oper.stion, air is drawn from the process pit (via a booster fan), the core region of the reactor, and the reactor room itself. This air is passed through the stack to the intake of the exhaust fan, located on the roof. The exhaust is directed into the booster fan i intake, and is mixed with outside air. The original reactor room ventilation / heating system has been , retained, but for heating purposes only. The- original outside intake / exhaust points have been closed ,off. The present system > recirculates the reactor room air. In the case of a building alarm, the heating system is shut down, and cannot be run until the ventilation system is running again. The' ventilation ratings are given in Table 9-2. Refer to figures 9-2, 9-3, and 9-4 for diagrams of VTAR ventilation components. Table 9--2 VTAR Ventilation Characteristics C,omponent , Description Blower fan Centrifugal Exhaust Unit 855 rpm l Blower fan Vain-axial booster fan 855 rpm J c Blower fan Dump tank centrifugal booster 1725 rpm, 1/8 hp fan (process pit) Total flow rate 12,700 cfm at stack discharge
- O 160-
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Booster Fan
~
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Centrifugal Blower Roof ) 1. Stack Monitor
/ 2. Sample Point
- 3. TLD
- 4. APFPM 14\ 5. TAP
- 6. Rabbit System Exhaust Room 308
\T 7. Reactor Room Ventilation Plenum
-8. Dumu Tank Booster Fan Duct
- 9. Core Ventilation Fan Duct
- 10. Fume Hood l
Room 208 2-O -
'3 _ .
Room 108 Room 10 (Reactor Room) 4-. 5 '-~ ( _J l 9 I
\ 7 m, l 6 U "Tb a Reactor Room 8 Damper Valve !
4 l k l Figure 9-2 VTAR Ventilation Diagram O 161
.. - . .. .. .~ . . ._ ~_ . ~.
O O o , (Upper) Operating Closure l I I I I 1 l
- 1. Central Stringer (Lower) Operating- -
- 2. Off-set Stringer Closure -
- 3. Core Tank Shutdown
\ Closure o
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,3 Core Shielding Ventilation .
g Duct w 1 2 3 Core llegion 0 :O Dump Tank Dump Tank l - Booster Fan Core' Overflow Drain Figuro 9-3 Schematic Diagram of Purge Points
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163
1 O 9.4 Other Systems 9.4.1 Pneumatic Rabbit Sys' tem The rabbit system is used to transport samples to be irradiated into the central core regon, or into the thermal column region. After a predetermined amount of time, the sample is returned to the transmitter, a glove box in room 106. Samples can also be shot in or out by manual control. There are presently two rabbit systems in operation. The southeast j offset stringer has been removed from the central core region where one of the rabbir. tubes terminates. The other rabbit tube terminates in the thermal column region. Each rabbit transfer tube is independently controlled. The thermal column rabbit presently uses sound sensors attached to the transfer tubes. The central stringer rabbit uses optical sensors and an integrated control circuit. The two systems are functionally the same except for minor differences in clock control. The rabbit operator selects the time interval required to activate a sample and dials the time on the thumbwheel switches. The operator then pushes the trigger button that causes the control logic to open the supply gas solenoid valve. This pressurizes the rabbit tube with nitrogen gas and propels the capsule into the reactor. As the capsule passes the optical sensor nearest the reactor, it sends a signal to the control logic instructing it to shut the supply gas solenoid valve and turn on the timing clock. The clock starts timing and the elapsed time
! (0 to 999 seconds) is displayed on the front panel. When the elapsed time equals the preset time the counter stops counting and signals the control logic to open the return gas solenoid . valve. This supplies i nitrogen gas to propel the rabbit capsule back'to the glove box. As it enters the glove box, an optical sensor on the tubing signals the i control logic to shut off the return gas solenoid. The system can then be reset for the next shot.
The ' exhaust. gas from the rabbit is passed through a prefilter and then through an absolute filter. The gas is then vented to the exhaust stack from the reactor room. (See attached block diagram, Figure 9-5) The rabbit system can be used_only if two conditions are met: l , (1) reactor operator must enable the rabbit 'from the reactor console to shoot a sample, and (2) the building alarm mst be in a non-alarm state. To use the rabbit to activate samples, the following initial conditions must be met: J
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. (1) ' The radiation monitor 'in the rabbit system glove box mst be operable 1 (2) The rabbit control system mst be energized and nitrogen gas )
available to the system; bottle pressure mst be at minimum 100 psi, and regulated pressure to the rabbit set at 30 to 35 psi; (3) The reactor ' m st be critical and power stabilized at the desired power level for sample irradiation; and,
' (4) The necessary irradiation requesc is properly completed and on the reactor run sheet.
The general procedure for the operation of the rabbit is as follows: The rabbit operator requests the reactor operator to enable the rabbit system from the reactor control con:: ole. he rabbit operator then
; shoots a blank rabbit (i.e., no sample material) to check out the rabbit system operation. If the system checks out as functional, the blank will be shot back and forth for a minimum of two more times to clear any Ar-41 that may be trapped in the system. The glove box receiver can then be opened and a sample.can be loaded into the receiver. Then the
. rabbit operator informs the reactor operator that a sample is ready ' to be shot, what the sample is, and the time it is to be irradiated. When permission is granted to shoot the sample, the reactor operator will
- enable the system from the console. The rabbit operator will then set the internal timer and shoot the sample in by triggering the controller.
^ When the sample enters the reactor, the operator insures that.the supply gas .is automatically turned off and the countdown timer is j running. ' Once the sample has returned, the rabbit operator _ resets the i controller and visually inspects the- sample to see that it is in the
- glove box receiver and intact. He then informs the reactor operator.
The rabbit operator observes the radiation level on the rabbit monitor. In che event the sample is > 2 R/hr, he exits immediately and informs the SRO and RRSO. If the radiation level of the sample .is within permissible limits, he retrieves the sample f ro's the receiver and l closes the receiver afterwards. i .After removing the sample from-the receiver, the rabbit operator i visually inspects the vial for any defects before removing it from the glove box. i The following. steps are followed when shutting down the rabbit
-system:
(1)' Shut down the controller; 0: 166 i : l
- _ _ , _ . _ _ _ . . _ ~ . - - , _ _ _ _ . _ _ . _ . . _ . __._ - _._ ._
O (2) Shut off the N2 gas, and bleed any residual pressure through l I the system relief valve; (3) Clamp the rabbit receiver cup back in the system; and (4) Shut the glove box lid. 9.4.2 central Stringer Irradiations The central stringer is used for irradiations which require a longer irradiation period than is practical with the rabbit system. The irradiaton facility consists of a 4 in x 4 in x 2 in graphite block situated in the center of the core region. This block has forty-nine positions available for 2/5 dras polyethylene sample vials as well as a lifting attachment for removing the block from the core. With the reactor shutdown, the graphite block is removed from the reactor and placed in a glove box. Required samples are then added and the block is returned to the reactor. The reactor is then operated at the required power level for the required time (typical values are 6 hour irradiation time at 100 kW). During this period, other samples may be irradiated in the rabbit system. At the end of the irradiation time, the reactor is shutdown and the samples are left in.the reactor for ten to twelve hours (usually overnight) to allow any short half-life elements to decay to lower levels. With the reactor still shutdown, the , ( block is then removed to the glove box where the samples are removed. i The block is then returned to the reactor. Figure 9-6 is a drawing of the central stringer block. i 9.4.3 Beam Port Irradiations Beas port irradiations are used for experiments which requgre a collimated beam ci' neutrons at a relatively low flux (= 10 n/cm -s).
, Whever a beam port irradiation is done, a beam catcher (shield) is positioned behind the experiment. This beam catcher is capable of reducing gamma and neutron fluxes to levels at or below those allowed by 10CFR20, in accordance with the ALARA philosophy. Areas which are accessible and through which the beam might pass are rcped off and appropriately labeled. To determine the adequacy of the shielding around the experiment, a step ascent to power is performed with radiation surveys-taken until the desired power level is achieved.
Beam port irradiations require review and approval by the facility Director, Reactor Supervisor, Reactor Radiaton Safety Officer, and a Senior Reactor Operator. Irradiations of special nuclear material require the review and approval of the Reactor Safety Committee in each case, as well as those of the above noted individuals. Figure 4-4 shows the available beam port facilities. l l O , i 167
CENTRAL STRINGER IRRADIATION BLOCK O s 0000000 0000000 0000000 0000000 - 0000000 O000000
~
O 0000000 4 = rn e = l !! ii i: :: i: li : l ll13 ll ll !! ll l~ l' ll I il is # ll s lt _ _ll.; t _ ll !! i_ l1l~ _ a t __; __.n ul _.a u__2 l lu l _ _a l l O risure 9-6 ceatret striaser Irreat tio" 3took 1 168
O 10. STEAM AND POWER CONVERSION l Since the VTAL is used solely for training, teaching, and research this chapter is considered not applicable. No steam production occurs and no electricity is generated by the VTAR facility. l O I . h i
.O 169
. - . - .-. .. .. , . - . . . . . . .- , . . , , , , . - - . . , .=
U 11. Radioactive Waste Management i 11.1 Introduction Solid, liquid, and gaseous radioactive effluents are produced by the VTAR. The following sections provide the details on the types of waste generated, the nuclides present, their activities, and how each type of waste is treated at the VTAR facility. 11.2 Solid Waste Management Low-level radioactive solid wastes are generated during routine reactor operations and maintenance. These wastes consist primarily of contaminated gloves, paper, plastics, tools, clothes, samples from activation analysis and ion exchange resins. 1sotopic analysis is performed on unknown solid waste using a germanium-lithium (GeLi) semiconductor detector. Solid wastes are then packaged in 55 gallon drums and shipped to an approved disposal site in accordance with
- applicable NRC and DOT regulations. The total activity of solid waste Table 11-1 Solid Waste Disposal, VTAR (Nov. 1975 - June 1983)
- O Radionuclide Quantity (mci)
Co-60 39.654
! 2n-65 13.592 Ag-110 13.050-
.; Th-232 0.0902
- Cr-51 0.0001 -
I Eu-154 0.0010 Cd-109 0.0042 Cd-113 0.0460 Ba-133 0.0562 Fe-59 0.750 , Ra-226 0.00003 Po-209 0.0000012 Lu-176 0.00006 i Cs-137. 0.1002 Ac-227 0.00003 I Pb-210. 0.000056 Co-57 0.000132
-Na-22 0.0001 4
Cs-134 0.1 Eu-152 -0.050 TOTAL 67.494 4 O 1 1 170-
- -- _ _ _ . . . ... _ __- _. ._ _. _, _ - . _ _ . . -- l
a O d disposed of between November 1975 and June 1983 consisted of 67.45 millicuries (Table 11-1). Cobalt-60, Zine-65, and Silver-110 comprised 98.2% of this total. i 11.3 Liquid Waste Management Liquid wastes generated from reactor operations consist primarily of contaminated water from the shield tank, dump tank, primary coolant, and rinse water from the decontamination of tools. Samples of the water to be disposed of are either analyzed on a Ge-Li detector or evaporated, with the residual solid. analyzed for gross alpha and beta activity. The water is discharged to the sanitary sewerage system if the activity levels are within the acceptable levels for release into an unrestricted area as cited in 10CFR20 Appendix B. The release is further d' luted by an average flow of approximately 900,000 gallons of sanitary sewerage per day. The liquid waste released from the VTAR facility into the sanitary sewerage system from 1977 to 1983 is presented in Table 11-2. Table 11-2 Liquid Waste Released from VTAR to Sanitary Sewerage System 1977 to 1983 Date Volume of Nuclide Activity O Water (liter) (pC1/mi) 3-25-77 9117.4 S-35 1.7x10 4 3-21-78 113.4 6 Ho-166 3-9-79 37.8 Co-60 3.6x10 2.4x10-6 Il 9-5-79 75.6 Unknot.m 9-12-79 37.8 &kaan 5.9x10 II . 9-26-75 75.6 Unknown 2. k10~ 9 1.2x10-10-11-79 75.6 Enknown 8.7x10-10 7 5-2-80 34.0 N2-22 2-25-81 109.6 Unknown 5 x 10'8 3-4-81 94.5 Unknown 1.3x10 1.0x10-8 4-9-81 680.4 Na-24 2.5x10~7 5-21-81 7098.8 Na-24 1.8x10-7 3r-82 5.0x10-8 . 12-1-82 8316 Unknown 2.7x10-8 7-22-83 8316 Unknown ND+ 7-26-83 831.6 Unknown ND
+ non-detectable Liquid wastes that cannot be disposed of in the sanitary sewerage system are placed in suitable containers and stored on site until arrangements for - off-site disposal have been made. The liquid is then 'O. !
171
O V absorbed into vermiculite in 55 gallon drums and shipped to an approved
. disposal site in accordance with applicable NRC and DOT regulations.
Liquid radioactive waste shipped off-site from. 1977 to 1983 consisted of 0.002 m Ci of Co-60 and 0.023 m Ci of Th-232. 11.4 Gaseous Waste Management The primary gaseous effluent of concern is Argon-41. This l radionuclide is produced in the VTAR by the neutron activation of Argon-40 in the air that has leaked into the core region. The maximum ground level concentration of Argon-41 in an unrestricted area was calculated by using the Pasquill-Gifford equation (Appendix 11-1). . Depending on atmgpheric 10~ conditions a range of 1.15 x 10-1 pCi/mi to 2. 2 x pCi/mi was obtained. The maximum permissible ground level concentration for Argon-41 in an nrestricted area as cited in 10CFR20, Appendix B is 4.8 x 10-8 pCi/mi. ANSI recommends that the concentration averaged over a year should be less than 2% of this limit. The calculated concentrations of Ar-41 in unrestriced areas around the VTAR represents 0.24% and 0.46% of 10CFR20 limits. Arggn-41 emissions are limited by VTAR Technical Specifications to 1 x 10~ Ci/see and 315 Ci/ year. Table 11-3 shows the Ar-41 stack discharge rate in Ci/see from 1978 to 1983. Table 11-4 shows the total
, amount of Ar-41 released per year for the period 1978 to 1982.
1 Table 11-3 Ar-41 Stack Discharge Rate + (C1/acc), VTAR 1978 to 1983 Date Discharge Rate , (Ci/sec) 10-25-78 5 t 4-21-79 3.00x10 3.28x10-5 10-23-79 4.86x10-5 11-29-80 4.77x10-5 l 11-3-80. 1.63x10-5 5-15-81 5.48x10-5 8-9-82 5 1-26-83 4.79x10 3.29x10-5 . + flow rate of 1x106 ,17,,, t 172
p& Table 11-4 Ar-41 Stack Discharge Yearly Totals, VTAR 1978 - 1982 Year Ar-41 (Ci/yr) kWh 1978 144.7 90,174 1979 169.9 105,846 1980 113 70,278 1981 73.6 45,845 1982 7.4 4,636 i 11.5 Conclusions Solid wastes produced at the VTAR consist of low-level radioactive wastes. These wastes are packaged and shipped to an approved disposal site in accordance with applicable NRC and DOT requirements. Liquid wastes are discharged to the sanitary sewerage system if the activities are less than those specified in 10CFR20, Appendix II.B for unrestricted areas. Liquid wastes not discharged to the sanitary sewerage system are packaged and shipped to an approved disposal site in accordance with applicable NRC and D7f requirements. O environment consist mainly of Ar-41. Gaseous effluents released to the Releases to unrestricted areas are well below the maximum allowable limits cited in 10CFR20, Appendix B. Release rates in Ci/see and C1/ year are less than the limits specified in the Technical Specifications for the VTAR. Baeed on the above analysis the radioactive waste management program at the VTAR faculty f s operating in a safe manner consistent with HRC and DOT regulations. O 173
m
-O Appendix 11-I Concentration of Argon-41 in Unrestricted Areas The maximum ground level concentration of Argon-41' can be calculated from the Pasquill-Gifford equation i + )) (Eq. 1) x(x,y) = ,*0y*z,exp(- ay f( az .
_ 3 where: x = ground level concentration of the effluent in pCi/m at point x,y x = downwind distance on plume centerline, meters t y = crosswind distance, meters Q = emission rate in pCi/see a = horizontal standard deviation, meters i y a, = vertical standard deviation, meters j u = average wind speed, meters /sec H = effective stack height, meters -
; Calculation of emission rate (Q)
' Air from the coge region is drawn into the ventilation stack at a
- flow rate of 1x 10 ai/sec and is discharged on the roof of Robeson Hall. A ~ booster fan on the roof mixes ougide air with the stack air.
i resulting in a total tiow rate of 6 x ,10 mi/sec leaving the booster j fan. The maximum equilibrium rate measursd leaviag the stack and
- estering the fan is 55 pCi/sec. Assuming ideal mixing reaults in a dilution to
6 (55 pCi/sec)(
- 6 '" ) = 9.17 pCL/sec 6 = 10 ni/see The reactor operates . a maximum of 8 hours a day, S' days a week. ,
Averaging this value over a period of a year results in a release rate
- of
- I l , (9.17 pCi/sec)(8 0
) = 2.18 pCi/see Calculating effective stack height (H) i The effective stack . height is a function of the actual stack height, the exit velocity and the temperature of the gas (ref. 11.2)..
0 174 't
, , . - ,,e., , . , _ _ , . , . , , . , , , _ y,,, ,y-%, ,_y m#. - , ,e.,
< I
- n b (Eq. 2)
H . = h + d() * (1 + f) where: h' = actual height of stack, meters. d~ = stack outlet diameter, meters
= exit velocity of gas, meters /see v
p = average wind speed, meters /sec ST = dif ference between ambient and gas temperature, OC T = absolute temperature of gas, OK
. Substituting the following values into Eq. 2:
h = 16.76 m u = 3.76 m/sec (ref. 11.3) O d = 1.07 m AT =OC v = 11.7 m/see T = 296 K f yields: 0 H H. a sec 0
= 16.76 m + 1.07 m (3.76 m/see)1.4(g , )
' 296 H = 22 meters Calculation of the downward distance (x) which yields the maximum ground 1 level concentration.
The maximum ground level concentration occurs on the plume centerline at the downward distance where the vertical standard deviation-is: (ref. 11.2) y = h W 4 Therefore: 22 me m o* =
, [
J, = 15.6 m l (From Fig. 11.8 of Ref. 11.2) A vertical standerd deviatio 1 of 15.6 m l corresponds to a dowawind distance of approximately 90 m for cxtremely I unstable atmospheric conditions (Pasquill category A) and approximately 1000 m for ' moderately stable atmospheric conditions (Pasquill category F). The horizontal standard deviation (o ) associated with these values ! can be found from fig. 11.7 of Ref. . II.D. These values are 20 m and l 38 m respectively. I f l O l L - -175
Calculation of the maximum ground level concentration (x)
~
Substituting the following values into Eq. 1: 1 xg = 90 m (Pasquill Category A) x2 = 1000 m (Pasquill Category F) y =0 Q = 2.18 pCi/sec , o yg
= 20 m c
y2
= 38 m
! o x1
= 15.6 m o
2
= 15.6 m i p = 3.76 m/see H = 22 m yields: Pasquill Category A 1 22m 2.18vci/see -bl5.6m)2 x = (90m, Om) = (w)(20m)(15.6m)(3.'76 m/sec) ,
-10 gj 3 x = 2.2 x 10 2 .
Pasquill Category F 1 22m 2 [- bl5.6m)23 E = (1000m:0m) = }wy.18$1/see 8mj(15.6mj(3.76 m/sec) , x = 1.15 x 10 -10 C1/cm This calcalation shows that the maximum ground level concentration of Ar-41 is well below the MPC of 4.8 x 10-g pCi/mi for an unrestricted 3 area as cited in 10CFR20, Appendix II.B. I i 176
)
, . . .-- - . ~ . - - - - . _ _ _ .. .- . .- -. ...
i .. I 3 12. Radiation Protection l 12.1 Health Physics Program 12.1.1 Responsibilities It is the responsibility of the Reactor Safety Committee and the Reactor Radiation Safety Officer to ensure that radiation exposures to workers and the public, and radioactive releases to the environment are as low as reasonably achievable. The Reactor Safety Committee is responsible for regulating the safe operation of the reactor facility and reports directly. to the Vice-President for Administration and Operations. 4 The Reactor Radiation Safety Officer assumes the primary l responsibility for assuring that those regulations adopted by the f Reactor Safety Committee pertaining to the Health Physics aspects of the reactor and its operations are carried out and maintained. Figure 12-1 shows the organizational chart for the Reactor Safety Committee and the Reactor Radiation Safety Officer. 12.1.2 Duties of the Reactor Radiation Safety Officer ! The Reactor Radiation Safety Officer it, a staff member of the i Office of Safety and Health Programs and as such is independent of the
- O cc c 11- '"i a . 1 < tici a 67 re ing, and experience to advise others in the safe use of ionizing radi-er a c ti - tr i -
t ation and to supervise the Reactor Health Physics program. The Reactor l Radiation Safety Officer will perform the following duties: (1) Act in a supervisory capacity in all aspects of the Reactor's { radiation measurement and protection activities. This includes
- responsibility for personnel monitoring, maintenance of I exposure records, . survey methods, waste disposal, and radiologie31 safety practices as specified by NRC regulations, 4
The Commonwealth of Virginia, or as approved by the Reactor : Safety Committee '
, (2) Ensure that radiation doses to wekers and - the public, and I radioactive releases to the environment are as low as reasonably achievable (3) Provide written approval for all activities and proceduras which involve actual or potential exposure of personnel to radiation or the release - of radioactive waterials to the environment. The Reactor Radiation Safety Officer shall bring '
- those activities not covered by established procedures to the Reactor Safety Committee 4
- O 177
(j President VPISSU Vice-President for Administration a operations
~
Campus + Police l i<adiation Safety Director Committee S.H.P. ! O Reactor Radioisotope a Committee i l l l t I [ l Radiation __ _ .L L _ Satety. Office l I Figure 12-1. Administrative Organizational Chart for the Radiation Safety Program at - Virginia Tech O 178
If, '
,,M'A i g Y, .
ll (d' ' 1
, : <;
- l /
T $. ( O g' ;
- q. > i
- Q jr ,
'A provide advice 1(4) Consult with / reactor personnel and on
,I f s radiological safety' practices-L6
. (5) Suspend any operation as rapidly and safely as possible that is f8 causing or da:eaed capable of causing an excessive radiation
'i hasard., The Raaetor ' Safety Of ficer shall notify the Reactor i
l
Safety;Committeewhichwillpromptlyreviewtheincident (6)Y Perform routine and special radiation surveys and inspections
,' as' deemed necessary in the interest of radiation safety 1 ~l (7) Prepare's quarterly ' report of. incidents, and materials received L ,q <
and transferred .J . I d (8) Maintain a current list of authorized users of the reactor facility ' (9) Ensure that users of the reactor facility are adequately trained in the haastd4 associated with radiation, and the methods to rapuce eEp'sures o to themselves and others
- s. Ii (10) Sarve as an ex t.'fficio. aember "
or. the Reactor Safety Committee. i 12.1.3 Routine Inspections and Surveys VO j'", . At least annucily, the' heactor 'Aadiation Safety Of ficer conducts a formal inspection of the reactor facility.
~
This inspection includes a
! & review of' the general conditions of the facility, survey logs, radio-l) i active material inventory, and adhere'ce principles.
n to ALARA ' radiation safety t t On a quarterly basis th[Reacter Radiation Safety Officer conducts
, tMe followiag contine radistien, surveys. '
1 , f . [ (1) area gamma and neutron radiation levels in restricted and j' > enrestricted areas j 4-(2) contamination' levels in res.t:Tieted and unrestricted areas i > 4 (3) sealed source leak tests i : j (4) Argon-41 sty samples in restricted ar.d unrestricted areas
- g. (5) Primary coblaii s'amples ,
13 , i 12.1 4 Radiarton Safety Training b % .
- n. ,
A _All personne17 desiring to work;in restricted areas at the VTAR aust h ' ,4 attend a video tape radiaton safety lecture series which includes the j' , 9o11owing informatton: '
<Li <
2 L ' M' k, l,, 4f g W: (1) general properties of ionizing radiation + 1, q j; P Me
.179 J\,
' ' f).
. u i <.: - -
I (2) principles of radiation dete: tion (3) radiation units and limits (4) biological effects (5) methods to reduce radiation exposure (6) federal, state and local regulations At the conclusion of the lecture series all individuals are required to pass a written examination administered by the Radiation Safety Office. 12.1.5 Conclusion The organization and function of the Health Physics program at Virginia Tech is adequate to ensure that workers, the public, and the environment are not exposed to excessive radiaton hazards from the operation of the VTAR and that exposures are maintained at a level that is ALARA. 12.2 Radiation Safety Instruments Radiation safety equipment available to. the VTAR is located at the O reactor facility, the radiation safety office, and in the radiation emergency kits. This equipment consists of portable survey instruments, stationary counting systems, and installed radiation monitors. 12.2.1 Portable Survey Equipment and Stationary Counting Systems The following portable survey equipment and stationary counting systems are available for monitoring radiation levels, contamination i levels, and radioactive effluent releases from the reactor facility: (1) Reactor Facility 2 - neutron monitors 2 - beta gamma ion chambers 1 - PAC-4G alpha monitor 1 - high volume air sampler 2 - Frisker contamination monitors 1 - beta gamma Q4 survey meter 2 - GeLi detector and multi-channel analyzer (2) Radiation Safety Office high range beta gamma ion chamber 1 - beta gamma Gi survey meter
.180
, -- r
O V 1 - alpha, beta gamma G1 survey meter 1 gas flow proportional counting system 1 1 - liquid scintillation counting system 1 - NaI(T1) scintillation counting system 1 - GeLi detector and malti-channel analyzer 1 - TLD reader 1 - neutron monitor 1 - Triton radioactive gas sampler 1 - Frisker contamination monitor (3) Radiation Emergency Kits 1 - high volume air sampler 1 - PAC-4G alpha monitor 1 - low range Qi survey meter 1 - high range ion chamber 12.2.2. Installed Radiation Monitor The installed radiation monitoring system provides a continuous indication of gamma radiation levels throughout the reactor facility. The readings from these monitors are displayed on the reactor control console. An audible alarm is triggered if any monitor exceeds a predetermined set point. n v Area radiation levels in the reactor room are monitored by two radiation monitors mounted on the east and west walls. Normal readings for the cast wall monitor are 1-2 mrem /hr. The alarm is set at 5 mrem /hr. Normal readings for the west wall monitor are 2-4 mrem /hr with the alarm set at 15 mrem /hr. A monitor is located in the stack near the exhaust end to monitor radioactive gas releases. This monitor normally reads 0.3 mrem /hr, with the alarm set at 15 mrem /hr. A radiation monitor that continuously examines the primary coolant for fission product activity is installed in a small holdup tank adjacent to the process pit. This monitor normally reads approximately 100 mrem /hr. The alarm set point is 250 mrem /hr. In addition, an air particulate monitor, consisting of a NaI(Tl) scintillation detector connected to a single channel analyzer, is located in the base of the stack to measure unplanned releases of airborne fission products. 12.2.3 Instrument Calibration and Quality Control Portable radiation survey instruments and fixed radiation monitors are calibrated by the radiation safety officer semiannually.
-181
. . -- ~~ . . - -
4 [_ 4 (~ - High range survey instruments are sent to an outside vendor for calibration. i Quarterly quality control tests consist of a calculation of chi-square operating voltages, and efficiencies for the counting systems in the radiation safety office. The GeLi and NaI(Tl) counting systems are calibrated prior to use. Control charts are currently being established to determine the reproducibility of the counting systems. TLD chips are exposed to known radiation fields and a standard curve is generated for the TLD reader prior to using it to count unknu n chips. 12.2.4 Conclusion I The radiation safety instruments available to the reactor facility are adequate to protect the general public and operating personnel from excessive radiation hazards and unplanned releases of radioactive gaseous and particulate matter. Calibration procedures and quality control tests are adequate to ensure that radiation survey instruments and counting systems are operating properly. ; , 12.3 Personnel Monitoring O 12.3.1 Occupational , Individuals working in restricted areas at the reactor facility wear whole body TLD's. Depending on the nature of the work, ring badges may also be worn. These badges are supplied by an outside vendor and are processed quarterly. Table 12-1 presents the whole body gamma doses received by workers at the VTAR from 1977 to 1982. For this period the
~
highest yearly individual whole body gamma ' dose was 0.788 rem. This dose represents 15.8% of the maximum allowable dose. The highest cummulative dose was 3.338 person-rem. If higher than normal doses are expected (e.g. during fuel transfer) or during non-routine operations', in-house . TLD's and self-l reading pocket - dosimeters may also be worn to provide an - immediate reading of the dose received. Neutron -badges are issued only to those individuals who are
- . expected to receive the highest neutron dose among the reaccor staff.
In order to evaluate possible internal ingestion of radionuclides at the reactor f acility, urine samples are collected yearly from select individuals and analyzed for gross' beta activity. Rasults from urine samples collected f rom 1977 to 1982 have been less than the action level defined by the Reactor Safety Committee of three times the MDA of the anlaysis procedure.
.182
i f 10 \ Table 12-1 Occupational Whole Body Gamma Doses, VTAR, 1977 to 1982 Total Occupational Highest individual 10iH's of Year Dose (person-rem) Dose (rem) Operation 1977 3.338 0.788 113,837 1978 2.493 0.456 90,174 1979 2.319 0.465 105,846 1980 1.203 0.266 70,278 1981 1.072 0.205 45,845 1982 0.540 0.090 4,636 12.3.2 Non-Occupational i Visitors entering restricted areas as defined by 10CFR20 are issued self-reading pocket dosimeters. When groups are touring the facility, pocket dosimeters are only issued to representative individuals. Table 12-2 shows the non-occupational doses received at the reactor facility 1 for the period 1977 to 1982. The highest non occupational dose for this
; period was 0.030 rem and the_ highest yearly cummulative dose was 0.142 l
person-rem. Table 12-2 Non-Occupational Whole Body Gamma Doses, l VTAR, 1977-1982 Total Non-Occupational Highest Individual J Year Dose (person-rem) Dose (rem) 1977 0.077 0.012 1978 0.108 0.014 1979 .0.142 0.016 1980 0.017 0.005 1981 0.067 0.030 1982 0.016 0.005 l I f 183
12.3.3 Conclusion Occupational and non-occupational doses resulting from operation of the reactor have been very low and are well below the maximum permissible doses cited . in 10CFR20. These low doses indicate that the VTAR is operated in a manner that is consistent with achieving doses that are ALARA.
- 12.4 Radiation Surveys-Restricted dreas 12.4.1 Shielding The reactor is shielded on the sides and top with 6 1/2 feet of ordinary concrete. A lead gamma curtain 2 inches thick is located between the core and the thermal column. Another 2 inches of lead are located on the side of the shield tank opposite the core.
Fourteen poured concrete blocks augment the shielding around the perimeter of the reactor and two custom built high-density concrete blocks with integral shielding augment the top shielding. Solid stacked concrete blocks provide additional shielding where necessary. 12.4.2 Radiation I4vels Neutron and gamma radiation levels at the reactor are measured m quarterly by the Reactor Radiation Safety Officer. Table 12-3 contains measured gamma and neutron radiation levels from 1977 to 1982. Survey points correspond to the survey locations shown in Figure 12-2. Data were taken with all shielding in place and the reactor operating at 100 kW. Figures 12-3, 12-4, and 12-5 present isodose curves measured af ter the installation ef the new shielding in October 1978. The average gamma dose rate around the bottom of the reactor (points 1-12 figure 12-
- 2) from 1979 to 1982 was 1.3 mrem /hr, and the average neutron dose rate was 0.7 mrem /hr. Radiation levels at the top of the reactor averaged 9.7 mrem /hr for gamma radiation and 3.0 mrem /hr for neutron radiation.
Reductions in the radiation levels at several points occurred in 1979, reflecting the effectiveness of the new shielding. The isodose line passing ' through generally accessible areas at the bottom of the reactor shows 2.5 mrem /hr for gamma, <1.0 mrem /hr for fast neutrons, and <0.1 meem/hr for thermal neutrons. The isodose line at the top of the reactor shows 15 mrem /hr for gammas, 2.0 mrem /hr for fast neutrons, and 0.2 mrem /hr for there.a1 neutrons.
.12.4.3 Contamination Surveys A loose surf ace contamination survey of the work area in the NAA laboratory is conducted daily when samples are irradiated. A similar survey of the entire laboratory and connecting unrestricted areas is conducted weekly. These surveys are performed and documented by the reactor staff.
O ' 184
O Table 12-3 Radiation Levels + VTAR. 100 kWt. 1977-1982
, Gamma (area /hr) Thermal & Fast Neutron (area /hr)
Points 1977 1978 1979 1980 1981 1982 1977 1978 1979 1980 1981 1982
- 1. 0.9 0.6 0.3 0.4 0.3 0.3 2.0 2.0 1.3 0.9 0.8 0.3
- 2. 1.7 0.9 0. 4 0. 4 0.4 0.6 0.9 1.8 1.3 0. 9 1.3 0.3 *
- 3. 2.5 1.7 2.1 2.5 1.9 1.6 0.6 1.5 1.6 0.6 0.7 0. 3
- 4. 2.8 1.8 2. 4 2. 0 0. 9 0. 9 0. 7 0. 9 1.8 0.5 0.8 0.1
- 5. 1.0 3.3 0.8 0.9 1.2 0.4 0.8 21 2.4 0.2 0.1 0.1
- 6. 3.5 5. 2 2. 5 1.6 2. 3 0. 7 1.6 2. 2 1.9 0. 5 0. 8 0.4
- 7. 2.0 3.7 1.8 8. 9 7. 3 2. 9 1.2 1. 9 1.5 0. 9 1.2 0.2
- 8. 2. 7 5. 0 1.5 1.1 1. 4 1. 7 0.7 1.4 1. 0 9.
- 0. 4 0. 5 0.1 1.2 0.7 0.2 0.2 0.2 0. 2 0.5 1.3 0.3 0.3 0.2 0.1
- 10. 1.3 3.1 0. 2 0. 2 0. 2 0. 2 0. 8 1.7 0.4 0.5
- 0. 3 0.1
- 11. 0.6 0.3 0.2 01 0.2 0.2 0.8 1.4 0.6 0.4 0.5 0.1
- 12. 3. 2 3. 8 0.8 1.3 1.3 0.7 2.1 2. 5 0. 9 0.7 0.8 13.
0.2 18.7 22.3 13.5 16 17 14.1 4.8 5.0 3.7 2.4 2.2 0.8 14, 46.7 32.5 30 35.7 42.3 32.8 31 18.9 15.0 14.9 26 4. 5
- 15. 3.3 3.5 2.8 -- --
1.6 0.7 1.7 1.5 -- -- 0.3 () 16. 17. 18. 7.0 11.7 11.7 10.2
- 8. 5
- 7. 3 6.2 12.3 12.7 5.1 12 4.7 12.7 12.3 3.8 7.3
- 1. 2 3. 8 13.4 15.3
- 2. 8 3.8 1.4
- 2. 5 2.1 5.4
- 0. 3 0.7 12.~6 6. 8 10.1 13.1 4.4 2.5 4.7 1.1
- 19. 13.5 7. 7 10.2 10.7 11.2 5.3 14.2 15.1 5.4 5.8 7.3 1.3 20'. 6.0 8. 0 6. 4 6. 0 5. 6 5. 0 2.1 9. 4 4. 2 2. 4 1. 7 0.7
- 21. 11.0 17.3 13.4 17.5 18.0 14.7 5.2 5.0 3.8 3.0 3.9 1.1
- 22. 17.5 6.2 4. 7 1.5 1.2 10.2 26.3 23.
- 7. 6 5. 2 4. 0 5.3 5. 9 10.0 8.5 8.8 9.2 10.8 4.2 1.0 1.8 1.6 1.0 1.2 0.4
+ Averaged of RRSO Quarterly Surveys
- See Fig. 12-2.
1 185
l WEST W FUEL O HOT - - - - - - - - S ORAGE [ g \' g h {"):33 is e :: -:
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. ~ . . .__
i ' ( Wock areas in the reactor room and tools are swiped for potential contamination'at the end of a work session by the reactor staff or the
' radiation safety . officer, if present. In order to ' ensure that contamination is not spread to unrestricted areas, personnel leaving the reactor ' room are required to. monitor themselves with a Frisker i contamination monitor prior to entering unrestricted areas.
- Decontamination is performed if the contamination levels exceed the limits presented in Table 12-4.
) Quarterly, the Reactor Radiation Safety Officer conducts a contamination survey of the reactor f acility and reviews contamination j surveys performed by the reactor staff. 1 Table 12-4' i 4 Contamination limits, restricted areas, VTAR i
- (a) General Areas. Equipesnt. Tools Type Contamination Gross 6y Alpha Loose 1000 dpa/100 cm 2' Non Detectable Fixed 1.0 ar/hr 9 1/16" Non Detectable l
f (b) Control Contamination Areas
- Type Contamination Gross 8 y Alpha i
10,000 dpa/100 cm 2 50 dpa/100 cm 2 Loose
; Fixed 1.0 ar/hr @ 1/16" 500. epa (PAC-4G)
{ ] (c) Skin and Personal Clothing l Type Contamination Gross 6 y Alpha i ); Loose Non Detectable Non Detectable Fixed 0.05 ar/hr Non Detectable f l
- Higher levels shall only be authorised by the Radiation Safety Office on an individual case basis.
i 12.4.4 Sealed Sources Sealed sources used at the reactor facility are leak tested by the RRSO on a quarterly or semi-annual basis, depending on the type of source. Contamination swipes from 1977 to 1982 have revealed no leakage i in excess of the NRC IIsit of 0.005 pCi/ swipe. 4 Sealed neutron sources are presently stored in a locked parafin-filled barrel surrounded by stacked concrete blocks. Radiation levels i O 190 _.u. . - m. ,. ~ , , , _ , -
, . . , - - - - - ,y - - , ,
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^on the sides of thesa blocks on contact are 2.2 mrem /hr for gamma i radiation and 0.2 meen/he for neutrons. Other miscellaneous sources are j stored in the reactor storage cabinet. Radiation levels at this cabinet are 1.5 ares /hr.
4 4 When the sealed sources are being used, exposure rates are l monitored with the appropriate survey meter. Sources are always handled I j
. .th tongs. All operations are conducted under the direct supervision ,
, of an authorized user. ' 12.4.5 Primary Coolant Analysis l s A sample of the primary coolant is analyzed for radionuclide content on a GeLi detector and multi-channel analyzer by the RRSO on a quarterly basis. Since this program has only recently been introduced, data are not available for an analysis of this procedure. ] 12.4.6 Argon-41 Sampling i Argon-41 is produced during . routine reactor operations by the f neutron activation of air that has leaked into the reactor core area. l Air from the core area and the reactor room is discharged through a duct j to the roof of Robeson Hall by an exhaust fan on the roof. a i On a quarterly basis the RRSO takes an air sample from the reactor . room and analyzes it for Argon-41 concentrations. Surveys from 1978 to : 1982 have shown no detectable Argon-41 levels in this area. , 12.4.7 Conclusion The data presented in Table 12-3 indicate that the radiation levels 4 around the VTAR during full power operations are generally very low. I Isodose curves measured 'af ter the installation of the new shielding again reflect the low levels of radiation associated with reactor ope rations. These low radiation levels correlate witn the low i occupational doses received by personnel (Table 12-1) and indicate that the VTAR is operated in a manner that is consistent with achieving doses j that are ALARA. l The contamination survey program in restricted areas is adequate to
- prevent the spread of radioactive materials to unrestricted areas. In j support of this conclusion, contamination surveys conducted in un-l restricted areas from 1977 to 1982 have shown no levels of contamination i
in excess of the limits ' adopted by the Reactor Safety Committee. i ) Sealed sources are stored and handled in a safe manner consistent with sound radiological safety practices and pose no danger to reactor
- personnel or the public.
I e O l 191 i l l
. . - .. ._ - _ . ~. . . - . . - -- .. . - - .
i. U Primary coolant analysis is expected to increase the safety margin of the reactor by identifying radionuclides in..the coolant. Argon-41 produced during routine reactor operations is effectively' removed by the exhaust fan and discharged through the exhaust stack at the roof and does not pose an undue health hazard to workers. 12.5 Radiation Surveys-Unrestricted Areas I Radiation. levels in unrestricted areas surrounding the reactor facility are monitored by environmental TLD's, contamination surveys, ! area radiation surveys, and air samples. 12.5.1 Shielding j The reactor is located in the northwest corner of Robeson _ Hall. [ Classrooms and laboratories of the Physics Department are located in
- . this building; however, only one classroom borders the reactor, on the
! top. Robeson Hall was designed specifically to accomodate the l reactor. . Inside walls surrounding the reactor are composed of 12 inch } thick re-enforced high density concrete. The ceiling directly above the i reactor is made of 24-inch thick re-enforced high . density concrete. Outside walls are composed of concrete 24 inches thick, f aced . with limestone. O 12.5.2 Environmental TLD's i ! Environmental levels of external gamma radiation are measured by TLD 's. . These TLD's are placed at the locations presented in Figure 12-l 6. Table 12-5 shows the results of the TLD environmental monitoring { program from 1977 to'1982.
- Radiation levels in areas 1 through 3 exceeded 500 ares /yr for 1977 i and 1978. Additional shielding was added to the reactor in October of i 1978 reducing these levels to below 500 mrea for 1979. Radiation levels l at the outside oak door (area 4) however. continued ~ to exceed 500 arem/ year. In 1983 additional shielding was erected around a RaBe source that was stored in close proximity to the outside door. In addition, sources that were stored in a cabinet near these doors were relocated further from the door. Since these. actions were taken
) radiation levels at the _ door have been minimal, suggesting that the observed radiation levels were from the stored sealed sources rather i than from reactor operations. ! Radiation levels from Argon-41 emissions are monitored by a TLD in i the stack where it discharges at the roof (point 6). For the period
- i. 1977 to 1982 radiation levels ranged from 30 to 256 aren/ year, depending '
on the hours of reactor operation. O 192
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>N First Floor NAA Offices Reactor i Room, h t'5)
- 1. Reactor door to hall 6. Stack at roof
- 2. N.E. corner, hall 7. Nitrogen holding area
- 3. Basement stair well 8. Rm. under rabbit on ceiling
. 4. Outside door "
- 9. S.E. wall, Rm. 106
- 5. Rm. 208 above reactor i
j Figure 12-6. Location of Environmental TLD's, VTAR. 193
P l-L L j O 12.3.3 Area Radiation Levels
- Aren gamma and neutron radiation surveys, to determine exposure l' rates, are performed on a quarterly basis by the RRSO. Radiation levels are determined at waist level directly against the outside walls or doors. Figure 12-7 shows the points where the surveys are conducted in unrestricted areas and Table 12-6 presents the results of these surveys from 1977 to 1982. Gamma radiation levels from 1979 (after the installation of the new shielding) have averaged 0.2 aren/hr, and neutron levels have averaged 0.1 meen/hr in -unrectricted areas.
Decreases in radiation levels in area 1 and 4 were noted after the addition of the new shielding in October of 1978. 12.5.4 Contamination Swipes Contamination swipes are performed by the RRSO on a quarterly basis . in unrestricted areas surrounding the VTAR. Contamination swipes from 1977 to 1982 have revealed no removable contamination in excess of the limits approved by the Reactor Safety Committee for unrestricted areas (Table 12-7). 12.5.5 Argon-41 Sampling Air samples from unrestricted areas are analysed for Argon-41 on a quarterly basis by the RRSO. Results f rom 1977 to 1982 have shown no O detectable Argon-41 levels in these areas. Air samples from the exhaust stack are also analysed for Argon-41 emissions na quarterly basis. Results from these samples are shown in Table 11-3, chapter 11. Argon-41 concentrations for the study period 1978 to 1982 have been consistently below the maximum allowable release rate into- an unrestricted area. The ground level concentration of Argon-41 was calculated for the maximum measured release rate using the Pasquill-Gifford Equation (Appendix 11-1). Dep ing on atmospheric e concentrations - ranged from 1.15 x 10~p$1/gi to 2.2Wi/mi x 10~gditions, which is less than the MPC of 4.8 x 10~ pCi/mi cited in 10CFR20, Appendix B, for releases of Argon-41 into an unrestricted area. 11.5.6 Conclusion Although radiation levels have exceeded 500 meen/ year in unrestricted areas around the reactor, these areas are not occupied constantly. Areas 1 through 4 are in hallways, stairwells and outside doors; not offices, classrooms, or laboratories. It'is therefore deemed unlikely that any one individual could have received a dose in excess of 500 meen/ year in an unrestricted area. Through the use of additional shielding and the relocation of radioactive sources, radiation levels have been reduced and are sufficiently low to ensure that members of the public will not receive a dose in excess of 500 ares /yr. O 194
Ground Floor i O Reactor Room i l
>N a
l First Floor 1 i NAA Offices Reactor + Room. i O @ .- w, i @ i ! 1. Door to hall i 2. Ground floor hall i 3. Stairvell .
'4. Outside door
- 5. Conference Rm.
i 6..First floor hall f
- 7. Rm. above reactor i ;
i ! Figure 12-7. Survey Points, Unrestricted 4 Areas, VTAR. i
- O j 195 i
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~
Table 12-5 Environmental TLD Results, VTAR, 1977-1982 Exposure (area) Location 1977 1978 1979 1980 1981 1982 1
- 1. Rx door hall 1479 981 497 394 252 30
- 2. NE corner hall 903 516 155 84 65 30
- 3. Basement stair wall 1444 935 360 212 163 120
- 4. Outside door 1472 677 540 808 431 550
- 5. Rm 208 above Rx 105 100 97 112 93 0
- 6. Stack at roof 256 107 124 104 69 30 0 7. N 2 holding area 173 130 216 111 73 0
- 8. Rm under rabbit 151 144 111 141 85 0
- 9. S.E. wall Room 106 247 356 247 205 96 0 kW-hr 113837 90174 105846 70278 45845 4636 Radiation levels from Argon-41 emissions have been consistently low from 1977 to 1982 and do not present a hazard to the general public.
Neutron and gamma radiation levels in unrestricted areas around the reactor facility have remained consistently low (< 0.3 meen/hr, gamma and neutron) and pose no danger to members of the public. Contamination swipes have revealed no levels in excess of allowable limits for unrestricted areas. This indicates that contaminated materials are handled in a safe manner in the reactor facility and that
. monitoring and decontamination procedures are adquate to prevent the l spread of radioactive materials to unrestricted areas.
196
TABLE 12-6 Radiation Levels *, Unrestricted Area VTAR, 10 0157, 1977-1982 Survey 1977 1978 1979 1980 1981 1982 Point + y n y a y n y n y n y n 1 0.6 0.5 0.5 0.7 0.1 0.2 0.1 0.4 0.1 0.4 0.2 0.1 2 0.1 <0.1 0.1 0.1 0.1 0.1 0.1 <0.1 0.1 <0.1 0.1 0.1 3 0.2 0.1 0.1 0.1 0.2 0.1 0.2 0.1 0.3 0.1 0.1 <0.1 4 0. 3 0. 3 , 0.4 0. 2 0. 4 0.1 0. 2 0.1 0.1 0.1 0. 3 0. 2 5 0.1 - 0.1 0.9 0.1 0.1 0.1 0.1 0.1 0.1 0.1 0.1 6 0.3 -
- 0. 3 1.3** 0.3 0. 2 0. 3 0.1 0.3 0.1 0. 2 0.1 t
7 <0.1 - - - - - 0.2 0.1 0.2 0.1 0.2 0.1-
- average quarterly R30 surveys in area /hr
( ** based on one survey
+ See Fig. 12-7 TABLE 12-7 Contamination Limits, Unrestricted Areas. VTAR Type Contamination Gross 8y Alpha Loose 220 dpa/100 cm2 Non-detectable Fixed 0.05 ar/hr Non-detectable l
l 1 l l 197
, , , - _ ______.__,_...r_ -v-., .,, , .,..e __ _,~.,,c_,_.. .
v, . Argon-41 levels in unrestricted areas have been undetectable and emissions f rom the stack have consistently been less than the limits
- specified in the VTAR Technical Specifications. Ground level concentrations based on calculations have been less than the maximum permissible concentrations cited in 10 CFR20. These data indicate that Argon-41 emissions from routine reactor operations do not pose an undue health risk to the public.
12.6 General Conclusions Based on the above analysis the radiation protection program at the VTAR is adequate to protect individuals working at the facility, the public, and the environment from excessive radiation hazards. The low occupational doses and-low area radiation levels associated with eactor operations indicate that the reactor is operated in a manner that is consistent with achieving doses that are ALARA. l 1 O f i i r O l 198 l l
. . . .- - ~ . _ _ . . .- - - - - - -_ _ __-_ _ _ _ _
I
- 13. CONDUCT OF OPERATIONS i
13.1 Organizational Structure of Applicant 1 13.1.1 Management and Technical Support Organization l ( The VTAR is owned and operated by Virginia Polytechnic Institute i and State University (VPI&SU), for the purpose of research and i instruction. The President of the University, Vice President for Administration and Operations, Dean of the College of Engineering, Head of Mechanical Engineering Department, Director of the Nuclear Reactor Laboratory, and the Reactor Supervisor all have responsibilities in line for the control of the reactor laboratory, for safeguarding the public i and adhering to all requirements of the facilities licenses and the j technical specifications. The authority control chain of the VTAR is shown.in Figure 13-1. i VPI&SU is under the direct supervision of its Board of Visitors. l All university affairs are directed by the President with council of the The Mechanical Engineer-University Provost and other Vice Presidents. ing Department is part of the College of Engineering and is under the supervision of the Dean of the College of Engineering. , j 13.1.2 Operating Organization i i 13.1.2.1 Director of the Nuclear Reactor Laboratory and Reactor i j Supervisor i j The Director of the Nuclear Reactor Laboratory (NRL) and Reactor , Supervisor are- in charge of the daily operation of the reactor i laboratory. They are responsible for the safe operation of the reactor,
- the physical protection of the laboratory, the scheduling and super-j vision of expensents using the reactor, the control of reactor fuel,
- the keeping of administrative logs and records, and the maintenance and
'. physical condition of the reactor and laboratory. They are also responsible for coordinating the teaching and research programs within , the facility as well as being the liaison with the NRC and other regu-i latory bodies. 1
- In all matters pertaining to reactor operation and technical speci-
! fications, the Reactor Supervisor and the NRL Director shall be respons-1' ible 'to the VTAR Reactor Safety Committee. For administrative matters the NRL Director shall be responsible to the Department Head of Mechanical Engineering. - The Reactor Supervisor shall be a licensed Senior Operator and l should have the following minimum qualifications (ANSI /ANS 15.4) for a level two supervisor. At the time of appointment the Reactor Supervisor 199 l
- ~
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President VPI&SU 1 Vice-President for 4 Administration & Operations i i i Camous Provost pogge, l Dean College of ! Engineering 4 Department Head Radiation Safety Director Mechanical Committee
~~~
S.H.P. Engineering i Director Reactor Nuclear Reactor Safety Radioisotope ]' . . CMuee Laboratory , committee
! (NRL)
- l i
.I l i I I I Reactor I U Audit i Supervisor 1-- L. - - L - - - Safety
- Gr up N.R.L. oggie,
! l Operating Admin. Personnel Personnel 1 Operating l Trainees l VTAR Organizational Structure Figure 13-1 i 200
1 shall have a minimum of six years nuclear experience. The individual i shall have a bachelor's degree or higher in an engineering or science field. Equivalent education or experience may be substituted for a degree. The degree will fulfill four years of the six years of nuclear experience required on a one-for one time basis. The Reactor Supervisor viects operators and technicians, supervises their training, med enforces operating procedures and i regulations. The Reactor Supervisor is selected by the Mechanical Engineering Department Head with recommendations from the NRL Director and reports directly to the NRL Director. 13.1.2.2 VTAR Reactor Safety Committee The VTAR Reactor Safety committee (RSC), is part of, and answers to, the University Radiation Safety Committee (URSC) as to unresolved i- review questions or votes. The RSC .is charged with reviewing and
- approving all safety evaluation procedure changes, equipment (or system)
! modifications, or any changes which directly effect the safety or con-trol of VTAR Laboratory. The purpose, rules and membership of the RSC are delineated in the following 5 paragraphs and the charter of the RSC
- is included as Appendix 13-1 to this SAR.
I ! 13.1.2.2.1 Purpose of the Reactor Safety Committee
; The pu pose of the VTAR RSC is to provide an independent review and audit of safety aspects. of reactor laboratory operations for the Virginia Tech Argonaut Reactor.
j 13.1 2.2.2 Charter and Rules of the RSC To assure the safety of reactor operations, the review and audit I functions of the RSC are conducted in accordance with an established i charter, written rules of procedure for subcommittee operation, includ- , ing provisions outlined as follows: - l A) The VTAR RSC meets not less than once per calendar quarter, at intervals not to exceed four months, and more f requently as
- necessary. Records are kept of these meetings.
I j B) A quorum for the RSC meetings consists of the chair or desig-
- nated alternate, and one-half of the members or alternate members of the ' RSC. A majority of members present shall be
! regular members. C) Minutes are disseminated, reviewed and approved in a timely j manner. O 201
N F, V 13.1.2.2.3 Membership of the RSC , Membership requirements for the VTAR RSC are outlined below:
-A) The .VTAR RSC censists of nine members including the NRL
- j. Director, Reactor Supervisor, Reactor Radiation Safety Officer, the Chairperson of the RSC and the Director of Safety and
] Health Programs plus four other technical personnel, (at least [ one of which is from outside the University), familiar with ! reactor operations or design, radiation, or systems in use at . or proposed for the reactor facility. These four members are recommended to the Vice-President for Administration and Operations by the Chair of the RSC. B) The Chair of the RSC is a member of URSC. The chair should be q. a Department Head or higher level administrator. j~ C) Appointed members to the RSC serve a 3 year term and are j eligible for re-appointment. 1 j D) Members missing 3 consecutive meetings without prior approval ! of the chair shall be considered-to have resigned. s E) Any member may designate, in writing, a duly qualified repre-sentative selected from an approved list, to act on a temporary j basis, in the members absence.- No more than 2 alternates shall participate on a voting basis in RSC activities at any one
- time.
13.1.2.2.4. Review Function of the RSC { To meet the review requirements of its charter, the RSC reviews shall include but not be limited to the items outlined below. 4 A) Proposed changes in equipment, systems, procedures, tests or ' i experiments. * ! B) All new procedures and revisions having safety significance, ! proposed changes in reactor . laboratory equipment on systems having safety significance. C) All experiments or tests significantly different from pre-viously approved experiments or tests or those that involve an unreviewed safety question. D) Proposed changes in VTAR Technical Specifications or licenses. ! E) Violations of VTAR Technical Specifications, licenses, and i internal procedures or instructions having nuclear safety significance. !O l 202
O F) Deficiencies having nuclear safety significance. I G) Events reportable to NRC in writing. H) Audit ' (I) Operating reports J) Qualifications of prospective reactor staff members. 4 13.1.2.2.5 Audit Function of the RSC The audit function of the RSC includes a selective (but compre-
, hensive) examination of operation logs, records and other documents.
Where necessary, discussions with cognizant personnel also take place. j The individuals immediately responsible for the areas of audit do not take place in the audit.. Areas of the audit cover, but are not limited to, the followingt
! 1) Laboratory operations and administration;
- 2) Qualification and re-training program of the operating 1
personnel; l O 3) oeficiencies e d actie - e
- n te ce<<ect that occur at the Laboratory; t* e aeficie cie-
- 4) S.N.M. Accountability and Safeguards;
- 5) Security plan, emergency plan and implementing procedures.
- Deficiencies found are fowarded to the operating staff and the RSC j in a written report prior to the next regularly scheduled meeting of the i RSC. Deficiencies discovered which ef fect reactor safety are reported to the chair of the RSC immediately.
ii 13.1.2.3. Radiation Safety Organization The Radiation Safety Organization at Virginia Polytechnic Institute and State University consists of the University Radiation Safety 1 Committee, its two subcommittees and the Of fice of Safety and Health Programs, under the direction of the Vice-President for Administration and Operations. The committee is comprised of the members of both the Radioisotope Committee and the Reactor Safety Committee. The Chair of
- one of these committees shall serve as chair of the URSC and the other I
shall serve as the vice chairperson. The URSC is responsible for advising the President of the University, via the Vice-President for j Administration and Operations, on all matters related to radiation
. safety. The primary purpose of the URSC is to act as the final authority pertaining to radiation safety.
203
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. 12.2 Trainin-13.2.1 Operator Trainig i Training of reactor operators at the VTAR is done on an individual basis to fit the trainee's needs. Schedules are arranged in a flexible manner in order to maximise the availability of the reactor as a research and teaching tool. Training procedures and requirements are l' determined by the Director / Reactor Supervisor of the Nuclear Reactor Laboratory. The trainees will receive academic and operational training to adequately prepare them for the written and practical examinations ,
i planned by the Reactor Supervisor and the NRC. ! I 13.2.2 Replacement and Retraining 1 j In the academic environment of the VTAR, reactor start-up, shut-downs, normal and abnormal operations are routinely encountered by the i licensed operators. The reactor operating staf f routinely meets each
' month and discusses the reactor status, maintenance and tests as well as j any other technical or administrative subjects considered pertinent to l
the safe operation of the VTAR. Written minutes of these monthly meet-J ings summarise the reactor operations, maintenance, training, tests and ; i calibration. Every licenced operator shall attend these monthly meet- , j ings unless previously excused by the bactor Supervisor. Changes in , ! procedures, technical specifications and regulations are reviewed and discussed prior to implementation. The reactor secf f participates as i i ) instructors and/or students in formal university cources involving the training of students or reactor operapJr training conducted at the VTAR. ; A training program for the periodic requalification of VTAR operators is i l conducted in accordance with NRC requirements and meets or exceeds the j requirements of 10 CFR 55 Appendix A and ANSI /ANS 15.4-1977. . i } Responsibility for the Adminiutration of the qualification progran rests with the Director and Reactor Supervisor of the Nuclear Reactor I f Laboratory. I ! All licensed operators are required to participate in all phases of l training except where specifically exempted. Persons in training for an- , operators license ' also participate in the requalification program. An i l operator receiving a license during a requalification period is required j to complete only- those portions occurring af ter the effective date the ! license received. I !
- i. The requalification training program ia force at the VTAR consists l
of nine areas discussed in the following sections. The requirements, i delineated in these sections, must be met to successfully complete the j requalification program. ; i I f
- 204
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(~ V) 13.2.2.1 Requalification Schedule The VTAR requalification program is conducted over a period not to exceed two years and then is followed by successive two year programs. 13.2.2.2 Lectures, Reviews and Examinations The requalification program is divided into groups of topics listed in Table 13-1 for which preplanned training or preparation is scheduled. The schedule is set up so that the entire program covering the topics is completed over the two year requalification cycle. Table 13-1 Topics for Requalification I. Reactor Theory and Principles of Operation II. Reactor Systems III. Radiation Safety IV. Routine Experiamnts V. Infrequent Operations VI. Technical Specifications VII. Applicable portions of title 10, CFR VIII. Emergency / Abnormal Procedures (reviewed every quarter) O An examination is administered annually, covering all topics included within that examination period. This examination shall be administered within the first quarter of the year. Any person receiving a license with the effective date of the license occurring after 1 August of the year just prior to the test may be exempted from partic-ipating in the annual requalification examination for that period. Results of this examination are used to determine the operators ptofic-1encies, weaknesses or deficiencies. A special training session is held prior to any fuel handling operation, incore maintenance work or as deemed necessary by the Reactor Supervisor or the RRSO. The requirements of the operation and proced-ures are discussed and practiced to assure proficiency of all persons involved. Any changes in procedures, regulations and technical specifica-tions, as well as any change with a safety significance to the labor-atory are reviewed by all licensed operators. Variots documents, letters and memo's are placed in the Required Reading folder prior to permanent filing. Each operator is responsible for reviewing the required reading folder in a timely manner in order to remain current with the information placed there. O 205
2 F O 13.2.2.3 Requalification Operations i i 4 Over the two year requalification period, licensed individuals nuet f perform a minimum of ten reactivity control manipulations in any com- , bination of reactor start-ups, shutdowns or significant reactivity changes. To insure operator proficiency over a range of ordinary i operations, the following schedule has been devised: A) Each licensed operator shall complete a minimum of one complete , reactor start-up and shutdown at intervals not exceeding four q ' l months.
'i' i At least one start-up and shutdown performed by each reactor B) j operator during the four month period will be supervised and i
- evaluated by a senior reactor operator.
i ; l C) At least one start-up and shutdown performed by each senior ', I
! reactor operator during the four month period will be super- ,1
] vised and evaluated by the Reactor Supervisor or designated 1 siter.' ate. , i D) The Reactor Supervisor must document at least one start-up and I' shutdown during the four month interval. It is the responsibility of each operstar to insure these requirements I O are se and to ed-t; 1 13.2.2.4 Emergency Drills ;
.l l Emergency Drills are held quarterly. At least once per year the
- drill will consist of an exercise which will involve participation of
- such energency organisations as the Blacksburg Fire Department, i University Rescue Squad or other local emergency organisations. Each '
l operator is required to participate in the annual site emergency drill. l A review of the det11s and applicable energency procedures is performed
- following each drill.
i f 13.2.2.5 Absence from Authorised Activities ( l An individual who has not been actively performing the certified ! functions for a period in excess of four months shall demonstrate to the cl
- Reactor Supervisor competsace in reactor operations and theory by com- l l
pisting the following requirements: 4 A) Forform a minimum of four complete reactor start-upe and shut-j downs under the observation of the Resetor Supervisor or desig- f l nated Senior Reactor Operator. j 2) Review all normal and eastgency operating procedures and obtain j an evaluation from the Reactor Supervisor or designated Senior i Reactor Operator. 206
. . ,_ -.. -_ _ _ _ , ___ __, _ _ _ . . _ _ . - , _ .., - - - . . - . _ _ _ _ . ~ . . -
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, C) Satisfactorily complete a comprehensive written examination.
, , , y, . 13.2.2.6 Enceptions 1 di the administrator of the6requalification program, the Reactor 4 ~$upervLlor is exempt from the annual written examination. However, the koactor Supervisor any designate one Senior / Reactor Operator to prepare L,' and 'drade tha annual written examination and, a this case, that senior ,
reactor oparator shall be the only person exemptu from the written exam for that year, but will take an oral exas administered by the Sup,atvisor. No person may be exempted from the written exas in two con-decutive years. , Any person completing a comprehensive written examina- l tion or NRC licensing examination af ter 1 Aug. of the previous year is l o , s i exempt from.the first quarter writ' ten annual exas of the present year. j i>
" 13.2.2.7 Evaluation of Operators
, l 13.2.2.7.1 Go the Job Training
){
,' . Eacn licensed individual is required to demonstrate satisfactory f '.
understanding of the op$ ration of the facility systems, operating pro-cedurm , and system changes ' during a quarterly walk-through of the facility . administered by ':he' Reactor Supervisor or a designated Senior 1 Reactor Operator. s 13.2.2.7.2 Grade RequirementA s
!' All opht score are required to complete the annual comprehensive etastnation satisfactorily according to the following requirements.
A) Anoverallgradeofif01isrequiredtobeconsideredpassing. '
, i An individual vectLon of the annual exas will require a grade i
8) i of 70% or greater.'
/
i C) An overall grade of less than 80% shall require renoval from
, licensed duttee and placement on an accelerated requalification progree uncti satisfactory terforesnee has been certified by k ,
\ ,
the Rosetor Sup'ervisor ae, evidenced by the individual success-fully passing a comprehehsive~ written examination.
! D)
A grade of less than 701 in any one section of the annual exas-O ination shall require the individual to review the applicable g esterial and successfully pass a written test on that section. X) An individual shall be considered' to have failed if the indi-vidual receives less than 70% on more than two sections ~ of the annual comptehensive essetnation regardless of the overall grade. This individual . shall ' be placed on an accelerated requalificatt.on progree af in part C above. ~ < ~ e 1 ,. k ') . 207'
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1 Y p i O 13.2.2.7.3 Accelerated training y' Accelerated training programs are completed within five months following the grading of the examination._ Furthermore, within one month after grading the . examination, an evaluation is made by the Reactor Supervisor, considering the individuals past performance, past test scores and current deficiencies. This evaluation shall be used to determine the specific areas of re-training to be used for that h individual. 13.2.2.8 Requalification Records , (? i Records are kept to assure that all the requirements of tl a VTAR Operator Requalification Program are met. i Each operator and trainee has an individual file which contains 3~ copies of written examinations taken, the answers given by the operator, g the results of any evaluations and documentation of any additional training the operator has received in deficient areas .or special
,. y activities.
i
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Written comprehensive examinations and answers given by each operator shall be retained for a period of at least two years or until
, permanent removal of the operator from licensed duties, whichever is the f \ shorter period. Also kept for this length of time are the emergency
- procedure review status of each operator's start up and shut down I
performance. Reactor operator meeting minutes, with _a sunuaary of lecture topics ist were discussed, are kept for five years.
' .I>' 13.2.2.9 Requalification Document Review l All operators are to review the required reading folder at frequent i ;i intervals. Each operator shall review the contents of all emergency and
+ abnormal procedures each quarter and all operators shall be cognizant of the facility Technical Specifications, design and procedure changes. b These reviews are left to each operator to perform, however, at each Reactor Operator meeting, deficiencies are sought and corrective actions taken. ) l 1 p 13.3 Plant Procedures ) This section describes the procedures pertinent to normal operation and administration of the VTAR Laboratory including the performance of. experiments, modifications, repairs and cests. The Reactor Supervisor 4 is' responsible for ensuring compliance with the established controls. The Reactor Supervisor _is responsible for the preparation of detailed written procedures for normal and emergency operations. These procedures are approved by the VTAR RSC before implementation. However, O 208
.i - . .
O if the need arises for a new procedure between scheduled meetings of the RSC en Interim Procedure may be used until the procedure is approved at the ' next regularly scheduled RSC meeting. The interim procedure shall have a safety analysis attached and shall be reviet"ed by all available licensed operators and the Reactor Radiation Safety Officer. Disapproval by any one of the reviewing persons shall be reason for not implementing the interim procedure prior to review by the RSC. Procedure changes which do not consistute an unreviewed safety question may be implemented by the Reactor Supervisor af ter approval by the Reactor Supervisor, all Senior Reactor Operators and the Reactor Radiation Safety Officer. This change in approved procedure is then reviewed by the RSC at the next meeting following institution of the change. 13.3.1 Administrative Procedures -- Access Control The outer doors of the reactor laboratory are normally locked. During operation of the reactor, the operator manipulating the reactor controls entry to the reactor area and can limit entry at any time. All personnel entering the reactor cell must wear personnel dosimetry. Visitors are escorted and must sign in and out of the personnel dosimeter log book provided. , When the reactor is not operating, unsupervised access is permitted to persons authorized by the Reactor Supervisor. Entry requirements are covered under the Physical Security Plan of the Laboratory. Facility keys are issued to persons specified by the Reactor Supervisor. Details on key control are contained in the Physical Security Plan of the VTAR, submitted separately, and are witheid from public disclosure, subject to 10 CFR 2.790(d). 13.3.2 Operating and Maintenance Procedures 1 13.3.2.1 Routine Operations and Records-Manipulation of the . VTAR reactor controls is permitted by a VTAR-Licensed Operator or by a non-licensed person, in license training, under direct supervision of a licensed ' reactor operator. A senior reactor operator is on call at all times that the reactor is operating and a second person, duly qualified as a Reactor Assistant,- is ' present in the laboratory to assist the operator at the console. An Operator-in-Charge (OIC) is designated for each reactor' operation. The OIC is ) responsible for ensuring that the following requirements are met prior. j to and during reactor operation. !
- 1) The OIC is in 'a position to operate the controls of the reactor. The OIC cannot leave the control console area. l
- 2) The correct starc "7 check and shutdown check procedures are I followed -and convertational log . sheets are filled out' for all operators.
209 L 2.
l l l I i
- 3) Any proposed experiment or irradiation is correctly authorized and any requirements noted have been complied with.
- 4) All irradiated samples moved within the laboratory are
- monitored.
- 5) The experimenter and senior reactor operator on call are informed in case of any unusual or unexpected occurrence, apparent equipment failure or other deviation.
The Reactor Start-up Procedure (VTAR OP-I.1) ensures that the reactor and experimental configuration are correct, and removable t shielding is in place or personnel are otherwise protected, the instruments are calibrated and functioning, the scram and interlock circuits are functioning and scram setpoints are properly set, and the facility is otherwise in proper condition for operation. Two log books of VTAR operations with information related to start-up checks and Reactor usage are maintained on the Reactor Control Room. The Reactor log includes the stiartup check list and the Reactor Run Sheet with the name of - the operator in charge, the experimental con-figuration, special instructions, periodic instrument readings and-con-trol rod positions. The conversational log book contains test results, f~ methods and reasons for shutdowns, changes of power or experimental configuration, and any other notations the Reactor Operator deems appropriate. The procedure manual includes Standard Operating Procedures (SOP), Emergency Operating procedures, Maintenance procedures, Security pro-cedures, Administrative and Radiological Procedures. 13.3.2.2 Routine Tests, Maintenance and Monitoring The Reactor Supervisor maintains a program for regular testing of all safety systems and certain reactor components.- In additon to the startup checks of instruments, periodic checks and maintenance are performed on a conditional quarterly, semi-annual and annual basis. All required maintenance items are listed in the VTAR Procedures Manual Part IV, which. lists-the requirements of each mainte- ! nance item as well as its periodicity. i Any compromise of the safety-related systems for the VTAR is suf-ficient cause for stopping reactor operation until the situation is corrected. The results of all periodic tests, checks or maintenance items listed in Part IV . of the Procedure Manual are recorded on a data sheet and placed in the appropriate maintenance file. O 210
O The Reactor Radiation Safety Officer routinely performs surveys of the reactor and reactor area, especially during reactor operation, to check radiation levels. . The results of such monitoring are recorded and maintained in a file .at the University Radiation Safety Office. The detection of any significant or abnormal radiation levels requires immediate investigation and subsequent corrective 4.ction to alleviate the problem. A TLD badge service is provided as part of the Personnel Monitoring Program. These badges are supplemented in the reactor area by pocket dosimeters which are also used for occasional visitors. The Reactor Radiation Safety Officer is in charge of badging and associated records. 13.3.2.2.1 Operational Startup Checks
; The startup checks are initiated and satisfactorily completed with-in four hours prior to reactor startup. The scope and detail of the startup checks are indicated'in VTAR SOP I.1, Reactor Startup Procedure, and are summarized below: -
- 1) All movable shielding is in place to minimize personnel
; exposure.
- 2) All fuel storage pits are covered and locked.
- 3) The primary and secondary coolant systems are correctly aligned
). and verified for operation.
- 4) The console equipment is checked to insure all items are func-tioning correctly.
- 5) The' proper functioning of the airborne particulate detector, portable radiation survey instruments, and the reactor room ventilation system.
- 6) The neutron source is verified to be out of the core area.
- 7) The calibration of all nuclear instrumentation is checked.
- 8) The proper functioning of the startup ' interlock system is verified.
- 9) The proper functioning .of the reactor safety systems including reactor trips and annunciators alarm systems is verified.
- 10) The proper functioning of any repaired instrument is verified prior to startup.
i
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sm b .13.3.2.2.2 Quarterly Maintenance Checks The following maintenance items are checked on a quarterly basis:
- 1) Temperature indicator calibration.
- 2) Control Rod Position Indicator Calibration
- 3) Area Radiation Surveys
- 4) High Water level scram test.
- 5) Pump on/ valve shut interlock operability test.
13.3.2.2.3 Semi-Annual Mainte' nance Checks The following maintenance items are checked on a semi-annual basis:
- 1) Permanent Radiation Monitor calibration.
- 2) Air Particulate Fission Product Monitor calibration J
- 3) Startup interlock test
- 4) Nuclear Instrumentation drawer calibrations
- 5) Heat Balance Reactor Power calibration
- 6) Automatic power controller calibration
- 7) Conductivity meter test
- 8) Control rod drop time measurements.
13.3.2.2.4 Annual Maintenance checks The following maintenance items are checked on an annual basis.
- 1) Control rod reactivity worth and rate measurements
- 2) Control rod speed measurements
- 3) Primary dump valve test
- 4) Reactor power flux calibration
- 5) Control rod drive mechanism check 6)- Control rod checks
-D
.212
O 7) Auto-heat exchanger valve check .i
- 8) Annunciator alare panel test
- 9) Reactor fuel inspection
- 10) Reactivity measurements 13.4 Emergency Planning The site emergency plan for the VTAR Laboratory is described in the
" Virginia Polytechnic Institute and State University Emergency Response Plan for the Training and Research Reactor," and in VTAR Operating Pro-cedures which detail the responsibilities of, procedures and actions to j be taken by the operations personnel in the event of emergency conditions.
The Director of the Nuclear Reactor Laboratory has overall respons-ibility for the handling of emergency situations in his position as the emergency director.- These duties include coordination with the Law Enforcement, Disaster Preparedness, local and state health agencies and with the Nuclear Regulatory Commission. The VTAR Emergency Response Plan has been submitted as a separate document and includes all applicable procedures to effectively conduct O~ the activities required by the Plan. 13.5 Security The plans for physical protection of the VTAR Laboratory are described in the Physical Security Plan, already submitted to the NRC under a separate cover and witheid from public disclosure pursuant to 10 CFR 2.790(d). O-9 , 213
i APPENDIX 13-I ARTICLES OF CHARTER FOR THE VTAR REACTOR SAFETY COMMITTEE I. 'The Virginia Polytechnic Institute and State University Reactor Safety Committee (RSC), has been established with the authority to advise and regulate the safe operation of the nuclear research reactor at. Virginia Polytechnic Institute and State University (VPI&SU) as defined in this charter. II. The Reactor Safety Committee shall consist of nine voting members, appointed by the Vice-President for. Administration and Operations. The Director and Reactor Supervisor of the Nuclear Reactor Laboratory (NRL) shall be voting members of the RSC by virtue of their positions (ex-officio). The Reactor Radiation Safety Officer and the Director. of Health and Safety Programs
- shall be voting members of the RSC by virtue of their positions (ex-officio). The other voting members of the RSC shall include the chairman of the RSC and four technical persons familiar with reactor operations, disciplines associated with reactor design or operations, radiation, systems in use at or proposed for ' the reactor facility, or state and federal regulations as they pertain to the reactor laboratory. Of the four technical O representatives, one is to be from ..outside. the University.
Adjunct members may be appointed for their consulting services but shall not have voting privileges. III. Candidates for membership in the RSC may be recommended to the Vice-President for Administration and Operations by the chair of the RSC. Candidate nominations are made by the present members and may be selected for recommendation by a review of overall qualifications and a majority vote of the quorum present.
- Members are appointed to the RSC for a three year term and are The chair of the RSC should be an
~
- eligible for reappointment.
i administrator of department head level or higher.- Members missing three consecutive meetings without prior approval of the chair shall be considered to have resigned. Any member may
- designate, in writing, an approved alternate to act on a temporary basis, in the members absence. A list of approved-I alternates will be provided. No more than two alternates shall
- . participate on a voting basis in RSC activities at any one time.
IV. The chair of the RSC shall call meetings not less than once per calendar quarter at intervals not to exceed four months; meetings may be called more frequently if ' necessary. - A quorum of the RSC. shall. consist of a minimum of five voting members, including the l chair or approved alternate and one-half of the total voting j members or their alternates, of the RSC. A majority' of voting i 0; 214
.,_ . , , . . _ , , . , _ , . . _ -~
O members present shall be regular members. Annually, a joint meeting with the Radioisotope Committee shall be held. This joint meeting shall constitute the regular annual meeting of the University Radiation Safety Committee (URSC). Minutes of the; RSC meetings shall be recorded and furnished within ten working days, to the members of the RSC, prior to the next regularly scheduled meeting. Minutes shall also be furnished to the members of the URSC at the annual-joint meeting. The Secretary of the RSC shall be appointed by the RSC Chair. V. The RSC review shall include but not be limited to: A. All safety evaluations for: 1. changes to procedures, equipment or systems; and 2. tests or experiments conducted without Nuclear Regulatory Commission (NRC) approval under the provisions of 10 CFR 50.59 to verify that such actions do not constitute an unreviewed safety question (Review for approval); B. Proposed changes to procedures, equipment or systems that change the original intent or use, or those that involve an unreviewed safety question as defined in 10 CFR 50.59 (Review for approval);
% C. Proposed tests or experiments which are significantly i different from previously approved tests or experiments, or those that involve an unreviewed safety question as defined in 10 CFR 50.59 (Review for approval);
D. Proposed changes in Technical Specifications or facility !' licenses (Review for approval); 1 E. Violations of applicable codes, regulations, orders, Technical Specifications, license requirements or of internal procedures or instructions having nuclear - safety
; significance (Review for action);
F. Significant operating . abnormalities, or deviations from normal and expected perfo mance of facility equipment that
- affect nuclear safety (Review for action);
j G. Events which require written reports to the NRC (Review for action); H. Audit reports (Review for approval); I. Operating reports (Review for approval); J. Qualifications of prospective reactor staff members (Review
-for approval).
LO I L l 215
.O VI. Audits of the laboratory activities shall be performed by an .
I Audit Sub-Committee approved by the RSC. In no case will personnal responsible for the items being audited solely perform the audit. Audits shall take place annually at intervals no less than nine months and not exceeding fif teen months. The Audit Sub-Comittee shall consist of three persons designated by the i RSC chair at the regular meeting just prior to the quarter in which the audit is to be performed. The Audit Sub-Committee members . mist have familiarity with the areas to be audited. The areas of audit shall include: A. Special Nuclear Material (SNN) shall be inventoried by the Reactor Supervisor, Reactor Radiation Safety Officer and one member of the audit group; i B. Reactor Administration, Operations and Training; C. Security and Emergency Planning and; j D. Safeguards. VII. Decisions by the RSC become effective immediately upon a majority vote; however, if a vote is not unanimous, a member may request,' at the time of the vote, that a review be made by the URSC.. At O any time, any two members have the right to call the committee together, notifying every member of the committee in writing of , the purpose of the meeting at least one week in advance. This-charter may be amended at any regular RSC meeting by a majority vote of the entire membership. If a copy of the proposed amendment is submitted in writing at the immediately preceding general meeting, then the charter may be amended by a two-thirds ~ vote of those voting if a quorum is present. i J 216 4
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_ _ . _ _ _ . _ _ _ ~ . 4 h 14. Initial Test Program 14.1 Specific Information to be Included in Preliminary Safety Analysis Reports l The enclosed documents in this report are serving as the VTAR I facility final safety analysis report for an operating facility;
- i. therefore, this .section is not applicable.
l 14.2 Specific Information to be Included in the Final Safety Analysis Report i Since there are numerous changes which have been made to the VTAR during the current overhaul program, an initial test program is considered appropriate. Due to extensive system modifications a Preoperational and Startup Test Program has been developed. -This format will be adhered to for any
~
significant modifications to systems in the future, as determined by the Reactor Safety Committee. Following significant modifications, a Preoperational and Startup Test Program will be effected in the following three phases: Phase I Preoperational Testing Phase II Low Power Test Program i Phase III Step Ascent to Full Power Operations Phase I will include all electrical tests, instrument tests and alignments, functional tests and " hot" ' operational . tests (if applicable). This phase will also include SAR' documentation, NRC and/or RSC inspections'and reviews. Preparations for fuel loading are included I in this phase. Also included in this phase is the- establishment of a Startup and Testing subcommittee by the Reactor Safety Committee to review data, drawings, safety anlayses, etc. and approval of changes to startup and test procedures. Fuel' loading will. not be performed until all changes, tests, l analyses, etc. are accepted by the Reactor Safety Committee and by the j NRC, if deemed' appropriate. l . Phase II, the low power test program, is then enacted. - This phase consists of fuel loading and if required, an approach- to critical. Following criticality, low power reactivity measurements are made (if required). Additional tests such as. functional automatic scram tests
.will be performed.
j 217
-_ . . _ -__ . . . . . ._ 1 . _ . . . . .~. -
Phase III is the step approach to 100% power. This includes flux calibrations, heat balances, system performance tests and baseline data at power levels of 25, 50, and 100 KW. Power range scram setpoints are reduced to 50%,100%, and 120%, respectively during the ascent. Note that this program implementation. is dependent on the nature !~ and tw;nitude of changes to reactor systems as determined by the Reactor Safety Conaittee. l Due to the extensive. system changes (although no core or physics
- changes were involved) this program will be adhered to following acceptance of this SAR by the Reactor Safety Committee.
O 1 1 A O 218
- 15. Accident Analysis 15.1 Introduction This chapter addresses the safety of the VTAR facility, including analyses of responses to postuisted process variable disturbances and malfunctions or equipment failures.
Previous' chapters have discussed the structures, systems and components important to safety. In this chapter, the effects of process disturbances and component failures are examined to determine their consequences and to evaluate the capability of the VTAR facility to j control or accomodate such failures and situations. Accident categories considered include nuclear excursions, loss of I coolant, loss of flow, safety control rod malfunctions, release of fission products to the environment, etc. In each case, a ury conservative approach to both problem formulation and evaluation was taken. The results presented are therefore considered to be conservative estimates of the severity of any given problem or situation. 15.2 Reactivity Additions
, 15.2.1 Cold Water Accident At an operating power level of 100 kW, the (worst case) operational design basis accident is a power excursion caused by the rapid addition of cold water into the core.
Under the worst possible conditions, the inlet secondary coolant could have a temperature as low as 320F(00C). The worst case scenario is as follows:- The reactor has been operating at 100 kW with primary coolant temperature just below the high temperature annunciator level of -1650F. The. reactor is then shut down normally, then re-started while primary coolant temperature remains at 1650F. As criticality is reached in the source range, with a reactor period of 30 seconds (excess &/k = 0.00143%), the secondary system goes to full flow instantaneously with secondary water at 320F. . At this point, the primary coolant temperature is assumed to drop immediately to i 320F (over the entire core, with no mixing). Since the measured moderator temperature coefficient of reactivity is -0.004% &/k/*F, this temperature transient could introduce a maximum positive reactivity change of 0.532% R/k, which, when added to the already existing ' 0.00143% &/k, gives a total of 0.53343% &/k. This figure does not approach that required for prompt criticality. Under these conditions, the maximum power achieved is approximately 246 kW and the energy released is .approximately 340 kWs. The maxi: mum temperature achieved O l 5 219
4 l during the excursion is approximately 2030F, which is below the boiling temperature of the moderator / coolant and poses no threat of fuel element failure. Also, since the temperature falls quickly, there is no danger of overheating to the fuel, clad, or graphite. 15.2.2 Sudden Addition of Maximum Excess Reactivity An inadvertent power pulse is the consequence of a step insertion of positive reactivity. Several mechanisms might provide this sudden i excess reactivity, including the addition of a reactivity source other than that available from the reactor itself. This section will consider the credibility and consequences of such a sudden insertion into the reactor which is operating in the normal steady state mode. The current license for the VIAR limits _ maximum excess reactivity
; to 0.6% &/k. Since both the fuel consumption and maximum xenon concentration are small at 100 kW power levels, nearly all of the excess reactivity may be available in normal operating conditions. Because of previous fuel burnup, it is assumed that, regardless of the primary system temperature, the maximum excess reactivity is no greater than 0.6% &/k. The main reason for this assumption is that in addition to limiting temperatures to known values, annual measurements of core excess reactivity are performed. For every startup, the total excess reactivity is determined, including experiments. The reactor may not be operated with a secondary temperature below 550 F or with primary inlet O temperature below 700F. Except by deliberate additions of positive reactivity or rearrangement of the core, there is virtually no likelihood of obtaining an excess reactivity greater than 0.6% &/k.
Experiments inserted into the reactor are limited to 0.3% &/k, with j- most experiments not exceeding a few hundredths of a percent. Any anticipated insertion of the maximum excess reactivity, whether credible or not, would result in a maximum power of 2.468 MW. The
- ~ maximum energy release for this excursion is 876 kWs, resulting in an average fuel plate temperature of 225 0F. Since the temperature at which fuel melting could be expected 'to begin is on the order of 1000 F, 0 no
} problem would be expected to occur in this regard. - A comparison with Spert I fuels and results can be made. Spert I had plate type fuel, similar to that in the VTAR design. Enrichment was j 93% U-235, again similar to that in VTAR. The Spert I destructive test indicated molten fuel eruptions at a temperature of approximately 10000F (5900C). The Spert I destructive test excursion was of some 30.7 MWs.. As _ can be seen from the results of calculations for the VTAR- (above), neither melting nor an . associated volume change -leading to cladding j f ailure is credible, _ since the peak fuel temperature rise is well below
- the eutectic temperature and the margin of 7750F is believed adequate.
Since the maximum excess reactivity is - below that required for i achieving prompt criticality, the peak power, energy release, and . fuel
- O l
220
. . . . . -- - .. - .=. ___-- -
temperatures are somewhat. more limited in this type of excursion than would be the case with reactors having a greater excess reactivity. Because 'the above results are based on the instantaneous insertion of
- the maximum available excess reactivity, any credible accident would, because of the longer duration, produce fuel temperatures which are much lower than the maximum 10000F (5900C). In addition, the above results were found with no heat transfer from the core by any means other than the normal primary cooling system flow. Since conduction would almost certainly occur with the aluminum core boxes and the graphite slab, lower peak temperatures would be expected to occur in reality.
- Therefore, there is no safety hazard from rapid insertion of the maximum available excess reactivity of 0.6% &/k. Melting of fuel or cladding .
would not result from this accident, and core disruption or l rearrangement, if any, would be minimal. : 15.2.3 Safety Control Blade System Malfunction
- Three types of control rod malfunctions must be considered. These
! malfunctions are stuck rods, dropped rods, and rods moving independently
, of control.
Since blade (rod) withdrawal increases the reactivity of the reactor, a rod which stuck on withdrawal would simply prevent a power increase. This situation is clearly not a safety problem. On the other i O hand, a stuck rod during insertion or during a scram could pose a hazard. The total worth of all the rods combined is about 2% &/k. Since the moderator worth is in the area of 30% &/k, even in the event that all rods remain in the full out position, dumping the moderator from the l core will provide - a sufficient shut-down margin to ensure that the- , reactor is shutdown. This action, while ensuring reactor . shutdown, ; constitutes a loss of coolant accident during 500 kW operations (see i } section 15.4). During 100 kW operations, dumping the moderator is a l
;. normal part of reactor shutdown and as such cannot be considered to
, constitute a hazard. Any single stuck rod anst be a lesser problem than l that noted above and is considered not to constitute a hazard. ! l A single rod dropping completely into its full in position could 1 conceivably cause a power drop followed by a "self-restart". The reason 4 this anst be considered is that as the dropped rod causes rapid power 4 loss, the moderator temperature may be expected to f all. The negative
- moderator temperature coefficent coupled with the lowered moderator
- temperature could result in an increase in reactivity. In order for i self-restart to occur, the - gain in reactivity from temperature drop 1 would have to more thaa equal the reactivity lost because of the dropped 1
rod. The three rods of highest worth are Safety 1. Safety. 2 and the shim rod, having worths of 0.675 to 0.533% */k. To equal these worths, moderator temperature would have to drop an amount between 1690F and O 221
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133 F from its normal core exit temperature of about 1300F. This is O. ' clearly not credible. .In the case of the regulating rod, however, the
, required temperature drop is only about 330F, which may be possible. In t
this event, a restart might occur with the reactor achieving a lower l power level than the operating level prior to the drop. In the worst L case of this event, the reactor power would fall with a period on the order of -300 seconds. The consequent rise would be of the same order of magnitude, thus affording the operator ample time to achieve a full reactor shutdown prior to achieving the self-restart. Should the operator fail to shut the reactor down . the power level could be expected to stabilize in the vicinity of 13 kW. This situation, should 4 it occur, could not pose a safety hazard. A rod moving into the reactor independently of control will cause a i decrease in reactivity and a consequent reduction in power. A rod moving in the opposite direction will cause an increase in power. Since the motor and gearing prevent a reactivity increase of more than 0.02% Ak/k per second, the power increase would be a relatively gentle ramp up rather than a spike such as would be caused by an instantaneous insertion of reactivity. The automatic scram bus limits both power and i period. If power rise exceeds a period of 5 seconds or a power level of j 120 kW, the automatic system would shut the reactor down even without i operator intervention. In the event of any rod moving independently of i control, the standard procedure is for the operator to shutdown the reactor manually (scram) as soon as the situation is recognized. t In reference 15.21, a fault-tree analysis of the VTAR safety system l is described. This analysis considered the history of safety and
; control rod malfunctions for approximately 10 years of operation. .The j
results of the fault-tree analysis indicated that tjeprobabilityfora j total loss of the safety system was about one per 10 years. 15.3 Loss of Flow Accidents (LOFA) 15.3.1 LOFA at 500 kW a Loss of flow accidents were investigated for two cases; loss of i flow with all control rods out (ARO) and loss of flow with all control i rods in (ARI). In the cases studied, the initial fuel temperature was taken to be the hot channel hot spot temperature which is 2000F. The initial coolant temperature was taken to be the coolant temperature in i' the hot channel at 0 the same axial location as the fuel hot spot. This temperature is 131 F. I 15.3.1.1 LOFA with ARO One scenario ' of this accident was considered with the -following i assumptions: 500 kW will continue to be generated, all 4 control rods l remain withdrawn, and an extremely conservative assumption that no heat . { transfer occurs between the fuel and the coolant. In this scenario, 222 L- _ . _ . _ - _ . - _ _ - _ - - -. .. - -
. . _ _ . - ___ ~ - -
fuel melting can occur in 40 seconds in the hot channel hot spot after flow stagnation (see Fig. 15-1). It is well to note that this particular scenario is_a worst case and is not considered credible; the study was made in this manner so that equipment and procedures could be designed to deal with this accident on its worst possible level. The second case of this accident is the least severe case, which will result in instantaneous initiation of natural circulation. In this case, the fuel temperature will rise and the coolant can begin to boil in approximately 5 seconds in the hot channel hot spot. This is again assuming that 500 KW continues to be generated. In either scenario, the best course of action is to dump the coolant from the core. The core will become suberitical by 30% ak/k. l The water will be replaced by air at room temperature. There is normally an induced air flow in the core due to the ventilation system, however it will be assumed that there is no air flow initially. Using the extremely conservative assumption that no heat is transferred to the , air, fuel melting could begin in 110 minutes (Fig. 15-2). If I circulation of the air begins, both the fuel and air temperatures will l rise, but auch less rapidly than in the above case (Fig. 15-3). This situation is the same as a loss of coolant accident (LOCA) which will be i discussed later. If any control rods are stuck, they should be forced in. The dump valve should be closed and the emergency core cooling system should be started. This system is more than adequate for removing decay heat since it is designed to remove a 100 kW heat load. O 15.3.1.2 LOFA with ARI
- In this case, coolant flow goes stagnant immediately after shutdown from 500 kW operation. Under this condition water could begin to boil in about 36 minutes in the hot channel hot spot (Fig. 15-4). This accident should be responded to by initiating the emergency cooling system.
- 15.3.2 LOFA at 100 kW Loss of flow accidents for 100 kW were evaluated in the same manner as 500_ kW except that experimental data is also available _ from 100 kW operation.
15.3.2.1 LOFA with ARO As in the 500 kW LOFA, the worst case scenario was considered with the following assumptions: all four control rods withdrawn, 100 kW continues to be generated, and no heat transfer occurs between the fuel 4 and the coolant.' In this scenario, fuel melting could occur in 213 seconds in the hot channel hot spot. However it is a worst case basis of procedures and equipment. l O 223
O . O o . 1200-- iioo - ioOo - 90 0-800-FIGURE 15 - 1 FUEL TEMP LOFA ARO 7U - NO HEAT TRANSFER ME OUT OF FUEL
- e soo-s y 500-B g 400-r
. @ 300- . d i g 200-e , l00 - i i i ' ' s O i a 35 40 45 O 5 10 l$ 20 25 30 ! TIME AFTER ACCIDENT INITIATION (SEC) i
O O O 1200-110 0 - 1000- . 900-FIGURE 15-2 ggg_ FUEL TEMP LOFA ARI / LOCA 700 - DECAY HEATING OF FUEL
. - NO HEAT TRANSFER i 5 {600- OUT OF FUEL
- g '
# 500-B
$ 400 -
r I
@ 300 -
Gl
@ 200 -
10 0 - O , , , , , , , O 1000 2000 3000 4000 5000 6000 7000 TIME AFTER ACCIDENT INITIATION ( SEC)
1 O i 300 - 280 - FUEL HOT SPOT TEMP ,- 260 - 240 - - m 220 -
- u. ,
C ,- g, 200 -
,/
5 I iso - 1 EO I I
% iso -
* ' \ AIR TEMP AT SAME AXIAL '
14 0 - LOCATION AS HOT SPOT FIGURE 15-3
~
!i - 8
- FUEL TEMP I
12 0 - E I LOCA E
' NATURAL CIRCULATION o 10 0 OF AIR i
z 4
-I
$ 80 - l n.
I F 60 O [;j 40 - 2 20 - , O - I I I I I O 1000 2000 3000 4000 5000 i O TIME AFTER ACC! DENT INITIATION (SEC) l 226
1 I I O l C > 2., , IL 2
- 220 -
FUEL HOT SPOT TEMP E 2 200
=
l 4 m#
@ 180 -
,s N - '
- 16 0 -
14 0
/ COOLANT TEMP AT SAME AXIAL o /
LOCATION AS HOT SPOT O
'20 -
o FIGURE 15 - 4 O =
* '00 -
Fuet Tsue LOFA ARI 2
= NATURAL CIRCULATION 80 -
OF COOLANT g o Z 60 - E 40 - 20 - O i i i i i i 8 ' ' ' 8 0 200 400 600 800 1000 1200 1400 1600 1800 2000 2200 T;ME AFTER ACCIOENT INITIATION ( SEC.) O 227 d
.- - . ~ _ - - -. - .- - .. - - - _ _
l 1 The least severe case of this accident is where natural circulation O of the coe1 ant deveteFs inseantaneous1F. 1n this case fue1 temFerature rises and the coolant could begin to boil in about 60 seconds. In the case of a LOFA at 100 kW, the best course of action is to shut down and dump the coolant / moderator. According to experimental data collected in previous years, temperature rise in the fuel is negligible (3-50F) after i a moderator dump from extended operation at 100 kW. Hence, no other action would be necessary other than, of course, correction of the problem that caused the LOFA. 15.3.2.2 LOFA with ARI Since experimental data indicates that decay heating is negligible at 100 kW, a LOFA with ARI is not considered a safety problem. 15.4 Loss of Coolant Accidents (LOCA) 4
- Since the primary mechanism of shutdown from 100 kW operation is to j dump the coolant / moderator f rom the core, and experimental data shows that fuel temperature rise is negligible af ter the dump, a LOCA at 100 kW is not considered a safety problem. However, in the absence of I
experimental and test data, loss of coolant accidents were evaluated for 500 kW operation in the same manner as loss of flow accidents. That is, extremely conservative assumptions were made so~ that accidents can be handled on their worst possible level thus lending a high degree of j conservatism to the way an actual accident would be handled. 15.4.1 LOCA with no Heat Transfer A LOCA may occur as a result of a rupture of the primary piping, ' rupture of a core tank, or operator error, such as accidentally opening ; the dump valve. In the worst case, where no heat is transferred out of the fuel, fuel melting could occur in about 110 minutes (see Fig. 15-2). Though incredible, these assumptions were made to assure consideration of the worst case. 15.4.2 LOCA with Natural Circulation ' In a LOCA where natural air circulation begins, temperature rise of the fuel will be much slower and no melting can occur (see Fig. 15-3). In the absence of experimental data, the best way to handle a LOCA at 500 kW is to ensure that the control rods are inserted, close the dump valve if it accidentally opened, and initiate the emergency cooling system. This system is designed so that, even in the event of a pipe break, 2900 gallons of cool water can be pumped through the core. This water would accumulate in the process pit which has a capacity of 2900 gallons. i i i f O 228 l
'15.5- Fuel Element Failure Several accident scenarios could possibly . result in the release of fission products into the environment thus making a study such as the following necessary. These accidents are: fuel melting, an extremely severe . earthquake, or some other form of core crushing or loss of cladding integrity. The following sections discuss the methodology and res.ults of a hypothetical total core failure considered for both 100 and 500 kW power levels.
15.5.1 Method' ology The analytical methods used to . predict thyroid and whole body radiation doses following a complete fuel failure at the VTAR facility are summarized in this section. The methodology presented includes basic equations and theory as well as basic input data and information for the calculations such as diffusion coefficients, release fractions i from the fuel and hypothetical transport of the released radionuclides. 15.5.1.1 Basic Equations 15.5.1.1.1 Transport of Released Radionuclides Assuming the reactor cell ventilation stops after the accident occurs (as would be the case) the transport model for the radionuclides consists of two compartments; the reactor cell and the environment. ,i The release into the actual environment, that is, into the atmosphere where it cannot be controlled, is further hampered by the placement of 1 the reactor cell. It was assumed that the only point of leakage j directly to the environment would be the west wall where the large
- equipment door is located. The radioisotope inventory in the reactor i
cell is lost by leakage to the environment and by radioactive decay. The environment loses radioisotopes only by decay.
. Given the two compartment model, the total activity of a particular 1 nuclide, i, present in the reactor cell is given in the following I equation:
-( Ag + LR)t A =fAgge where A go = initial activity of nuclide i A g= activity of nuclide i at time t in Reactor cell
' i fg= fraction of nuclide i released from the fuel 1 = decay constant for nuclide i LR = fractional loss rate from the reactor cell. O 229
v The activity that escapes to the environment per unit time is given by: A = LR f A b* Therefore, the total ac.tivity of a nuclide that has escaped in a time interval of Tg to T2 is given by the following LRaf A b
"(Ai + LR)T ~* -( A + LR)T 2 i
Ai + LR 4 15.5.1.1.2 Dose equations l The isotopes of greatest significance in case of an accident are , the noble gases and radioiodines. The noble gases contribute solely by means of the immersion dose, while radioiodines contribute primarily to the thyroid dose through inhalation. The equations used to calculate the whole body dose and the thyroid dose are: WHOLE BODY DOSE O D=Q+m + CF g THYROID DOSE D=Q *
- BR = CF where D = Dose in rem Qt = Release of isotope type i to the atmosphere (C1) f = atmospheric dispersion factor (sec/m )3 l
CF 1 = whole body dose conversion factor for isotope type 1 3 Rem-a
- - [C-cec)
CF) .
= Thyroid dose . conversion factor for radioiodine type j Rem (Inhaledcurie) l BR = Breathing rate (m 3/sec)
O 230
- i
~ l O 15.5.1.2 Input Data 15.5.1.2.1 Radioisotope Inventory in Core l Since power level and operating times are extremely varied at the VTAR facility, it is very difficult to know the activity in the core at any given time. Therefore, for simplicity and to make this a worst case study, it will be assumed that an equilibrium inventory of the isotopes of interest is present at the time of the accident. This inventory is shown in Table 15-1 for both 100 kW and 500 kW. It would take over a Table 15-1 Equilibrium Gaseous Radionuclide Concentrations
- at 100 kW and 500 kW Maximum Power 100 W Activity 500 W Activity Isotope (Curies) (Curies)
Xe-131M 10 49 Xe-133M 744 3839 Xe-133 5172 26687 Xe-135 5164 26646 Xe-135M 856 4416 Xe-138 4889 25227 Kr-83M 4153 21428 l Kr-85M 1025 5289 Kr-87 1989 14504 Kr-88 2810 14499 Kr-89 3658 18880 I-131 2046 10557 I-132 3316 17110 1-133 5309 27389 I-134 5968 30794 l I-135 4965 25619 I-136M 1559 8044 These values obtained by modifying information from University of Florida Test Reactor (UFTR). Determination was found using ratios of
. flux and fuel loading.
231 1 I
,. ._ - . . . _ _ _ - - _ . .- . . . _ - - .-- - ~_ .
I i month of continuous operation to reach these levels of activity. The k VTAR facility operates a maximum of 8 hours per day for 5 days a week making this a conservative assumption. 15.5.1.2.2 Dispersion Factors and Leak Rate Leakage of . isotopes to the outside environment will occur through the west wall of the reactor cell at a rate of about 2% per hour. This
! is considered as a ground release. Dispersion factors (x/Q) were obtained by modifying the values found in NUREG/CR-3012 (IRDAM) to average local meteorological conditions.
15.5.1.2.3 Dose Conversion Factors Semi-infinite dose conversion factors (CF) were obtained from the , r , U.S. NRC Regulatory guide 1.109. These factors are listed in Table 15-2 along with the decay constants for the isotopes of interest. For thg igdine absorption caculations, an adult breathing rate . of 3.0 x 10- i a /see was assumed. Infant iodine doses can be obt the dose by 0.09 (infant breathing rate is 2.7 x 10-gingd m /sec). by nultiplying 15.5.1.2.4 Fuel Release Fractions , .i . It was conservatively assumed that 25% of the radioiodines and 100% of the noble gases were released from the failed fuel into the reactor , cell as recommended by the NRC through the ANSI /ANS-15.7 standard. 100%
! of the fuel was assumed to be failed for these calculations. This 100%
failure is a common assumption made by the NRC in the evaluation of the radiation doses for a Loss of Coolant Accident associated with commercial light water reactors. 15.5.2 Dose Calculations
; A computer code was written following the methodology in the l
previous sections and was then used to calculate the radiation doses for this hypothetical accident. Using the input data specified previously, the code would calculate the dose for each isotope separately for a specific time period, total up each individual dose contribution, and output a total dose versus distance for each time period. The time I periods investigated were: 0-2 hours, 2-24 hours, 1-4 days and 4-30 days. Distances investigated were 250-3000 meters. i 4 O , 232 1 l l 1
r ( Table-15-2 Radionuclide Decay Constants and Dose Conversion Factors Whole Body Thyroid Dose Decay Constant DoseConverajon Conversion Factor Radionuclide (sec-1) Factor C ec Kr-83M 1.035 x 10-4 1.045 x 10-2 ___ Kr-85M 4.297 x 10-5 3.770 x 10-2 _ _ _ Kr-87 1.520 x 10-4 3.437 x 10-2 ___ Kr-88 9.358 x 10-7 4.601 x 10-1 - Kr-89 3.655 x 10-3 5.156 x 10-1 - Xe-131M 6.689 x 10-7 4.180 x 10-2 __ Xe-133M 3.597 x 10-6 8.250 x 10-3 - Xa-133 1.512 x 10-6 7.500 x 10-3 --- l Xe-135 2.099 x 10-5 6.150 x 10-2 ____ Xe-135M 7.549 x 10-4 1.055 x 10-1 ---- Xe-138 8.133 x 10-4 7.075 x 10-1 ---- I-131 9.975 x 10-7 - 1.48 x 10 6 I-132 8.424 x 10-5 5.35 x 104 I-133 9.255 x 10-6 --- 4.00 x 10 5 I-134 2.196 x 10-4 - 2.50 x 10 4 I-136 1.444 x 10-2 ----
.1.24 x 105 15.5.3 Results of Dose Calculations A sensitivity analysis was performed for the whole body and the thyroid doses for periods of 2 hours, 24 hours, and 30 days. The distance was varied for. the VTAR operating at 100 kWeh and hypothetical 500 kWth maximum power. The results of these calculations are presented graphically in figures 15-5 to 15-12. Cases for mild conditions and !
severe conditions are shown for 100 kW and 500 kW maximum power for the thyroid -dose and severe conditions for the whole body dose for both power levels. t l l l 234 t. I
- - . . . - _ . _ _ -- . _ . = _ . . - . . - _ . _ - - . . . _ . - - .
- 4 4
- O ricuse is-s
. 2 HOUR WHOLE BODY DOSE 10 0 AND 500 KW SEVERE CONDITIONS i i i - I 1 ^ n . i i . ! OsO 500 KW
.01 eums i 100 KW i
0 01 , , . . . O 500 1000 1500 2000 2500 3000 DIS 1ANC E ( METERS) 235 l __. _ - _ - _ _ _ _ - - - - , _ _ _ - _ . . _ . _ _ . _ ._ ___--_.D
u ,,m..maamaw-.a- u- --,--_.,u . m -, ..- - - , - - -
. ' P si 4
)
\, \
(( \ , I , LQ FIGURE 15 - 6 24 HOUR WHOLE BODY DOSE 100 AND 500 KW
. SEVERE CONDITIONS o <
I
- 4 e
L 1, . .. l 500 KW 4 O O 2 - O ! ' ua 100 KW n 8 4 s
.01 I'
1 .
)
r
)
(
. .k
?
f O ooi . 50'O 1000 1500 2000 250 0 3000 DISTANCE ( METERS ) ~.r: 236
- , . -. . . , r , - , - - - . , - .
4 ,..---...,-,-,a . . , -- . _ , - - - - , - - - . . , . . . , _ _ , . , , . . - , , - - - . , . , , - . , - - , ,
O . FIGURE 15 - 7 i 30 DAY WHOLE BODY DOSE 100 AND 500 KW SEVERE' CONDITIONS , -1 f . Y p? I h
~
6 .
- t
- w
.e i: 8 500 KW iL O.
.i t
i 100 KW
\,
01 ((. 500 1000 1500 2000 2500 3000
;)' :
DISTANCE ( METERS ) O , s.
, 237 l 1
l l
o FIGURE 15 - 8 10 - . 2 HOUR THYROID DOSE 500 KW LOCAL METEOROLOGY e
=
i
, SEVERE INVERSION l.0 M
e l 1 e w l E
.I .
s Q m MILD LAPSE
.01 I
G O 0 01 Sbo 1000 1500 2500 2500 3dOO l
.O ' DISTANCE ( METERS ) I 238
I \ l LO i FIGURE 15-9 j
- 10. 0 -
2 HOUR THYROID DOSE
- 100 KW , LOCAL METEOROLOGY n
l.0 - SEVERE INVERSION l I w 0: w O 8
.I e
j MILD LAPSE < .01 1 l
.0 01 500 10'0 0 15 0 20UO 2500- 3000 OlSTANCE ( METERS )
239-l
. i
- - " a _ m a, _,
e o _ I
. FIGURE 15 -10 24 HOUR THYROID DOSE 10 0 KW to --
SEVERE INVERSION W W W W l J O 2 - i E w I ~~ co . 8 - MILD LAPSE e
.01 .
W un A
.0 01 , , , , , ,
500. 1000 1500 2000 2500 3000 DISTANCE ( METERS ) 240
se* 4 O riouas i5-ii 24 ~ HOUR THYROID DOSE 500 KW 10 0 SEVERE INVERSION 10 - O eu 6 N 2 O : 8 . WILD LAPSE
.I -
I h
.01 , ,, , , , ,
500 1000 1500 2000 2500 3000 O oi=Taace ( > Tras ) 241 p i l j
1 O l 1 FIGURE 15 - 12 30 DAY THYROID DOSE 100 AND 500 KW 1000 6 l O 10 0 - - 500 KW SEVERE INVERSION i n ^ 5 a w o 10 - Q l 00 KW SEVERE INVER$10N l i 4 1 500 1000 1500 2000 2500 3000 1 O DISTANCE (METERS) l l 242 ,
,,c~ ,- n, w ~ , , , , , , - ,e,-.- .,.,--w I
~bV. 15.5.4 Selection of Site Parameters Based on Dose Calculations i l
In regard to Research Reactor Site Evaluation, the ANSI /ANS 15.7- l 1977 standard presents the following definitions: (1) Site Boundary "The site boundary is that boundary, not necessarily having
! restrictive barriers, surrounding the operations boundary wherein the reactor administrator may directly initiate emergency activities. -The area within the site boundary may be frequented by people unaquainted with the reactor operation."
(2) Urban Boundary "The urban boundary means the nearest. boundary of a densely populated area or neighborhood containing population of such numbers or in such a location that a complete rapid evacuation l is difficult or cannot be accomplished within two hours using ] available resources." r The dose commitment for the " site boundary" is 5 Rea to the whole body or 15 Rem to any other organ for a two hour period. The dose commitment ! for the " urban boundary" is 0.5 Rem to the whole body or 1.5 Rem to any I other organ for a one day pariod.' This study indicates that the thyroid 4 dose limit is the critical dose. ] Due ~ to actual site conditions for the VTAR facility, the urban boundary distance, by its definition, can be estimated to be 790 meters (approximately 1/2 mile) ' which is the distance to the nearest housing. development. A development would be difficult to evacuate in comparison to a university building. This boundary would hold for both 100 and 500 kW maximum power levels. The " site boundary" for 100 kWth operation would be the immediate area around the facility, approximately 20 meters around the outside of the building. For 500 kWeh operation, the site boundary would be larger as expected. This boundary would be approximately 200 meters under-conditions of severe inversion. l . This study is done under the assumption of total fuel meltdown' or
~
complete loss of cladding integrity. These assumptions are unrealistic for the VTAR facility under any credible accident conditions with current safety requirements. Even at 500 kWeh maximum power, so many failures would have to occur at the same time that the probability of' this accident is - indeed very small. In addition, the VTAR f acility would not operate long enough to attain equilibrium radionuclide concentrations. On a . six hour per day operating cycle, the actual concentrations would be about one-half the concentrations used in these - calculations. Therefore, it'is' concluded that the study presented here l
-is very conservative;and unlikely to occur under any circumstances.
l 1 243 1 1
O 15.6 Earthquake As noted in Section 2.5, the VTAR facility is located in an area j which has had no maj or (damaging) seismic activity in the past century. The largest known quake in Virginia reached a magnitude of about 5.8. The structure of Robeson Hall, the building in which the reactor is housed, is capable of withstanding an earthquake of magnitude 6.5 without sustaining major structural damage. The VTAR is equipped with a seismic motion sensor which scrams the reactor automatically upon sensing .a lateral ~ motion of 0.14g, which corresponds to approximately 3.0 on the Richter scale. According to the Cort study (NUREG/CR-2198), the worst case earthquake-type accident in an Argonaut reactor is one in which coolant is lost but the fuel itself is uncrushed. For all intents and purposes, i then, the worst case earthquake accident becomes a loss of coolant accident, which is addressed in Section 15.4. Essentially, subsequent studies. by the reactor staff indicate that even at 500 kW operating power levels, a loss of coolant accident similar to that which may be , caused by a large earthquake would produce some fuel melting from decay
- heat only af ter two to four hours with no cooling and no heat tr-'isfer i from the core.
History would tend to indicate that the probability of damage to O the VTAR from earthquakes is small, though not non-existent. damage would allow the reactor staff some four hours to re-establish Such some type of cooling in the core to remove 'a sufficient amount of decay heat to prevent melting. If the reactor is not operating at the ti'ne of , the damaging earthquake, there is no danger of a fuel melt and consequently the' danger to the public is minimized. 15.7 Fuel Handling Accident
, Since the VTAR reactor is de-fueled and inspected bi-annually, the i possibility of a fuel handling accident cannot be ruled out. Because j the fuel elements are transferred singularly, a dropped fuel element
- scenario would be the most likely case to consider. This scenario has 4
been addressed previously in NUREG/CR-2079 and will be used exclusively in application to the VPI&SU reactor. S l The first step would be to determine the fraction of total core inventory in a single fuel element. This can be calculated from the following: f nr* . f= (1)
" total O !
l 244
4 jL where n, = number of fuel plates in the element j r = power density ratio of fuel element location to average in the core - i i n 1 = total number of fuel plates in the reactor 1 -AppSylngvaluesfortheVTAR i 4 f=
)= 0.125, or 12.5% of the core activity in a single j element.
NOTE: See Reference 15.12 for a basis of the r value selection. l Next, the activity release fraction would have to be determined. This requires determination of the total' (active fuel) surface area for
- j. one fuel element. For one plate:
Area = Length x Width I f = (2.75 in) (23 in) l l _.
= 63.25 in2 x 2.(both sides of the plate)
= 126.5 in 2 j therefore, one element yields:
A = (126.5 in )(12)(2.54incm)2 f A = 3,855.72 cm 2 } j NOTE: Dimensions are from American Standard drawings. i To calculate the volume.of exposed fuel: l 2 ! V = (1.37 x 10~3cm)(3855.7 cm ) l
= 5.28 cm3 l
where 1.37 x 10-3 cm is the range of fission fragment recoils. See
- Reference 15.12.for a basis of value selection.
j i Next, determine the total fuel volume for 1 element. For one plate:
- V = (2.75.in)(0.04 in)(23 in)
I
= 2.53'in 3
- .O
~ 245
1 0 1 Hence, for one element 3 V = (12)(2.53 in )(2.54 cm)3
= 497.5 cm3 Again, dimensions are from American Standard drawings. Finally, to obtain the activity release fraction:
V 3 exposed 5.28 cm 3
= 0.0106 (3) ,
available 497.5 cm or 1.06% of the gaseous fission product activity could escape. The last step is to take the values calculated in Reference 15.12 for hypothetical activity, curies released, and dose to a downwind observer and apply a fractional equivalent to obtain estimates for the VTAR. This fractional equivalent is determined as follows:
^ '
1 = 0.24 FRE = (4) (RI) (RI) where FRE = total fractional release equivalent for VTAR i f = fraction of core inventory for a fuel element in VTAR reactor, 0.125
= fraction of core inventory for a fuel element. 0.07 f(RI) 1/3 = fraction of plausible operating time for VTAR (8 hours per day at 100 kW for 1 year) as opposed to values used in Reference 15.12 (24 hours per day at 100 kW for 1 year).
A = activity release fraction for VTAR, 0.0106
= activity release fraction in reference 15.12, 0.0106 A(RI)
Multiplication of the values in Reference 15.12 by this fraction are summarized in Table 15-3. Conclusions reached are that the maximum dose to a downwind, ground level observer in the event of a fuel handling accident would be 0.3 Rem to the whole body (f rom noble gases) and 6.21 Rem to the thyroid from radioiodines. O i 246
Table 15-3 Activity and Dose Equivalents from Maximum Credible Fuel Handling Accident for the VTAR Curies Plume Dose Nuclide Released Concentration (C1/g3 ) Equivalent (Rem) ; KR-85" 1.25 2.5x10~0 0.02 KR-35 .02 4.9x10-3 _ KR-87 2.26 6.5x10-6 0.15 KR-88 3.45 9.5x10-6 _ XE-1335 .18 5.0x10-7 - XE-133 6.43 1.8x10-5 0.03 I XE-135" 1.02 2.8x10 4 - XE-135 7.39 1.8x10-5 0.10 { 0.30 Rem Total 4 Whole Body Dose Equivalent 1-131 2.62 7.2x10-6 3.10 I-132 3.93 1 1x10-5 0.18 i 1-133 6.43 1.8x10-5 2.16 i l I-134 6.79 1.9x10-5 0.15 I-135 6.07 1.5x10-5 0.62
- 6.21 Ram Total i
Equivalent Thyroid Dose Numerous assumptions and items to note are listed in Table 15-4. O 247
, l i
4 Table 15-4 Assumptions and Notes for VTAR Fuel Handling Accident l 1. Removal of the fuel element from the core immediately following full , j power operation. t l 2. All gaseous activity within range of recoil particles (1.37 x l 10 cm) would escape. 3
- 3. Entire surface of all 12 plates removed (0.02 in cladding).
!- 4. 100% release of fission gases.
- 5. Uniform distribution of fission gases.
{
, 6. Radioactive decay disregarded.
. 7. No credit for any anchanisms which would reduce the maximum exposure j (containment disregarded, no sutomatic reactor ventilation trip or dilution, X/Q value (10 s/n 3)(extremely conservative), no radio-
-active plateout, etc.)
! 8. Assuess I hour release - in the Reference 15.12 derivation (of dose l O 2tvtc t>=* are applicable to an instantaneous or protracted release. ei4 =a=*ae t- ,i
- 9. Observer at ground level - but VTAR reactor room ventilation
! discharge is on building roof, which is accessable only to staff ! personnel.
, 10. The only less conservative assumption made was 1/3 of the operating i j tian used in Reference 15.12. ! 1 , j 11. These values 'shall be incorporated into the Energency Plan Procedures as the Protection Action Guide. 15.8 Storage Pit Criticality Calculation i Por safety purposes a simple one group criticality calculation was made to determine k,gg in the event the pits are accidentally flooded. In the event of a secondary cooling system rupture there is a likelihood that the storage pits could. become flooded. . If this occurs the pits will be evacuated, the water sampled for activity prior to . release, and the fuel inspected for damage. I l If corrosion is detected, fuel shall not be used until an approved QA method of verifying integrity is employed. O I 248 ,
-.r , , - - , - . - - . - . - . ..-~-----.sn--- . , ,e ,n,._,e.,--_w-.- ,-m,---,-..-v- ,. ,. --,,.r- v-+. .n.. ,..
. - - .... .. . - . - - .... - . - - = - .
i i O '*- <t =- ett computed using the one group criticality equation. tei> tic et - < cc < *- 11 44 - k ,e l 1+LB i where k, = nf since c = p = 1
! The storage pit was considered to be a 4 x 4 array of cells spaced ! 13" apart. Each cell consisted of one fuel element in an eight inch diarseter water filled cylinder in the center of a 13" square of ordinary I concrete. Various cell parameters are shown in Table 15-5.
! The calculations yield an f of 0.310 and .n of 2.01 to give k " = 0.623. For a 16 element storage pit of effective dimensions 26.6 x 52 x 52 inches, by using an age of 35.6 cm, a thermal diffusion length of 4.07 cm and a reflector savings of 5.2 cm.
-TB i
e = 0.923 O and 1+LB = 1.029 t I' Then k,gg = * = 0.5158
.02 i
i Therefore the storage pit can be expected to remain sub-critical when i flooded. 4 I i O 249
. - . - - , ~ , ,4.ww,, .-,.-,w- ,4 e ,.e , .c...pm..- -,,,.,,.,.m,-.-.,-...-.v..m.. .,,- - ~..
Table 15-5 $ Parameters for Multiplication Calculations of UTR-10 fuel storage pit. }
'(12 plate elements) th g th th th f Material Fraction I g g Dg U-235 .00023 28.51 24.10 .497 .670 0 .67 i
- U-238 .000015 .117 0 .497 .670 0 .67 Al .0168 .0123 0 .084 3.97 0 1.41 i
Fe .0280 .202 0 2.07 .161 Neglect HO 2
.266 .0197 0 2.08 .160 .036 1.19
- Concrete .704 .00596 0 .798 .418 .0512 1.81 i
l lO ) i Data Given: j U-235 loading per place = 22 g Enrichment = 90% Meat thickness = 0.040 in ] { Cladding thickness = 0.020 in l Water gap = 0.40 in
! Rectangular steel guide = 0.089 in thick x 5-7/8 in x 3-1/16 in i
! Steel liner = 1/8 in thick x 8 in diameter
- 16 cells set in concrete on 13 in centers
- Calculation
i : U-235 loading / assembly 12 x 22 = 264 g ( Meat thickness / assembly 12 x 0.040 x 2.54 = 1.2192 cm l O l [ 250
. . - . _ _ _ . . . . _ _ _ . _ . _ _ _ . _ _ . . . _ . _ -.. . . . ~ . = _ __, _ _ _ , _ _ . . , __.
O V Cladding thickness / assembly 24 x 0.020 x 2.54 = 1.2192 cm Water gap thickness / assembly 11 x 0.40 x 2.54 = 11.176 cm Total thickness 1.2192 + 1.2192 + 11.176 = 13.6144 cm Volume fractions in cell (cross section of element) l* Meat y3 4
- 0.08955 l
Cladding = 0.08955 6 Water 44
= 0.82089 l Meat Composition i Volume of meat / plate - 0.040 x 24 x 2.75 x 2.543 = 43.262 cs3 Assume 93% enrichment, then total U per plate is:
0 93 = 23.66 g of which 1.66 g is U-238 Density ((0.93)(235)3
+ 0.07] 18.7 - 18.48 g/cm
! Volume of U = = 1.28 cm Volume of Al in meat - 43.26 - 1.28 - 41.98 cm l Volume fractions in meat o=;;28 , = 0.02,58 l A1 = = 0.97042 Volume fractions of materials in cells excluding steel and concrete 4 (Cross Section of Element) U-235 = 0.93 x 0.02958 x 0.08955 = 0.002463 cm2 U-238 = 0.07 x 0.02958 x 0.08955 = 0.000185 cm 2
= (0.97042 x 0.08955) + 0.08955 = 0.176450 cm 2 A1 H2O = 0.820892 cm2 Steel (Cross Section)
Guida l O 4 251
i i 2 . 0.089(2(5.87+3.0625)]-2.54 = 10.26 cm 2
- Liner 0.125 x 8 x x 2.54 20.27 cm Total Steel - 30.53 cm Concrete (13 x 2.54)2 - (4.00 x 2.54)2 3= 766 cm 2 Water (4 x 2.54)2 w - 12 x 0.080 x 3 x 2.542 = 306 cm2 (Plates are
- 3" wide)
Cross Section of cell = (13 x 2.54)2 = 1090.3 cm2 2 Cross Section of element = 13.61 (3 x 2.54) = 103.7 cm Volume fractions in entire cell I U-235 = 0.00246 x 103.7/1090.3 = 0.00023 U-238 = 0.00018 x 103.7/1090.3 = 0.000017 j A1 = 0.1765 x 103.7/1090.3 = 0.0168 i ' Steel = 30.53/1090.3 = 0.0280 O H2 O =-[306 - 103.7 + (.821 x 103.7)]/1090.3 = 0.2660 Concrete = 766/1090.3 = 0.7040 i i E Y Y 25 25 E Y V
- 25 25 + E28 + Y28 + EA1 A1
~
24.1 x 2.30 x 10 ' x 2.46 0.01364 0.00663 " 2.30 x 10-k x 28.51 + 1.5 x 10-5 x .117 + 0.168 x .0123 E E Y j g, 25 25 28 28
- EA1 A6 All
,l 0.00863 0.00863 I " 0.00863 + 0.0280 x .202 + .266 x-.0197 + .704 x .00596" 0.02372 = 0.364 nf = 0.364 x 1.580 = 0.575 252-1
- , - , . ,m.- .....,n.-- - - , . , . - - . ,- - , - - , ---n...-.
O f E 31
= 0.036 x 0.266 + 0.0512 x 0.704 = 0.0456 ca" 0.266 , 704 ~
=
,g = 0.244 + 0.389 = 0.613 ca d
Dg .= 1.63 cm f j T= = 1.63/0.0456 = 35.8 cm S1 D 2 eh L = g a 1 0.704 0.266 -1 i D " 0.413 ,' O.160 + 0.0280 0.161 + 0.0168 3.97 = 3.5460 cm th O Dg = .2820 cm i E,th = 0.0218; L2,0 8 9 28 = 12.9400 ca
~
i I D Reflector Savings = 6 = [ L r r 6= ( ) = 0.6740 x 7.70 = 5.2 cm } pie dimensions are 30.6" x $2" x 52" or 77.7 cm x 132 cm x 132 cm 2 -2 B l
* (88 10) + 2(142.40) = 0.00127 + 0.00097 = 0.00225 cm
. O ; 253 I
6 e -Bb=exp(-35.8x0.00225)=0.9230 1+La = 1.0 + (0.00225 x 12.90) = 1.0290 K = 0.9230 x 0.5750 = 0.5158~ ' < eff 1.0290 Similar calculations for 13 plate elements yield k,gg = 0.6030 (see Appendix 15-I) 15.9 Explosive Chemical Reactions In reference 15.12, " Analysis of Credible Accidents for Argonaut Reactors" (NUREG/CR-2079), it was thought that the possibility could
, exist for an explosive reaction between aluminum and water under adverse conditions. The reaction is:
2A1 + 3H2O + A1 02 3 + 3H2 + 389 kcal O For the reaction to occur, the aluminum would have to be either in a molten state or exist as finely divided particles. No credible accident mechanism could be postulated to produce the conditions necessary for melting or production of finely divided aluminum particulates. Although the aluminum-water reaction is exothermic, heat is
, required to initiate the reaction. Studies indicate that 174 cal /gm of fuel plate are required before damage is apparent. This is far in excess of energy available from an inadvertant maximum reactivity insertion. Hence, this is not considered a credible accident. ! The only other possibility for an explosive reaction would be through samples inserted into the central grappite region for irradiation. However, samples are kept small (< 1 cm ) and irradiation of explosives or highly corrosive material is expressly forbidden in the Technical Specifications. Therefore, this is not considered a problem. The possibility of a sample causing a fire in the graphite region is discussed in section 15.11.
i 15.10 Experimental / Rabbit System Failures , 1 A failure of the rabbit system is in itself not detrimental to the reactor. However, a rabbit system failure could lead to possible radiological hazards. In order to minimize the hazards involved with a O 254
d E 4 2 h-d potential system failure, the maximum dose rate of an irradiated sample
, has been set as 2R/hr at 18 inches. This dose rate and consequent
- maximum irradiation time are determined either by calculation or by a three second irradiation measurement extrapolated to the maximum allowable irradiation time without exceeding 2R/hr.
There are three types of f ailures of the rabbit system which can ; lead to a radiological hazard. They are:
- 1) a sample lodged in the reactor
- 2) a sample lodged between the rabbit receiver and the reactor l
- 3) ruptured or leaking sample capsule '
The following sections describe the methods used to deal with these 4 failures. 15.10.1 Sample lodged in the Reactor [. I If a sample does not return to the rabbit receiver and the " sample
! in" indicator does not de-unergize, it is assumed ' that the sample is ;
- lodged within the reactor. If the sample has not exceeded the maximum allowable irradiation time, an attempt can be made to dislodge the sample by alternately applying " insert" and " return" gas pressure. If ,
j several tries do not free the sample, then the reactor is shut down and j the sample is removed manually. Should the maximum allowable time be , j exceeded, the reactor will be shut down immediately. Any sample removed manually would be treated as highly radioactive using standard radiological control techniques. 15.10.2 Sample lodged Between Rabbit Receiver and Reactor
- If a sample doas not return to the receiver and the " sample in" indicator de-energizes, it is assumed that the sample is lodged in the transfer tubing. The rabbit operator will immediately verify (visually) i that the transfer capsule is actually lodged in the tubing. If
! verification cannot~ be made readily, the sample would be treated as lodged in the reactor. ,1f the capsule is ' lodged in the tubing, attempts can be made to dislodge it by applying alternate insert and return pressure or by manually shaking the tube. If, after reasonable l attempts, the capsule remains stuck, the reactor would be shutdown and the capsule removed manually, employing standard radiological handling
- techniques.
15.10.3 Ruptured or Leaking Sample Capsule Samples ' irradiated in the pneumatic transfer system are encapsulated before' insertion into the rabbit capsule. Normally this internal capsule is a sealed polyethylene vial. However, other O , 1 l 255
.. - - .-. . . - . - - . . - - . - . - . _~ - - . - - - .- --
4 I I containers, such as a quarts vial, have been used in the past. Ruptures of the polyethylene capsules could occur due to expansion of the ' - encapsulated material, a poorly sealed capsule or, in the case of an
- . alternative enclosure, by physical breakage of the sample vial.
e The operators of the rabbit systems are given specific training in the event any evidence of loss of integrity of the samples is 4 observed. The receiving station is immediately closed with the rabbit sample inside. The Reactor Radiation Safety Officer is immediately notified and decontamination is performed so as to protect the workers ) and the general public. l 1 i There have been several instances in which potential risks to
; personnel. could have occurred. One was the result of a broken quartz i vial containing a bromine compound. An inexperienced graduate student from another department attempted to retrieve the sample. Moderate contamination of the immediate area of the receiving station occurred
- but no one received any significant exposure. A more serious incident l occurred in 197 en, due to a lack of calculation, an approximately.1 l gram sample of was irradiated for 3 minutes instead of 5 seconds.
j The capsule ruptured because of high temperatures. Significant 4 contamination of the pneumatic system, room 106, the adjacent hall and i an area near the wall on the outside of the building occurred. No one
- . was injured and complete decontamination was effected in approximately 3 1 working days. Physical changes to the equipment and much more rigorous i O t at ci ca ' tt at
- ecia = t =ici or
- the incident unlikely. Specifically, all irradiations of special l nuclear material must now be cleared in advance by the RSC. Further, all samples receive a preliminary 3 second irradiation to allow
! estimation of a maximum safe irradiation time. Procedural rules require j an immediate shutdown of the Reactor should this irradiation time or the i requested irradiation time (if shorter) occurs. 15.11 Fires 1 i 15.11 1 Introduction { A self-sustaining graphite fire requires oxygen, high temperature j and an ignition source to initiate combustion. The fire may be j suppressed by eliminating fuel or oxygen or by reducing the temperature i of the reactants below the combustion value. The specific actions to be i taken in the event of a fire are covered in detail in the Emergency ' } Plan. A brief analysis of credible sources of oxygen, high -temperature j and ignition will be considered in this section. Intentional sabotage and its effects are beyond the scope of this section. 15.11.2 Oxnen Sources It would be possible to have access to both the graphite on top of. i the reactor and in the thermal column. These locations are reached by O 256, d
~- . , - - , _ . _ _ - - _ _ . - - . - - . . - _ . . , , . . , . _ . . _ , _ , - . . . . - . . , , . _ . _ . . _ - . . , . . - , , , , , - , - - , _ _
1 J O removing the movable concrete shield blocks. A substantial part of the j total reactor inventory of graphite would then be exposed to room air and ambient conditions, though Tech Specs limit core exposure times and conditions. In the VTAR facility, air passing through the reactor core is diluted with room air, which is then drawn f ros the reactor room and through a roof vent by two ventilating fans at a rate of 2117 cfa. Air i flow through the graphite stringers is estimated at 250 cfm (per NURgG/CR-2079). At this flow rate, 2.1 kg 02 per minute are potentially l available to the graphite. This amount of 02 could completely oxidize t l 800 g of carbon tg CO 2 per minute (assuming complete combustion), i yielding 6.2 x 10 cal / min based on 7800 cal /g for the heat of j combustion of graphite. This energy would be sufficient to raise 12 kg l of graphite to 6500 C, assuming that no energy is carried off by the air
- flow. If the fan is turned off, as it is when the stack monitor reaches a higher than normal (15 mR/hr) level, the rate of 02 delivery will fall
} considerably. In this condition, a rapid drop in graphite temperature j would result probably leading to the graphite temperature falling below ! the ignition temperature. s
- Failed beam or rabbit tubes would allow only a slight increase in i air flow with the fan in operation. Again, with the fan shut off, the increased air flow through such failed tubes would be negligible and graphite temperature would be expected to fall rapidly.
Technical specifications for experiments introduced into the
- reactor effectively prohibit the introduction of significant amounts of
, oxidizing materials. Because of administration procedures required for , . new experiments, the possibility of a problem in this area, though not
- non-existent, is considered very small.
1 l 15.11.3 Fuel Sources l All classes of flammable materials - solide, liquids, and gases - i though potentially available are very restricted in the reactor area. I Deliberate introduction of flammables into the reactor is for
- experimental purposes only and the nature of the materials is reviewed and severely restricted.
f The most obvious source of fuel for a reactor fire is the j graphite. Graphite. when made to burn, yields 7800 cal /g in the
- conversion to CO 2. The energy released in the combustion of Ig of I graphite will raise 38g to the ignition point if ' no energy is lost.
j Then the actual energy delivered would be obviously less, but once A ignited it will continue to burn until the oxygen supply is diminished j or the fire is extinguished. I Various chemical mixtures have the potential for reacting ~and j releasing considerable energy. Regardless of the nature of such O 257
. ~. - _ _ . - _ __ . _ .~ _ . - _ - _ _ _ - - _ _ , . . _ , _ . _ _ , . _ - - _ _ _ _ . , _ - . _ . - - . . - . - - _ .
i materials, however, their introduction into the reactor is severely l
- limited and closely monitored. Though the potential for such accidents
- cannot be neglected, a combination of conditions and events mast occur i for this potential to be realised. In addition, there are some steps
- between the introduction of such materials into the reactor and actually reaching ignition by further aischance.
j 15.11.4 Innition Sources Many events can be postulated to lead to the' production of ignition i temperatures. -To pose a significant problem, the ignition event must occur at the location of and in the presence of a material which can f'i readily ignite and can also burn long enough to ignite a larger fuel ; 1 supply. Some ignition sources are sufficiently energetic to potentially j start the graphite burning directly. One of these is an electrical ! malfunction. !- Detectors and other instrumentation requiring high voltage may be j installed in the graphite penetrations. Current drain would have to j exceed the circuit breaker capacity before the are would be j terminated. While such an arc would not be likely to ignite the ! graphite, it could conceivably ignite insulation or other flammable , material. , Maintenance in the reactor room, such as welding, as well as a j building fire, could cause ignition of flannable asterials. h only j real possibility of such an ignition source igniting the graphite is to j have burning material fall into an opening in the shield, then provide a
- sufficient oxygen flow to produce and sustain a graphite fire.
1 The Wigner Effect is another possible ignition source. According
! to NURgG/CR 2079,1ghe ratg of energy storage for reactor graphite at
! 500C is 4.7 x 10- cal /cm 3 The highest flux in the VTAR occurs in l the graphite in ral portion of the core. The flux at d poweg is 1.2 x 10go n/cm cong-s, and at 500 kW, is expected to be 6.0 n/cm -s. Thus the grap in the center island will store energy at ' approximately5.64x10pitecal/g-s, or 0.049 cal /g-day at 100 kW. At 500 i j kW, energy deposition will be appronistely 2.82 x 10- cal /g-s, or 0.244 j cal /g-day. Assuming an intentionally _ overestimated capacity' factor of j 15% since the present graphite was installed in thg 1967 upgrade to 100
- kW, total operation would correspond ' to 8.76 x 10 kW days. For this L reactor, then, the center island graphite may have accumuisted a stored i energy of 43 cal /g. This energy, if released, is insufficient to raise 4
the temperature of the graphite more than 570C, which is considered j trivial. 15.12 Due Valve Failures i Fotential dump valve failures any be divided into two general l categories: Lelectrical and mechanical. These two categories will be f addressed.in this section. 258 9 '-g_,1+ -
-94,y,-gre =-+g --g --w-s( tw --
**e---,-"1iir-ge.wy*Nm N ye em tv-.g e ww - g-- m ryy p--*,feyygi, r-,t-y ygy,eyw wy g g m ,+ wiyw w w y-qp- + 4
15.12.1 Electrical Faults There are two separate electrical systems which could experience faults. These are the dump valve motor and the dump valve f electromagnetic drive clutch. ! 15.12.1.1 Dump Valve Motor Failure i t In the event of a motor failure with the value open, the value could not be closed. Since the dump valve must be closed in order to start up the reactor, the situation is a relatively simple repair problem. . Motor failure during 100 kW operation would not cause a cooling ' problem, since the valve would already be closed. The inability to close the dump valve af ter shutdown from 100 kW operation is still no more than a maintenance problem because decay heat is insufficient at 100 kW to cause temperatures which would cause fuel element f ailure. If 500 kWeh operation is to be realized, alterations must be made to the dump valve to allow manual closure of the dump valve in the event of the occurrence of a scenario similar to that described in section ! 6.3, loss of primary flow. 5 l 15.12.1.2 Clutch Failure [ A dump valve clutch failure at 100 kW operation will result in an
- iemediate reactor shutdown because of the reactivity margin provided by ;
j dumping the water (-30% Ak/k). The reason' for this shutdown is that [ l clutch failure results in the dump valve opening under spring tension. l 1 Since draining the core by opening the dump valve is a part of normal : i shutdown at 100 kW ope ration, it is evident that core temperatures ! I during loss of coolant at 100 kW operation do not result in damage to f j the fuel assemblies. Again, during 500 kW operations, this situation is i considered a LOCA, as addressed in section 15.4. - I j If a clutch failure occurs when the reactor is shutdown, the i situation is similar to that of a motor failure in that the dump valve - j is unable to close. I 15.12.2 Mechanical Failure j Because the valve has a stainless steel disc and seat and an i aluminum valve body, rapid temperature changes any cause the valve to j bind or come off its closed seat. l I
- 15.12.2.1 Duas Valve Stuck open If the dump valve should come off its seat, several resultant indications are available to the operator. The limit switch indicating j O i l
l 259
-. - - _ . . . - - . -. - _ . _ - . . - - . ~ - - . . - - - - _ -
t i i. O " dump valve open" requires approximately one quarter inch movement of the stem. If the movement is insuf ficient to cause the limit switch to indicate the valve in the open position, two other indications will j i inform the operator of this problem.-
- t With the dump valve off its closed seat, the core is partially i robbed of flow, or short-cycled. In this case, the hot leg temperature, i and the resulting temperature differential across the core, will j increase. This increase would continue until either the operator !
.i observes the abnormal AT or the primary coolant high temperature alara occurs. Since the flow detector is upstresa of the point where the flow li is forced to the core tanks, a partially open dump valve would not cause a reduction in flow indication. t ' The second indication available to the operator would be an increased reading on the primary coolant fission product monitor. The ! reason for this increase is that at the lower flow rate in the core, the coolant would remain in the core for a longer time, resulting in l increased N-16 production. l [ If any of the above indications occur - abnormal AT, high f temperature alarm, or increased primary coolant radiation - the operator l i aust be relied upon to shutdown the reactor. , i 15.12.2.2 Dump Valve Stuck in the Closed Position j Since either safety rod or the shia rod being fully inserted will i 1 easily meet the required shutdown criterion, sticking of the dump valve l in the closed position does not cause a "can't scram" condition. t ! Because a situation involving all rods stuck out and the dump valve stuck closed could conceivably occur, the system is so designed that the f core tanks may be drained by opening a single valve (by hand) in the 1 ! process pit. At 100 kW, this operation is not hasardous and no fuel j damage would result. Above normal activity levels would be expected in the primary system, but the amount would be expected to be small and short lived. ! I In the " stuck closed" situation, once the reactor is shutdown the dump tank heater is used to raise the primary coolant temperature to or l l above that which occurred at the time the valve was closed. When ! !. sufficient time has been allowed for the various valve parts to reach ! thermal equilibrium, the manual scram is activated in an attempt to open , ! the stuck valve. If this procedure is unsuccessful, a small amount of , ! force is applied to the valve stem to f ree the valve. If the valve j still remains stuck, the core is then drained annually and force is again applied to the valve. If this attempt fails, the valve is then t disasser. bled. i O 260 i i'
Following a stuck dump valve, leak rate tests are performed to determine if the seat was galled and to verify normal subsequent operation of the dump valve. 15.13 Secondary Coolant System Rupture Secondary coolant systen rupture is by far the most likely source of internal flooding at the reactor f acility. No other single source, 1 with the exception of city water supply lines, contains the amount of water which would be required to flood the reactor room. A " water wall" of sealed concrete blocks, eight inches high, is installed around the area containing the secondary system piping, pump,
- and primary to secondary heat exchanger. This wall will, in event of a i rupture, allow the water to spread only in the direction of the process pit. The secondary system contains 3800 gallons of water. The process
- pit can contain up to 2900 gallons. In event of a secondary rupture in
- the reactor room, only the remaining 900 gallone would be available to flood the remainder of reactor room. Present plans are to add a
. partially removable eight-inch wall to the south edge of the process i pit. This addition would provide a near-complete confinement of water l to the area of the process pit and secondary system. With this 1 arrangement, the trapped volume would increase to about 4100 gallons, I which is sufficient to contain the entire volume of the secondary ( system. The only immediate concerne to the areas outside this containing dike are seepage through and under the dike and splashes in the area of such a secondary rupture. [ By containing the large volume of water in this manner, the rupture ! becomes a clean up probles of much smaller severity and the staff is ] afforded additonal time in which to determine the appropriate response to the problem. i In addition, should the reactor roca become flooded for any reason, l extensive wetting of the reactor graphite results in a negative } reactiv.ity insertion, tending to shut the reactor down. This phenomenon
; was shown in 1976 when the core tanks overflowed, dumping water into the l graphite. The reactor was operating at the time of the event and power dropped almost immediately to zero. After drying the graphite by l
heating at very low power levels, no additional problems were found and l the reactor was returned to normal operation. 1
- 15.14 Leaks and soitte Leaks and spills in the reactor area are considered here in a
- single section because both present similar problems and similar j solutions.
I in event of a leak or spill, the reactor would be shut down untti i such time as the Radiation Safety Officer and the Reactor Supervisor !O ! 261 j I i
~!
i
.O agree that the situation no longer exists. Any leak or spill in the I area will be treated as radioactive until proper determination of the radiation level, if any, is made by Radiation Safety personnel.
- Immediately following the detection of a leak or spill, the
; Radiation Safety Officer will be contacted. Radiation Safety personnel
! will then determine the radiation levels and supervise cleanup. If a i leak is determined to exist, Radiation Safety personnel will also l supervise the repair of the leak. If the leak or spill is determined to i be non-radioactive, clean-up will consist of simply a normal " sop-up". l If radioactive, the asthod of clean ,sp and disposal of the active l material and tools will be determined by the Radiation Safety personnel 4 present in cooperation with the Reactor Supervisor and staff. . Since the primary and secondary coolant systems are newly installed !
! and tested and since past samples of the primary system have shown radiation levels ranging from very low to none, coolant system leaks, in l ,
themselves are not expected to pose a radiation problem. l i j 15.15 Strikes by Personnel } The town of Blacksburg is supported primarily' by the University. t
- There has never been a strike or f riction between labor and management i of any severity at the University. The possibility of a labor strike i cannot, however, be ruled out.
Strikes effecting the VTAR facility may be of three forms. i l (1) A strike involving the Campus Police force }' (2) A strike involving the University faculty and staff (3) A strike involving Campus Police, faculty, and staff. In the event of a strike involving the Campus Police, the primary i security force becomes the Virginia State Police, who would provide patrols of the campus and reactor facility. The not result is that there should be no interruption of the availability of a security force ;
- should the need for such a force arise. In the event that the l requirements of the security plan cannot be met, a special meeting will j be called by the reactor supervisor to determine the necessary course of 1 action.
A strike involving the University faculty and staff is unlikely to , occur without warning. In the event of such a strike, the reactor would j be shut down and the control system disabled in such a way- that the j reactor could not be operated in any way until the control system is j restored to operating condition. O i i 262
. _ . _ _ _ _ _ . - _ _ _ . _ _ _ _ _ . . _ - . _ _ _ _ . _ _ _ _ _ _ _ . _ . . . . _ _ . ~ _ _ _ . _ . _ _ _ _ _ . _ . _ _ _ _ . . _ _ _ - . - _ _ _ - _ _ -
4 - l 4 .:" J l l i , i a I In event of a strike involving faculty, staf f, and campus police, I the reactor would again be shutdown and the control system disabled to , i prevent operation. 4' j i 1-4- t ! i
- i i
4 4 } i ( i ' t i ( 4
; O 5
4 6
?
i }
) I 1
f [ I I l 1 i l l 1, > I lO 4 i 263 ' i
r-a r v s' ~ O r Appendix 15-I 13 Plate Storage Pit Criticality Calculations ig l' Table 15A-1
,< All parameters for maltiplication calculations of UTR-10 fuel storage pit. (13 Plate Elements) ,
Eh th th g th e Material Fraction I,Eh Ig I D Dg U 235 0.000254 28.51 24.10 0.497 0.670 0 0.67 U 238 0.000019 0.117 - 0.497 0.670 0 0.67 , A1 0.0182 0.0123 - 0.084 3.97 0 1.41 Fe 0.0280 'O.202 - 2.07 0.161 Neglect------ u .l
',N, HO 2 0.2605 '0.0197 -
2.08 0.160 0.036 1.19 s, Concrete'0.7026 , 0.00596 -- 0.798 0.418 0.0512 0.81 O s ke.2i i 2 2 = 1 (at criticality)
, 1+LB Assuming: c=p=1 i o Then given: O, .,'
Water gap,. =0.36[in
,t Heat thickna e = 0.040 in i.
Clad thickness = 0.020 in W Calculation: , s s' i. U-235, loading /assy 13 x 22 = 286 g r) st tfcat thickness /assy: '13 x 0.040 x 2.54 - 1.3208 cm Clad thicko ms/assy 26 x 0.020 x 2.54 = 1.3208 en t Watergspthickbess/assy: 12 x 0.36 x 2.54 = 10.9728 cm Total thickness = 1.3208 + 1.3208 + 1.3208 + 10.9728 = 13.6154 cm T! . m , s f s o. .~' u \
,,. 264
,3
;L
f 4 O Volume fractions in cell (cross section of element) Meat
= 0.0970 36 4 Clad: 0.0970 p
Water: = 0.8059
; 1 6 4 l Meat composition:
Vol. of meat / plate: 0.040 x 24 x 2.75 x 2.543 = 43.262 cm3 238 Total U per plate: 3 = 23.66 gm, 1.66 g U Density (0.93( ) + .07] 18.7 - 18.48 ga/cm 6 .[ 5 Volume U = 8
= 1.28 cm 3
Volume of A1 = 43.262 - 1.28 = 41.98 cm >r Volume fractions are meat:
= 0.02958 ff U=4 6 O ^1 - 121-o97o*2 Volume f ractions of materials in cells (excluding steel and concrete)
(Cross sect. ion of element) 235 U = 0.93 x 0.02958 x 0.0970 = 0.002668 cm 38 U = 0.07 x 0.02958 x 0.0970 = 0.0002008 cm A1 = (0.97042 x 0.0970) + 0.0970 = 0.19113 cm H2O = 0.8059 cm2 Steel (cross section) Total steel = 30.53 cm2 Concrete: 766 cm 2 I Water: (4 x 2.54)2n - [13(0.080)3(2.54)2] = 304.1638 cm 2 Cross section of cell = (13 x 1.54)2 = 1090.3 cm 2-
+
Cross section of element = 13.6154(3 x 2.54) = 103.75 cm A >
. kl ,, ,
265 J' .
O Volume fractions of entire cell: U235 = (.002668 x 103.75)/1090.3 = 0.0002539 U 238 = (0.0002008 x 103.75)/1090.3 = 0.0000191 i A1 = (0.19113 x 103.75)/1090.3 = 0.01819 Steel = 30.53/1090.3 = 0.02800 Water = [304.164 - 103.75 + (0.8059 x 103.75)]/1090.3 = 0.26050 4 Concrete = 766/1090.3 = 0.7026 25fY25" E Y V t25 25 + E28 28
- EA1 A1
~
24.1x2.54x10 x2.46
~
(2.54x10 x28.51)+(1.9x10" x0.117)+(1.82x10 x.0123)
~
1.50x10 7.467x10
-3 = h91H - n g, Et25 25 + 28 28 + EA1 Al all f,
7.4676x10~ , 7.46x10-3+(0.028x.202)+(.2605x0.0197)+(0.7026x0.00596)
~
7.4676x10
-2
= 0.3327
======
=f 2.2443x10 K , = nf = 2.0165 x 0.3327 = 0.6709 I
I g1 = (0.036x0.2605) + (0.7026x0.0512) = 0.045351 ~ 0.0454
=
- 0. ,O. 6
= 0.2189 + 0.3882 = 0.6071 cm
~
Df .- 1.647 cm "1
,= -;6';=36.28cm2 S1 g 2,D eh !
ath 266 l
f O 1 .7026 -1
= 5.094 cm D " 0.418 +0.160 0.26050.161+ 0.280 3.97 + 0.182 Dg = 0.1963 cm E = (0.000254x28.51) + (0.000019x0.117) + (0.0182x0.0123) +
(0.0280x0.202) + (0.2605x0.0197) + (0.7026x0.00596) =
~
0.0224 ca L = 4
= 8.7470; L = 2.9575 cm Reflector saving (6)
D 6 = f L, 0.418 0"0.1963(0.00596)1/2 0.418
= 3. 3 cm Pit dimensions: 77.6 cm x 132 cm x 132 cm to/6: 85.46 x 139.86 x 139.86
~
B
= (8 46) + (13 .86) = 0.00135 + .0.001009 = 0.00236 cm
~T -0.0856 e = exp(-36.28 x.0.00236) = e = 0.918
~
-2 1+LB = 1 + (8.747 x 2.36 x 10 ) = 1 + 2.06 x 10 = 1.0206
-B t
=
=
0.6709 x 0.918 = 0.6030 k etf 1,gB 22 1.0206 O 267 1
0 16. TECHNICAL SPECIFICATIONS The VTAR Technical Specifications are virtually rewritten from the i original specifications for several reasons: (1) to conform more closely to ANSI /ANS 15.1. The Development of Technical Specifications for Research Reactors; (2) to address the new systems and equipment installed, and (3) to meet the need for maximum operational flexibility. Most of the limiting conditions for operation and limited safety ;
. system settings are unchanged from the original set points that were established when power was increased to 100 kW. These established set points have been shown to be sufficiently conservative to_ prevent unsafe operating conditions. The Safety Limits are based on actual experimental data (SPERT, BORAX, etc.) and theoretical data or a combination of both. ,
In conformance' with ANSI /ANS 15.1-1982, the statements under subheadings of specifications provide data, conditions, or limitations which bound a system or operation. Statements of Bases provide background or reason for the choice of specifications. Therefore, only the " specification" statements are governing. i O . 4 O 268
l im i V 16.1.0 DEFINITIONS 16.1.1 Abnormal Occurrence An abnormal occurrence is any one of the following: A) Operation with actual Safety System Setting (s) less conserva- . tive than the specified Limiting Safety System Setting (s). j B) Operation in violation of Limiting Conditions for Operation. C) Operation with a Safety System component malfunction which renders, or could render, the Safety System incapable of per- ! forming its intended safety function. D) An observed inadequacy in the implementation of administrative or procedural controls such that the inadequacy causes or could have caused an unsafe condition with regard to reactor oper-ation. E) A release of fission products from the reactor fuel element (s) of a magnitude to indicate a failure of the fuel element clad-i ding. F) An uncontrolled or unanticipated release of radioactivity to O the environment. 16.1.2 Channel The combination of sensor, lines, amplifiers, and output devices which are connected for the purpose of measuring. the value of a para-meter. 16.1.3 Channel Calibration the adjustment ~or the channel such that it's output responds, within acceptable range and accuracy, to known values of the para:neter which the channel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm, and trip. 16.1.4 Chanael (Operability) Check Qualitative verification of acceptable performance by observation of channel behavior. 16.1.5 Channel Test The introduction of a signal into tne channel to verify that it is operable. 10 i 269
O 16.1.6 Confinement A closure on the Reactor Room Cell (Rm 10) which controls the j movement of air into and out of the cell through a controlled path. i l 16.1.7 Experiment An apparatus, device, or material placed. in the Reackor core, ) experimental facility, or in line with a radiation beam emanating from l the Reactor, excluding devices being employed to measure reactor ! l characteristics such as detectors and foils. 16.1.8 Experiment, Fueled All experiments containing Special Nuclear Material (SNM). 16.1.9 Experiment, Movable An experiment where it is intended that the experiment may be moved in or near the core or into and out of the reactor while the reactor is operating. 16.1.10 Exreciment, New An experiment which, in the opinion of the Reactor Supervisor or O the Reactor Radiation Safety Officer, differs from experiments previously carried out at the facility. 16.1.11 Experiment, Secured 4
' Any experiment, experimental facility, or component of an experi-ment that is held in a stationary position relative to the reactor by mechanical means. The restraining forces imast be substantially greater than those to which the experiment might be subj ected by hydraulic, ,
pneumatic, buoyant -or other forces which are normal to the operating environment of the exp:;ciment, or by forces which can arise as a result of credible malfunctions.
! 16.1.12 Experimental Facilities A device for.the insertioa of an experiment. They shall'be limited to the following:
! . A) Graphite Thermal Column (located on tha west face) i B) Shield Tank (located on the east face)- C) Removable Graphite Stringers - in the Central Graphite Reflector Region (located between core tanks)
~
l 270 l R
O 1 D) North and South Beam Ports (located on the north and south l faces) E) Pneumatic Rabbit Systems . 16.1.13 Limiting Conditions For Operation (LCO) The lowest functional capability or performance level of equipment required for safe operation of the reactor. 16.1 14 Limiting Safety System Settinga (LSSS) Maximum or minimum settings for automatic protective devices or alarms related to those variables having significant safety implica-tions. 16.1.15 Measured Value The value of a parameter as it appears on the output of the channel. 16.1.16 Operable A system or component which is capable of performing its intended function in its normal manner. 16.1.17 operating A system or component which is performing its intended function in its normal manner. 16.1.18 Operator, Certified An individual authorized by the Nuclear Regulatory Commission and .; this facility, to carry out the responsibilities associated with the position. 16.1.19 Operator, Class A Reactor Any individual who is certified to direct the activities of a Class-B Reactor Operator. Normally referred to as a Senior Reactor Operator. 16.1.20 Operator, Class B Reactor An individual who is certified to manipulate the controls of a reactor. -Normally referred to as a Reactor Operator. O 271
1 l 3 s/ 16.1.21 Protective Action Protective action is the initiation of a signal oc the operation of equipment within the reactor safety. system in response to a variable or condition of the reactor facility having reached a specified limit. A) Channel Level I At the protective instrument channel level, protective action is the generation and transmission of a trip signal indicating that a reactor variable has reached the speci-fled limit. B) Subsystem Level . At the protective instrument subsystem level, protective action is the generation and transmission of a trip signal indicating that a specific limit has been reached. NOTE: Protective action at this level would lead to the operation of the safety shutdown equipment.
- C) Instrument System Level At the protective instrument system level, protective action is the generation and transmission of the command
() signal for the safety shutdown equipment to operate. D) Safety System Level At the reactor safety system level, protective action is the operation of sufficient equipment to immediately shut down the reactor. , 16.1.22 Protective Instrument Channel That combination of discrete modules and interconnections necessary to sense one reactor variable or condition and to initiate and transmit i a protective signal if and when that ' variable reaches a specific limit. -(See Figure 16-1) , 16.1.23 Protective Instrument Subsystem The _ combination of protective instrument- channels and any decision logic units necessary to determine that one of the reactor variables or conditions has been reached .and to transmit the necessary protective signals. (See Figure 16-1) i . O 272
9.a a o O - s sn
*E CONTROL -e c.........,
s p....... 3 gaz- O e 8 : DRIVE ' W SYSTEM iCONTFOLLER8 --------~ j 8 2 INPUTS ! ! 8 3 8
-bg i..........s c . .. .,. . . . a >
,gg CONTROL SYSTEM INTERLOCKS %$$
i l i l I m km i I " *a E k i ! d MAGNET MAGNET SAFETY POWER CURRENT ---=- MAGNETS - RODS >h'- SUPPLY " SWITCHES" o - o
=
m SAFETY 1 INTE h g m r
- DECISION - 7, l LOGIC UNITS MANUAL l
8 I (S.YS.T.E.M LE_V_E_L.).J t.. INITIATION
/ OTHER OTHER $-
SUBSYSTEMS - - ' SUBSYSTEMS M DECISION . I LOGIC UNITS I
! N
- u(SUBSYSTEM LEVEL)s! e g
/ \ >
E >E ~ D 5 E Z w a m' D = n > m w 3 %
- g U
$ 0 g w l
l l l $ w SE c. ' Is E E E
- x ~
8 y$ rc as E E
$c o
m w m T I I & x ' 0 " z W l O Figure 16-1 Generalized Reactor Safety System 273 (
() 16.1.24 Protective Instrument System That part of the Reactor Safety System which is the total of all > protective instrument subsystet* necessary to sense an occurrence of all reactor safety variables.and conditions and to initiate the operation of the safety shutdown equipment. Provisions for manual initiation based upon the decision of the reactor operator are included. (See Figure 16-1) 16.1.25 Reactivity, Excess The amount of reactivity which could be inserted if all control ; I rods .were fully withdrawn f rom the point where the reactor is exactly critical. (k,gg = 1.0) 16.1.26 Reactivity Limits The reactivity limits are those limits imposed on the reactor core excess reactivity. Quantities are referenced to a Reference Core Condi-tion. i 16.1.27 Reactivity Worth Of An Experiment The reactivity worth of an experiment is the maximum absolute value of reactivity change that would occur as a result of intended or antici-pated changes or credible malfunctions that alter the experiment posi-(N /~T tion or configuration. i 16.1.28 Reactor Coolant Core Overflow Line The core tank outlet piping line which allows primary coolant to be returned to the storage (dump) tank during normal operations. 16.1.29 React 3r Operating (Mode I) The reactor is considered to be operating whenever it is not in a , secure.d. shutdown, or standby mode. 16.1.30 Reactet Standby (Mode II) The reactor is considered to be in standby whenever: A) The dump valve key is inserted; and, B) 'The minimum. shutdown _ margin of section 16.3.1.1(B) is ) maintained.
~
16.1.31 ' Reactor Shutdown (Mode III) The reactor is considered to be shutdown whenever: (:) 274 I
l A) The primary core tanks are devoid of water, and B) The dump valve key switch is in the "off" pcsition with the dump valve key removed. i 16.1.32 Reactor Secured (Mode IV) l The reactor is considered to be secured whenever: A) It contains insufficient fissile material to attain criticality under optimum available conditions of . moderation and reflec- i tion; or, B) The reactor console key switch is in the "off" position with the console key removed,' and: 4 4 1) The primary core tanks are devoid of water; and,
- 2) The dump valva key switch is in the "off" position with the dump valve key removed; and,
- 3) All control rods are fully inserted.
16.1.33 Reactor Safety Systems Reactor safety systems are those systems, including their associ-O ated input channels, which are designed to initiate automatic reactor protection or to provide information for initiation of manual protective 4 ' actions. 16.1.34 Reference Core Condition The excess reactivity condition, when the core is evaluated at a ' temperature of 840F *2 (290C) and with a negligible reactivity effect of l Xenon. Any changes of excess reactivity' should be compared with previous values to assure conformity with the anticipated value changes due to the specific change of core configuration. More specific constraints.will be noted in the reactivity measurements procedure. j I6.1.35 -Research Reactor A device designed-to support a self-sustaining neutron chain reac-tion for research, development, ' educational training, or experimental purposes. 4 i J 16.1.36 Rod, Control l A device fabricated from neutron absorbing material which is 'used to establish neutron flux changes and compensate for routine reactivity changes.. O 275 _ _. . ._. _- . , - ~-. _. . . ._._. -
. b LJ 16.1.37 _ Rod, Regulating The regulating rod is a low worth control rod which does not have scram capability and is used to maintain a desired power level. Its position may be varied manually or by an automatic controller.
16.1.38 Rod, Safety A control rod which is coupled to its drive unit in such a way as to allow it to perform a safety function when the coupling is disengaged. 16.1.39 Scram Time The elapsed time between initiation of a shutdown signal at the subsystem level and the instant a control rod becomes fully inserted. 16.1.40 Shall, Should Or May The word "shall" is used to denote a requirement; the word "should" to denote a recommendation and the word "may" co denote permission, j neither a requirement nor a recommendation. 16.1.41 Shutdown Margin The minimum shutdown reactivity necessary to provide confidence that the reactor can be made suberitical by means of the control and safety systems starting from any permissible operating condition with the highest worth shim or safety rod and the regulating control rod remaining in their most reactive position (i.e. , fully withdrawn), and the reactor remaining subcritical without further operator action. 16.1.42 Shutdown, Unscheduled Any unplanned shutdown of the reactor caused by an actuation of the reactor safety ' systems, operator error, equipment malfunction, or a manual shutdown in response to conditions which could adversely affect safe operation, not including ehutdowns which occur during testing or check-out operations. 4 1 O l 276
i P 16.2.0' SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 16.2.1 Safety Limits 16.2.1.1 Applicability i These specifications apply to the neutronic and thermal-hydraulic characteristics of the reactor core. ) 16.2.1.2 Obj ective The objectives are to insure fuel element and cladding integrity. i Specifically, the following conditions must not occur. A) Bulk or saturated boiling of the primary coolant, or B) Subcooled or' surface boiling of the primary coolant, or i C) Fuel melting in the event of a loss of coolant accident (LOCA) or a loss of flow accident (LOFA). 16.2.1.3 Specifications A) Maximum steady state thermal power is (200 kW). O B) Minimum primary coolant flow is (15 gpa). C) Maximum steady state primary coolant outlet temperature shall not exceed (2000 F). 16.2.1.4 Bases l The. process variables are listed in order to reflect the basic physical candition of the reactor. By not excaeding the established Safety Limits - it can be assurad that the integrity of the principal physical barrier (ie. fuel . cladding) will not be breached. No one safety limit can assure this but using all combined, will give this assurance. The maximum steady state power limits the amount of heat that will be added to the system. The minimum flow establishes the minimum heat removal capability of the - system. Finally, aetting a maximum outlet coolant temperature allows for consideration of all other 1 systems important to cooling the reactor but more importantly, will not allow for bulk boiling to occur and as such maintain fuel clad
' integrity.
i r O. 277
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- m. . , - ~, , .,mv. - . . . . ~ .. er ,. .-- -,-
~ .- . - . .- - - . . .. . . - - - - -. .
16.2.2 Limiting Safety System Settings (LSSS) 4 16.2.2.1 Applicability These specifications apply to the Reactor Protection Channels. 16.2.2.2 Ob_tectives ! The objectives are to ensure automatic protective action is available to prevent a safety limit from being exceeded. 16.2.2.3 Specifications t
, A. The following conditions will initiate automatic protective actions (SCRAM):
- 1) Power greater than o'r equal to one-hundred twenty-five (125) kW steady state thermal,
- 2) A five (5) second period or shorter,
- 3) A core -tank water level one (1) inch or less from the top of the core tanks,
- 4) A primary coolant flow rate of less than or equal to tw;nty (20) gpe,
- 5) An earthquake causing 0.14 j's lateral acceleration or greater (three (3) on the Ricater scale).
! 16.2.2.4 Bases i
All specifications of this section 16.2.2 are established to prevent the violation of the established Safety Limits. Prevention of Safety Limit violations is assured by requiring that any Limiting Safety System Setting setpoint, when reached, . will initiate automatic protective actions. Limiting a maximum steady state power l'evel will prevent operation in excess of a known flux level such that any credible accident scenario will not result - in fuel melting or core dismantling. , It is necessary to maintain a safe and controlled power increase rate ..
- ' and this is accomplished by establishing a limit ' on reactor period.
- From past operating experiences it was found that wetting graphite introduces a large amount of negative reactivity and that there is very
. little control of the positive reactivity effects introduced - while the 4
graphite ' is drying. - With' this fact known, a limit was set as to the
, maximum core tank water level, ' to prevent overflow and . the ~ subsequent wetting of the graphite. The limit on primary coolant flow rate was set so that the reactor would be shutdown prior to reaching the safety limit
_ established in section 16.2.1. Finally, to provide assurance that in the event' of _ a significant. seismic disturbance, the ' reactor would be O 278 4 4
, ---a - - , - - - - , - ,w - -,r. - <, n ,, -+
3 i l
- - - shutdown before the disturbance reached an intensity large enough . to cause damage to the core or components, a lateral acceleration limit was established.
e l l 1 i i I I 1 i ! i O 4 r 4 Y i. l I, ' LO 279 1!
. - c. - ._ . . _ . . - . , - _ . . . _ - - - . - . _ - . - _ - - _ - _ _ _ _ - - - . . . - - _ . , _ . - - . _ _ _ _ . . . . , . . . . - . . , ~ , . _ _ -
16.3.0 LIMITING CONDITION FOR OPERATION (LCO) Limiting conditions for operation are the lowest -functional capa-bility or performance level required for safe operation of the facility. The reactor shall not be taken critical unless the following criteria have been established. 16.3.1 Reactivity Limitations These specifications apply to the reactivity condition of the reactor, control rods, and experiments. The specifications help prevent exceeding a LSSS or SL and provide assurance that the reactor is oper-ated with the maximum amount of safety measures available. 16.3.1.1 Specifications i A. Excess Reactivity The maximum positive excess reactivity, including the total from experiments, shall not exceed 0.6% &/k. B. Shutdown Margin The minimum shutdown margin shall be 0.5% &/k, with the
~
highest worth shim or safety rod and the regulating rod fully withdrawn with the moderator present. 4 C. Control Rods 5 The maximum reactivity insertion rate of any control rod shall
- not excead 0.02% &/k par second. This reactivity insertion rate shall not be applicable when the reactor is in mode III or IV.
D. Experiments The maximum absolute reactivity worth of any single experiment shall not exceed 0.3% &/k. Movable experiments shall not have , a reactivity insertion rate, or be introduced into or out of J experimental facilities, with values exceeding 0.05% &/k per second. This amount (0.3% &/k) corresponds to a period of about 130 seconde. E. Coefficients Void and Temperature The primary coolant void and temperature coefficients of reactivity shall be negative over the facility lifetime. O 280
es a ~ o x O F. Neutron Source A neutron source of approximately 10 6neutrons per second and i able to indicate 3 counts per second (eps) or greater on a source range instrument shall be positioned during startup in the central graphite reflector region of the core. 16.3.1.2 Bases These specifications are provided to limit the rate of power increase to the lowest possible values in the event of any credible accident. They also provide for assurance that a reactor shutdown can be achieved even with the failure of two control rods and the dump valve. Also, they assure a negative overall core coefficient .of reactivity and limit the maximum reactivity addition rates to levels commensurate with safe reactor operation, from the addition from control rods or experiments. Disallowing the control rod reactivity insertion rate when the core is devoid of the water moderator (Modes III and IV), gives flexibility for the maintenance and repair of control rods, while still maintaining the reactor in a highly safe condition with no possibility of the reactor becoming critical. Lastly, a maximum excess reactivity of 0.6% ak/k precludes a prompt critical excursion under any circumstance. 16.3.2 Reactor Control And Protection System This section applies to those components which actuate interlocks, alarms, indications, or otherwise establish minimum performance charac-teristics essential for safe reactor operation, plus specifying the minimum control and protective criteria for safe reactor operation. 16.3.2.1 Specifiestions 16.3.2.1.1 Control Rods A. All control rods shall be operable B. Total scram time for each safety rod and the shim rod shall not exceed 0.6 seconds C. All control rod top and bottom limit indicators shall be oper-able D. The shim and regulating rod position indication meters shall be operable. i' E. The maximum rate of control rod controlled motion shall not exceed 6.6 inches per minute for the shim and safety rods, and 33.0 inches per minute for the regulating rod. O
, 281
F. The automatic regulating rod control circuit may be inoperable provided remote manual control (with position indication) is available at the control console. G. The shim and regulating rods shall be- capable of both inward and outward controlled motion provided that neither is at a travel limit. H. Rod withdrawal is prevented if any one of the following condi-tions is not met:
- 1) Dump valve fully closed. I
- 2) Core tank water level greater than the core tank overflow pipe level.
- 3) Safety rod one fully withdrawn prior to withdrawal of any other rods.
- 4) Safety rod two fully withdrawn subsequent to safety rod i
one withdrawal and prior to withdrawal of the shim and regulating rods ' I. Dump valve closure is prevented if any one of the following conditions are not satisfied:
- 1) Neatron count rate on' source range of three counts per second or greater.
- 2) Reactor room ventilation fans energized.
- 3) Scram bus continuity present.
- 4) Primary coolant pump etergized.
- 5) Secondary coolant temperature greater than 55 0 F at inlet and cooling tower pan RTDs.
- 6) Pri7ary coolant temperature greater than 700F at inlet RTD.
16.3.2.1.2 Reactivity Insertion Rates Maximum reactivity insertion rates are listed in section 16.3.1 Reactivity limitations. 16.3.2.1.3 Nuclear Instrumentation A) At least two channels shall be operating on scale through all ranges of reactor power. l O 282 l m -
4 0 B) At least one channel having a high neutron flux level trip shall be operable (see Table 16-1 note 3). C) Intermediate range five second period trip shall be operable. D) Intermediate range period circuit shall be operable. E) At least one source range channel shall be operable and have indications of three counts per second or greater with the neutron source in the core. Nuclear instrument capabilit'ies including channel, range, and function are listed in Table 16-1. 16.3.2.1.4 Scram Channels A) Manual scram capability shall be available at the control console and from at least one remote scram switch in the reactor room. B) Automatic earthquake scram capability shall be operable. C) Automatic primary coolant low flow scram shall be operable. D) Automatic core tank high water level scram shall be operable. E) Automatic fast period intermediate range scram shall be oper-able. F) At least one high reactor power level automatic scram shall be operable (See Table 16-2 note 1) G) The scram channels shall be constructed such that no single electrical fault which partially or completely disables the automatic scram function can, in any manner, impair or disable the manual scram function or vice-versa. H) All reactor scram channels shall be capable of initiating reactor shutdown independent of one another. The reactor scram channels shall be fail safe. Table 16-2 lists the minimum reactor protection system parameters, set-points, and actions. 16.3.2.1.5 Backup Scram Channels _ A back up scram channel is provided for each main scram channel. These back-up scrams use the same detection mechanism that the main scrans use; however, the actuation' circuitry is totally independent. O 283 ) i l l
J-i j TABLE 16-1 Nuclear Instrument Capabilities 4 .-
# Channels
' Channel required Range Function
-12 Indication, Recording, Keithley 1 10 - 10 amps Automatic Power Controller
- Signal 3 Indication, over power trip Power Fange. 2 0-150% Power Intermediate '
Range 1 10 - 10 neutron flux (nV) Indication, Period trip, Period Alarm Source Range 1 1 - 10 counts /sec. (CPS) Indication, 3 CPS Start-up 4 Interlock i NOTES: l 1 4 1. At least two channels shall be on scale at all times. l 2. -The Keithley and Source Ranges have two instruments each available, however, only one each shall be required for operation. I 3. Two channels shall normally be operable. In the event only one Power Range channel is operable, reactor operation may be allowed under the following conditions: { 5
- a. The functioning Power Range channel setpoint shall be reduced to 100% overpower trip.
! b. The scaximum operating reactor power level shall f he 90% full power. < c. Reactor Safety Constittee approval. J t i ! T a !Q . 1
- 284-
., - , . - - , . , , , r-- , - - . - - -r, - --r
(j s TABLE 16-2 Reactor Protection System Parameters Set Point Action Keactor Power 1 > 125% Full Power Full Scram Intermediate Range Period i 5 Second Period Full Scram Core Tank High Water Level i 1 inch from top of core tank Full Scram Primary Coolant Flow 1 25 g.p.m. Full Scram i Earthquake > 3.0 Richter Full Scram Manual Console, Reactor Room Full Scram (3 locations) NOTES:
- 1. - Two Protection channels shall normally be operable. In the event only one power range protective channel is operable, operation may be allowed under the following conditions:
- a. The functioning Protection channel Set point shall be reduced to 100% overpower trip
- b. Maximum operating power shall be 90% full power.
- c. Reactor Safety Committee approval
- 2. All manual scram channels shall be operable. In the event some manual scram channels are not functioning, reactor operation may be permitted when the fc11owing conditions are mets
- a. Manual scram capability shall be available at the control console and at least one remote scram switch-in the Reactor Room
- b. Reactor Supervisors' approval
,.)
, 285
. = - , . , , . . .. . - . . .- _ __ . _ _ . -. . ._ ..
5 a ' 16.3.2.1.6 Bypass Conditions During normal operation at low power levels certain safety func- " tions are not required and must have by pass provisions. Additionally experiments at low power levels require by pass criteria. Allowances I must be made for routine surveillance which also require certain inter-locks and functions to be bypassed. Unless specifically listed in Table 16-3 no functions, interlocks, etc. will be bypassed if the reactor is operating (Mode I) 16.3.2.2 Bases w h The integrity and gerability of the Reactor Protection System must be verified. The reactor protection system provides the operator with parameters from which to safe?.y control the reactor. The operability of rod position indicators and the rod travel rates provide for the precise control needed of the reactor, due to reactivity additions f rom Xenon, experiments, etc. by the operator. Having a scram response time of 0.8 seconds assures that the reactor will be shutdown before excursions reach safety limits. Parts G and H of 16.3.2.1.1 list the conditions r which allow for a stepwise approach to criticality and insure that the
- reactor is in an appropriate operable condition for criticality.
p Nuclear instruments provide for observation of reactor pcwer through all levels with the necessary redundancy and protective capabilities. All L g scram and redundant back-up scram channels provide for a safe operating margin (allowing for scram response time) prior to reaching a safety r limit in the event of an excursion. Provisions are made ' for trip setpoint reductions, in the event of a channel failure, that allow 5 operation at a reduced power levei while still maintaining a safe 7 operating margin. { f; 16.3.3 Coolant Systems g This section applies to the coolant system components, primary, secondary and emergency, which are required for heat removal and neutron [ moderation of the reactor. This section also specifies the minimuu % operating equipment and limits required for safe reactor operation. 0 These specifications are also listed in Table 16-4. E f 16.3.3.1 General Specifications h A) Any moisture in the vicinity of piping where the cause is l- unknown shall be considered a leak until determined otherwise. ' In the event of a loss of power all pumps shall not restart B) m automatically. L r a O - 286 L K
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O TABLE 16-3 Bypass Criteria _ - Item Number Condition Bypass Provisions 1 Maximum control rod motion May be by-passed, by a class A operator rate (section 3.2.1.1-E) control rod tests or maintenance with . reactor in modes III or IV. 3 2 Control rod withdrawal inter- May be by-passed, by_a class A opera-locks (section 3.2.1.1-H) tor, for control rod tests or main e tenance, such as rod drop tests. j 3 Neutron count rate less May be by-passed,*.with R.S.C. appro-than set point (section val- provided a pico-ameter channel 3.2.1.1-I-l and section is fully operational in the same range 3.2.1.3-E) (> 0.1 watt) and on scale. 4 Shield tank water level May be by-passed, with RSC approval, (section 3.3.2.3-D) for specific experiments. 5 Primary Coolant / Moderator May be by-passed, with RSC approval, temperature specifications for specific experiments. () (section 3.3.2.3-G) 6 Primary Coolant / Moderator May be by-passed, with RSC approval, ; j flow specification for specific experiments. l (section 3.3.2.3-H) f 7 Secondary Coolant tempera- May be by-passed, with RSC approval, . ture specifications (section for specific experiments. 3.3.3.3-E) ( , 8 -Secondary Coolant flow May be by-passed, with RSC approval, ' ' specification (section for specific experiments. 3.3.3.3-H) l 9 E.C.C.S. Flow Specification May be by-passed, by a Class A opera-
- (section 3.3.4.3-D) tor, for reactor operations restricted
- to 100 kWt or less.
10 E.C.C.S. B.I.T. Water Inven- May be by-passed, by a Class A opera- ! tory Specification (section tor, for reactor operations restricted 3.3.4.3-E) to 100 kWt or less. May be by-passed,'by a Class A opera-
~
11 1 E.C.C.S. Boric Acid Inventory . Specification (section bor, for . reactor operations restricted 3.3.4.3-F) to 100 kWt or less. 12 E.C.C.S. Cooling Isolation May be by-passed, by a Class A opera-Valve Specification (section . ' tor, . for reactor operations reetricted
-3.3.4.3-G) to 100 kWt or less.
-287
, . , - - . , -. ~ , . . - , , , .
h Table 16-3 (continued) I i g Item i tV - Number Condition Bypass Provisions 13 E.C.C.S. Valve Alignment May be by-passed, by a Class A opera-
. Specification (section tor, for reactor operations restricted 3.3.4.3-H) to 100 kWt or less.
14 E.C.C.S. BIT Water Inventory May be by-passed, by a Class A opera-Surveillance requirement tor, for reactor operations restricted ! (section 4.4.3-A) to 100 kWt or less. 1 15 E.C.C.S. Boric Acid Inventory May be by-passed, by a Class A opera-and Location Surveillance tor, for reactor operations restricted requirement (section 4.4.3-C) to 100 kWt or less. f
+
O d i i O. l 288
~
__ _ __ _ . - _ , _ _ . ._m _ _ _ _ - - 1 o O taste 16-4 Major Cnolant System components Comprwwint g System Range / Capacity Function Setpoints Action Coolant Pump Centrifical l'rlmary 0-120 gpm Primary coolant locomotion 400*C Statorvinding High Temperature Alarm (canned rotor / sta ror) Flow Monitor Paddle Wheet Primary 0-200 gre Monitor Primary Coolant Flow < 25 gpm tow riow Scram Neat Exchanger Para 11e1 Plate Primary / ,0-500 kwt acuove heat from Primary coolant NA NA Temperature R.T.D. Primary 0-200*r Monitor inlet and outlet - 70*r inlet C Temperature Alarm Monitors High Remperature Scraw coolant temperatures 1 170*r outlet Radiation C.M. Tube Primary 0.1-10 Mn/hr Detection of fuel element failures 5 a normal Radiation in Primary Coolant Munitor Alarm 14 vel Pressure Dia- Primary 0-10 PSID Determination of core and ahteld Detectors 3,hram micro Above overflow pipe Had control Interlocks tank levels in core tanks '
""A *
< 2 ft. from top of' Shield Tank tow tavel Alarm shield tank Dump valvo 6 in. Angle- Primary 0-5.0 PSID Direct crimary Coolant to core Open limit Globe Indication and Rod Control tanks or dump tank Closed limit Interlocks Domineralizer Mined resin Make-up O.18 MG gurity Filter and purify make-up water < 0.2 pil Indication (local only) bed for primary system g Coolant Pump Centrifical Secondary 0-750 gym Secondary Coolant 8 - tion MA On.off indication Co
@ Throttle Valve 6 inch Butter- Secondary 0-1004 (Iow Regulate Secondary coolant Plow NA Indication fly riow Monitor Sonic Secondary 100-1000 gpa Monitor Secondary Coolant Flow < 125 gpa Alarm Cooling Tower rotecd draf t Secondary / 0-500 kW
- Atmospheric th 'I """ *"**'~ ""# I""*'I '
a ture Temperature Alarm Temperature R.T.D. Secondary 0-200*r Monitor Secondary Coolant Temper- > 55*r inlet . tow Temperature alarm Rod Control sennitors tures into and out of N.E. Interlocks
< 115'r outlet Nigh Temperature Alarm Coolant Pump Centrifical E.C.C.3. 0-25 ggas Provide emergency coolant flod NA Off-on Indication (canned rotor / through core tanks stator) i Neat Ex- Tube and Shell E.C.C.S. b-100 kW Remove heat from E.C.C.S. coolant NA NONE Changer Core Tank Sight Class Primary' O-74 of core Provide for core tank high level C 1 inch from top of High Water 14 vel Scram tevet tar.k s scram core tanks moron Injec- NA E.C.C.S. 0-30 gal Provide for dilution of Doric acid > 10 gallons se0NE tion Tank
[ Disap Tank IsA Primary 0-220 gal Trovide surge volume and head for > 200 gallons NOME j primary coolant i conductivity Dual Platiness Primary 0-3.0 naho/cm Monitor primary water purity 4 1.5 nauho/cm memote and local indication Probe Electrode
1 f I 16.3.3.2 Primary Coolant System 16.3.3.2.1 Major-Components A) Primary coolant pump i l B) Flow monitor
\
C) Conductivity monitor J D) Primary to secondary heat exchanger i E) Temperature monitors .] F) Primary coolant radiation monitor and hold-up tank G) Core tank / shield tank water level monitors H) Dump valve I) Dump tank l- J) Core tanks t. K) Shield tank j 5 L) Demineralizer M) Associated valves. H lters, strainers, and interconnecting piping
- 16.3.3.2.2 -Flow Path
- The primary coolant flow shall be provided by water pumped from the dump tank, via a two inch line, through the , heat exchanger and into a six inch line leading to the bottom of the core tanks. A four inch overflow pipe near the top . of each core tank shall return the ' coolant water to the dump tank by gravity. . A scram condition shall open the dump valve connecting the six inch dump-inlet pipe to the dump tank, draining the core tanks and bypassing the core tanks in the coolant ,
loop. 16.3.3.2.3 Specifications A)- Primary coolant water- shall- have a maximum value of 1.5 y ahos/cm conductivity (0.67' Megohns cm resistivity). During operation, the actual conductivity may fluctuate as high as- 2.0 y ahos/ca (0.5 MQ-en) if ' the less than or equal to 1.5 y ahos/cm value can be met when averaged over four hours. O . 290 4
- - - . , , , , - - - - . , , - . , - - - - - - , . - - -- - - , - ,~r.. -
B) The components of the primary coolant system with the exception of core tanks, make-up demineralizer, heat exchanger and con-necting piping, and fission product detector holdup tank shall be located in a process pit in the reactor room.
- 1) The process pit shall be large enough to contain the entire volume of primary coolant.
- 2) A drain into che process pit from beneath the core tanks shall be provided.
- 3) A sump pump, which is normally de-energized, shall be the primary means of emptyf.ng the process pit.
- 4) A drain into the process pit from beneath the primary piping run to and from the heat exchanger shall be pro-vided.
C) Make up water shall be supplied to the primary by demineraliz-ing city water. D) Shield tank water level shall be within two feet from the top of the shield tank for reactor operation greater than 10 watts reactor power. E) Core tank water level shall be equal to or greater than the O core tank overflow pipe level any time the reactor is operating (Mode 1). F) The dump valve shall be fully closed any time the reactor is operating (Mode 1). G) Primary coolant temperature shall not be less than 700F and not greater than 1700F when measured at the primary coolant temperature monitors. H) The primary coolant flow shall be greater than 25.0 gpm for reactor operation.
- 1) Primary coolant radiation monitor shall be operating whenever the reactor is operating in Mode I.
16.3.3.3 Secondary Coolant System 16.3.3.3.1 Major Components The major components of the secondary coolant system shall consist of: A) Secondary coolant pump O 291
- t; i
te . g I I2 j g B) ' Throttle valve C) F1ow monitor , D) Primary to secondary heat exchanger E) Temperature monitors i F) Cooling tower ,.
, d) Associated valves and interecnnecting piping.
3 If,.3.3.3.2 Flow Path
\ The secondary coolant flow sha11 be provided by water pumped from the cooling tower co1d water basin through the primary to secondary heat exchanger, to the hot distribution basins of the cooling tower.
16.3.3.3.3 Specifications A) Secondary coolant water shalt have coo 1 ant activity levels less than those values 111sted in 10 CFR 20 Appendix B, Tabte II. B) Make-up water shal1 be supp11ed by city water. O C) Ceetins tcwer ce1dwater basin tevet sba11 be maintained within six inches below the overflow drain, during reactor operation. ,, D) The secondary throttle va1ve sha11 be shut prior to stopping or
% starting the secondary coolant pump.
E) Secondary coo 1 ant temperatures shalt not be less than 550F and not greater than 1150F for any reactor operations. F) The secondary coolant system shal1 be maintained at a higher pressure than the primary system. G) The secondary coolant system water shal1 be treated with corro-sion and biologica1 growth inhibiting chemical (s). H) Secondary . coo 1 ant flow shall be greater than 125.0 gpm for reactor operation. 16.3.3.4 Emergency Core Cooling System (ECCS) The emergency core coo 11ng system is not required for 100 kW oper-l ation and shal1 only-be used for tes' ting and, training. This section has p" been included in ~ anticipation . of an upgrade in reactor power. This [, section shall not be considered binding unti1 the actua1 upgrade occurs. t s 292 y
- ,, 7, b N ,.
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. 9 '% yt O x 163341 " ser ce ae e t-A) ECCS pump B) ECCS heat exchanger t. C) Boron injection tank D) Dump tank ' E) Shield tack F) Cooling water isolation valve G) Associated valves e.nd interconnecting piping 16.3.3.4.2 Flow Path Emergency core cooling flow shall be provided by water pumped from the dump tank or shield tank through the ECCS heat exchanger and into a six inch line leading to the bottom of the core tanks. A four inch overflow pipe near the top of the core tanks shall return the emergency coolant to the dump tank by gravity. The shield tank shall be used as a secondary water source in the event the dump tank becomes unavailable. 16.3.3.4.3 Specifications A) The components of the emergency core cooling system with the exception of the shield tank, boron inj ection tank, cooling water isolation valve, and associated interconnecting piping shall be located in a process pit in the reactor room. B) ECCS cooling water sources shall be (in order of priority):
- 1. Dump tank
- 2. Shield tank
^'
- 3. City water main C) The dump valve shall be fully closed any time the emergency cooling system is operating.
D) Emergency cooling flow shall be greater than 18 gpm. E) The boron injection tank water inventory shall be maintained greater than 10 gallons. F)' There shall 'be a ' minimum of 1 kilogram boric acid crystals (H2B03 ) in close proximity to the boron injection tank. O V 293
G) The ECCS cooling isolation valve shall be operable during all reactor operations. H) The emergency cooling system valves shall be aligned in such a way that the system can be placed in operation from the control console. 16.3.3.5 Bases 16.3.3.5.1 Primary Coolant System The primary coolant system must provide the main source of heat removal from the reactor core. In' order to assure that this function is accomplished safely it is necessary to specify particular settings. Leaks shall be treated as potentially contaminated until otherwise determined. Manual restart of pumps following a loss of power will prevent an
; inadvertent insertion of cold water to the core.
Primary coolant conductivity levels have been determined so as to prevent corrosion of the primary system and thereby reduce the activ-ation of corrosion products, and to ensure that fuel element integrity is not degraded due to corrosion of the clad. O Maintaining as mch of the primary equipment in the process pit as
!~
possible minimizes the areas in which contamination, due to primary coolant, ' may occur. This will aid greatly in the control' of " spills" which may occur. By keeping the process pit sump pump de-energized, the possibility of discharging contaminated fluids prior to sampling is kept to a mini - mum. For operation above 10 watts ' the shield tank level muat be no
, lower than two feet below the top of the tank to minimize radiation levels at ~. the top of the shield ' tank. Maintaining core tank levels above the overflow pipe level ensures that the fuel elements will be fully immersed in primary coolant during operation. The _ dump . valve shall be ' fully closed for all reactor. operations to ensure that all coolant flow is diverted through the core tanks and core short-cycling does not occur. Primary coolant temperature limitations-are set so that
, there will be no bulk coolant boiling at the high temperature limit and that no reactor operation will ' occur in a region where core parameters have not been determined,'in the case of the low temperature limit. The low primary flow limit is addressed to ensure ~ that the core has sufficient cooling flow to ~ prevent exceeding high temperature limitations.
~
.The coolant radiation monitoring system is required to ensure that fuel element failure monitoring capability is functioning whenever the reactor is at power.
O 294 l
.- \
16.3.3.5.2 Secondary Coolant System The secondary coolant system must be capable of removing the entire heat load of the primary coolant system without allowing any coolant to cross between systems. The secondary coolant system is maintained at a pressure higher than that of the primary to ensure that if there is a loss of system integrity in the primary to secondary heat exchanger, flow will be into the primary system and thus not contaminate the sec-ondary or cause an uncontrolled release of any kind. To ensure that there is no release via the secondary system, the activity level of the . secondary system is set at the minimum detectable level. To ensure that the secondary system has sufficient coolant a minimum cooling tower basin level has been established. Secondary coolant minimum temper- I l atures have been. established to minimize the effects of an inadvertent 1 cold water addition to the primary. The high secondary coolant temper- ) ature limit is established to assure that adequate heat removal capabil-ity exists under all anticipated operational conditions. The secondary flow rate specification is imposed to assure that ' there will be the
, necessary coolant flow required for heat removal at the maximum reactor power level. In order to prolong the life and ensure system integrity the secondary throttle value shall be shut prior to starting or stopping the secondary coolant pump. This also minimizes the chances of unknow-ingly adding cold water to the heat exchanger and causing a cold water excursion. The system shall be chemically treated with corrosion and biological growth inhibitors.
O 16.3.3.5.3 Emergency Core Cooling System The emergency core cooling system (ECCS) has been installed in anticipation of an upgrade in reactor power and at the time such an upgrade should occur this system would be required for decay heat removal and shutdown capability following the onset of a loss of coolant 4 or loss of flow accident. By locating most major components within the confines of the process pit system leakage and the possible associated . contamination will be controlled. The ECCS has three sources of cooling water which allows a sufficient supply in the event the systets is 4 needed. The dump valve must- be fully closed for system operation in order to divert cooling water to the core. To assure sufficient decay I heat ' removal capability a minimum ECCS coolant flow has been estab-lished. Maintaining a minimum water inventory in the boron injection tank ensures that enough water will be readily available to dissolve a kilogram of boric acid crystals (H280 3 ). The location and ease of access will be maintained by keeping the boric cid crystals in close proximity to the beon injection
- n/ .. . TM dCCS shall be aligned so that an operator, at the -. .a console, can place the system in opera-tion by stceting . the ECCS pump and operating the ECCS cooling isolation valve.
O 295
l 4 - 16.3.4 Confinement This section - applies to the operations which require confinement and shall state the actions necessary to achieve confinement. There shall be two modes of ' confinement; (1) normal confinement, and (2) emergency confinement. ' 16.3.4.1 Normal Confinement 16.3.4.1.1 Operations That Require Normal Confinement (Mode I) Normal' confinement shall be required whenever: A) The reactor is in Modes I or II; or B) There is irradiated fuel handling taking place; or C) There is in core maintenance or repairs being done; or D) There is movement of irradiations to or from the core; or E) There is movement of experiments which could cause a change in reactivity worth more than 0.20% Ak/k. 16 3.4.1.2 Equipment To Achieve Mode I Confinement O Normal confinement- shall require as minimum equipment: A) All Reactor room accesses are maintained closed and locked, i . except for personnel transitions. I B) The reactor room ventilation is operating and maintaining a negative pressure (with respect to surrounding areas) in the , reactor room. C) The reactor room negative pressure monitor is operating. 16.3.4.2 Emergency Confinement j 16.3.4.2.1 Operations That Require Emergency Confinement (Mode II) Emergency confinement shall be required whenever there is an alarm
- -or indication showing a high radiation. level in the following
A) Ventilation stack radiation monitor; B) Building radiation monitor (west wall of~ reactor room). O 296
l l l ( 16.3.4.2.2 Equipment To Achieve Mode 11 Confinement Emergency confinement shall require as a minimum the following: A) The reactor room evacuated and all accesses closed and locked; B) The reactor facility building (Robeson Hall) evacuated, accesses closed and entry restricted to radiation personnel; C) Reactor room ventilation secured. 16.3.4.3 Bases Confinement is required to minimize the release of radioactive contaminants and fission products from the reactor to the environment. By maintaining a Mode I confinement, minor surface and fixed contamin-ation is assured to remain within the confinement area. Maintaining a negative pressure within the confinement area assures that small releases will be kept in this area. Emergency confinement is required for events having the possiblity of major radioactive releases. In this case ventilation is shutoff to ensure the maximum amount of effluents will remain in the confinement area. The reactor facility building is evacuated and entry restricted to reduce radiation hazards to the staff, emergency personnel, and general public. ( 16.3.5 Reactor Room Ventilation These specifications apply. to operation of the ventilation system and shall state the minimum equipment necessary for operation along with any associated interlocks and alarms. 16.3.5.1 Components The minimum equipment necessa y for operation of the Reactor Venti-lation System shall be: A) Reactor main ventilation fan, B) Reactor booster ventilation fan, C) Dump tank booster fan, ,, D) Reactor core argon duct, E) . Dump tank Argon duct, F)' Reactor room main ventilation duct, G) Ventilation duct between reactor room and the discharge of the reactor main ventilation fan.
'297
O 16.3.5.2 Specifications A) Reactor room ventilation shall be operating whenever the reactor is operating in Modes I or II or to meet the confine-ment requirements of section 3.4. B) Reactor room ventilation shall be automatically secured when an alarm condition is received on the Building Radiation Monitor or Ventilation Stack Radiation monitor or both. C) Reactor room ventilation shall be considered operating when the fan is on and there is a negative pressure within the reactor room, with respect to surrounding areas. D) The reactor room ventilation system shall be equipped with a loss of negative pressure alarm. E) The reactor room ventilation system shall have monitors for radiation and air particulate fission products. 16.3.5.3 Bases The purposes of the ventilation system are to maintain a negative pressure within the reactor room and to vent short-lived radioactive v gases, specifically Argon 41, produced during reactor operation. To meet these requirements the system is operated whenever the reactor is operating. The radiation monitors are installed to indicate and record the' radiation levels of all effluents. Should radiation levels in the ventilation system become significant (limits are listed in section 3.7) the system will automatically shut-down to minimize the uncontrolled release. Maintaining a negative pressure (at low radiation levels) assures an air flow into the reactor room thcs reducing the airborne contamination travel to the surrounding area, i 16.3.6 Emergency Power Emergency power applies to the power source which is required to be available to maintain af ter shutdown reactor functions, controls, and indications in the event of a loss of normal power and specifies the type of equipment, and minimum operating time required. 16.3.6.1 Components The minimum components necessary for emergency power shall be: , 1 A) A converter unit B) A battery bank l 298
J O- .C) .An inverter unit. D) Interconnecting wire l 16.3.6.2 Speciffcaeion A) The battery bank shall be capable of delivering 48 VDC to the inverter unit for a minimum of 15 minutes under full load. The inverter shall be capable of delivering a single phase 220
~
B) VAC output. 16.3.6.3 Bases Emergency power is desirable to provide power to reactor instrumen-tation controls, and functions in a-loss of power situation. This shall 4 allow for the monitoring of reactor power levels and coolant tempera-tures to assure a complete reactor shutdown has taken place following a - loss of power. l NOTE: In anticipation of an increase in reactor- power level in the l near future, this system shall 'also provide power -for decay heat 4 removal, for up to one hour .under a limited load condition, follow- ) ing a loss of power. The limited load condition will not be. addressed at this time. , h 16.3.7 Radiation' Monitoring System And Effluents This section shall list the minimum radiation monitoring equipment required for the VTAR facility and also the minimum number of effluehts that shall be monitored. This section shall give the specifications on instrument types, effluent ' types, release limits and instrument c set points along with the bases for- which these are determined.
. Specifications are also listed in Table 16-5.
16.3.7.1 Fixed Radiation Monitoring i The . fixed (permanently installed) monitoring system shall have the following capabilities: l A) West reactor room area radiation monitor (building monitor), one detector.
- 1. G-M type detector with proportional compensation i
- 2. Alarm setpoint; 15 ares / hour or less (except as noted in Table 16-5) 3., Initiates building evacuation -alarm and de-energizes reactor room ventilation.
- LO l
- 299-t i
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'U TABLE 16-5 Permanent Radiation Monitors Item Location Range Setpoint Function Reactor Room Area Monitor East Wall 0.1-10 Indication and area overtoler-Reactor Room r -> 5 "hr ance alarm 4 "#*"
Reactor Room Building West Wall 0.1-10 ~> 15 "#*" Indication, building evacuation Monitor Reactor Room alarm and sequence 4 "#*" Reactor Room Vent Stack 3rd Floor Library 0.1-10 ~> 15 '"** Indication, building evacuation Monitor Stack alarm and sequence Ventilation Air Particulate Behind Console, 0-10 counts > 300 CPS Indication and Fission Products Fission Product Monitor Control -Room per second 'in Stack Alarm 4 mrem Primary Coolant Fission N Delay Tank 0.1-10 5 x normal Indication and Radiation in g ' Product Monitor reading Primary Coolant Alarm o NOTE: During irradiated fuel handling times, this setpoint may be increased. This increase will be to the lowest practical setting, as determined from past radiation surveys conducted during irradiated fuel movements, while allowing for irradiated fuel movements near the detector. The setpoint will be re- ! turned to its normal setting after completion of each days activities.
1 1 B) East reactor room area radiation monitor (area monitor), one detector.
- 1. G-M type detector with proportional compensation
- 2. Alarm setpoint; 5 mrea/ hour or less l
- 3. Initiates area over-tolerance alarm.
C) Reactor room ventilation stack radiation monitor (stack moni-tor), one detector.
- 1. G-M type detector with proportional compensation,
- 2. Alarm setpoint; 15 mrem / hour or less
- 3. Initiates building evacuation alarm and de-energizes reactor room ventilation.
D) Primary coolant fission product radiation monitor (primary monitor), one detector
- 1. G-M type detector with proportional compensation
- 2. Alarm setpoint; 5 times normal 100% power reading or less
.O 3- 1 ici te ri ie ored ct- 1 grim r7 ce t e 1 rm-E) Reactor room ventilation stack air particulate fission product monitor (APFPM), one detector
- 1. Scintillation detector
- 2. Alarm setpoint; 300 counts per second, or less
- 3. Initiates fission products in stack alarm F) All radiation monitors listed in this section (16.3.7.1) shall be required to be operating for all reactor operations, and for certain experiments, whers reactor operation is not required, as determined by the RSC. If any single monitor becomes l inoperable, except those activating a building evacuation l- alarm, appropriate suitable survey instruments, with similar
! radiation : monitoring capabilities to that of the inoperable instrument may be -substituted until completion of repair activities, or 30 minutes whichever is less, where upon continuation of activities shall require all instruments being fully operable. Whenever portable survey instrument substi- l I tution is required the following conditions shall be met: O 301
- 1. Surveys shall be 'one continuously during the repair; and, l 2. All results shall be logged by time, date, and type in the reactor conversational log book.
The building and stack monitors shall be in continuous opera-I tion except as required for maintenance or repair and then the conditions in section 16.3.91.B (building evacuation) must be met. i 16.3.7.2 Portable Radiation Monitoring , -There shall be at the VTAR facility operable portable radiation
- survey instruments available for measuring: i A) Beta gamma instrument capable of monitoring dose rates.from 0.1 meem/ hour to 20.0 rea/ hour; '
B) Neutron instrument capable of monitoring dose rates from 0.1 ares / hour to 10.0 ren/ hour; 1 C) Portable air sampler for collecting airborne activity; and, D) G-M type "frisker" detector (or other suitable detector) for determining airborne particulate activity levels and for personnel contamination monitoring.
- 16.3.7.3 Bases The capabilities of the radiation monitors is determined by antici-pated operations and accident scenarios. This also specifies certain
- functions that initiate measures to inform and alert operators, minimis-l' ing release paths, and protect the safety and health of the staff and public.
i The fixed monitor setpoints are determined in such a manner as to
- allow for operational transients, yet initiating alarms, actions, etc.
l in a prompt fashion, and thereby maximizing protection of both the staff i and the public. Portable monitor ranges allow for monitoring radiation levels, airborne levels, and contamination levels over .al?. expected operating conditions in addition to providing assurances of monitoring over the range of credible accident scenarios. i The primary coolant fission product monitor indicates gross fission ' product contamination of the primary coolant. Evidence of a fuel ; element failure would be indicated by an increase in primary coolant. i activity / radiation. At the setpoint the fuel failure could be detected and with a prompt shutdown, further release minimized. J i lO 302 l
16.3.7.4 Effluents To Unrestricted Areas 16.3.7.4.1 - M ' The following operational limits shall govern the discharge of Argon-41 to the environment: A) 1 The maximum exceed 1 x 10 r4 ease rate from the ventilation stack shall not Ci/second; _ B) The total release of Ar-41 shall not exceed 315 Ci/ year; v. C) TheconcentrationofAr-41releagedtounrestrictedareasshall not exceed the MPC of 4.0 x 10~ pCi/mi as cited in 10 CFR 20, Appendix B, Table II. 16.3.7.4.2 Other Effluents All other effluents shall not exceed the MPC limits as cited in 10 CFR 20 Appendix B, Table II. 16.3.7.4.3 Bases A radioactive effluent specifically monitored at the VTAR is Argon-
- 41. This radionuclide is produced by the neutron activation of Argon 40 in the air that is drawn into the core area. Fission product releases f rom the VTAR are not considered likely because of the limitations on excess reactivity and the extensive operating history of Argonaut reactors. However, fission product activity is monitored via primary and secondary coolant water samples and an airborne particulate monitor.
16.3.8 Limitations On Experiments-These specifications apply to all those experiments and experi-mental devices in the reactor core or its experimental facilities. The objective of this section is to maintain operational safety and prevent damage to the reactor facility, reactor fuel, reactor core and associ-ated equipment; to prevent exceeding the Reactor Safety Limits; and to minimize potential hasards from experimental devices. 16.3.8.1' Specifications A) General The Reactor Supervisor, Reactor Radiation -Safety Officer and the Nuclear Reactor Laboratory Director shall review and approve . in writing all proposed ' experiments prior ' to their performance. The Reactor Supervisor shall refer to the Reactor
. Safety Committee (RSC) the safety evaluations'of all new experi-ments/ irradiations or those which require the use of special O
303
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nuclear material (SNN). The Chairman of the RSC shall sign for approval af ter review by the RSC if applicable, in addition to those approvers mentioned previously. Adjustments or changes e to previously approved experimental irradiations shall require a reevaluation to determine if it requires RSC. review and
- approval.- When experiments / irradiations contain substances
' which irradiation in the reactor can convert into a material with significant potential hazards, a determination will be , made, by all reviewing bodies, of the acceptable reactor power, amount of material and length of time of the irradiation. This review shall take into account such factors as: isotope iden-tity, chemical and physical form, containment type, toxicity, , potential for contamination, problems in removal and handling, j transfer and eventual disposition. Experimental apparatus, 1 material or equipment inserted into the reactor shall be
- reviewed to assure non-interference with the safe operation of e the reactor.
1 B) Experimental Facilities ! Facilities for the insertion of experiments shall be > limited to:
- 1. Graphite thermal column (located on the west face)
- 2. Shield tank structure (located on the east face)
- 3. Pneustic rabbit systems
- 4. Beam ports (located on the north and south faces)
I 5. Removable graphite blocks in the central graphite l reflector region (located between core tanks) The thermal column door shall normally ~be locked, and access i permitted only when the reactor is shutdown (Mode IV) or under ; special circumstances which would require RSC approval. Pneumatic rabbit systems may be installed and these shall be operated with nitrogen (N2 ) gas, with the gas vented through a
. prefilter and a HEPA filter into the exhaust ventilation duct.
s C) Classification of Experiments t I Class I i Routine, previously performed sample irradiations using the pneumatic rabbit . system, not containing SNM. This class shall be reviewed for safety by the Reactor Supervisor, Reactor Radiation ~' Safety officer, the Nuclear Reactor ~ Laboratory Director, and an SRO. At a' minimum, three separate individuals must review the experiment. L 304
. .- = ~ - . - _. .-- -.-
y f Class II Rotttine, previously performed sample irradiations, not containing SNM, placed in the central graphite reflector region of the core. This class shall be reviewed by those persons l' approving a Class I experiment. Class III l Experiments, previously performed and not containing SNM, which are to utilize the enternal irradiation f acilities of the j VTAR. This class shall be reviewed by those persons approving class I experiments. Class IV 4 i Experiments or irradiations which pose questions regarding the safety of the reactor, the personnel or the public or are new experiments or contain special nuclear material. This . class experiment shall be reviewed and approved by the RSC, the ' Reactor Supervisor, the Reactor Radiation Safety Officer and the Nuclear Reactor Laboratory Director. D) Installation Of Experiments l O No experiment shall be installed in the reactor in such a j manner that
- 1. it could measurnhly affect the nuclear instrumentation .
- system monitors, or
- 2. failure of the experiment could interfere with the control i rod system, or;
- 3. failure of the experiment could credibly result in fuel element damage.
l E). Reactivity Limitations On Experiments
- 1. The reactivity worth of any single movable experiment in the reactor core or experimental facilities shall not '
exceed 0.3% ak/k.
- 2. Experiments having moving parts shall have reactivity insertion rates less than or equal to 0.05% ok/k/second.
- 3. When determining the absolute reactivity worth of- an -
experiment, no credit shall be taken for temperature , effects. l 305 i
- . ~ , , , ,.----er - - - - . , - , - = , , - - . , . - --.,.y, + v % - = . - - - - - - -
1 l t =
- 4. An experiment shall not be inserted or removed unless all control rods are fully inserted or the experiments l absolute reactivity worth is less than that which would j- cause a positive stable period of 60 seconds (0.092% Ak/k).
F) Materials i' 1. No explosive materials shall be irradiated. >
- 2. The amount of special nuclear material contained in an i experiment to be placed in the rabbit or central graphite reflector experimental facilities shall not exceed 1 gram.
, 3. No experiment shall be performed involving materials whfah could: , a) credibly contaminate the reactor coolant systems-l causing corrosive action or reactor components or
- experiments; b) cause excessive production of airborne radioactivity, exceeding -25% of 10 CFR 20, Appendix B, Table 11; or, c) produce violent chemical reactions under the conditions of the experiment.
l 16.3.8.2 Bases , The general specifications assure that an adequate review process j is followed to determine the safety, conditions, and procedures for 'all experiments. The experimental facilities' defined, are the only places ! where an insertion of an experiment asy occur. The classification of - experiments delineates the responsibility for approving ' experiments according to their potential hazards; to assure that potentially hazardous experiments are analysed for their , safety implications; and i that appropriate procedures are established for their execution. The i installation of experiments details how an experiment any be installed l in order to prevent obstruction of safety related equipment and The reactivity limitations on experiments are established -{
~
instruments. to prevent prompt criticality by limiting the worth' of movable experiments; to prevent a reactivity insertion larger than the stipulated maximum step reactivity insertion in the Accident Analysis, and to allow for reactivity control' of experiments within the reactor i control system capabilities (60 second positive period limitation). These specifications prohibit the irradiation of explosive materials and ' limit the amount of fissile materials that can be irradiated in the reactor. Explosive materials are defined 'as those asterials normally j used to produce explosive or detonating effects, materials which can lO k~ 306
. . . . . , . _ , . ,,, - _ - ~ , - _ _ - . , _ . - ~ - - . . . - - , . - _ - - , - . , - ,m.m - , , , m - -. - _-._,,...-._.%%,, . . _ - . - - . -
I chemically combine to produce an explosion or detonation, or any ; material' which can undergo explosive decomposition under the influence - of neutron, gamma or heat flux of the reactor, or as defined by applicable standards. 16.3.9 Building Evacuation System f
.These specifications apply to the equipment required for the f receipt and initiation of a building . alarm for the evacuation of l personnel from the reactor facility and the reactor - building - (Robeson !
Hall). These specifications shall list the minimum . requirements necessary to initiate an evacuation alarm and also the actions to be taken in the event the system must be inoperable for any period of time, i (i.e. maintenance or repair). ! I 16.3.9.1 Specifications All building evacuation alarms received shall be treated as actual alarms, no matter what the cause, until verified by a Senior Reactor f Operator and the Reactor Radiation Safety Officer or his designate. ! I A) The reactor f acility and . reactor building shall be evacuated I when any of the following conditions exists ( 7
- 1. A building monitor radiation alarm is received (greater than or equal to 15 ares /hr) or the building evacuation i alarm is received; !
i
- 2. A ' ventilation stack monitor radiation alarm is . received I (greater than or equal to 15 meen/hr) or the building .!
evacuation alarm is received;- l
- 3. When the reactor operator detects a potentially hasardous l radiological condition, where preventive actions are !
required to protect the health and safety of.the operating l personnel or the public,-and manually initiates'the evacu-ation alarm. [ i B) The building and stack radiation monitors shall be in contin-uous operation. If ' either radiation -monitor becomes inoper-able, the reactor will be shutdown. 'After the shutdown activities, the ' substitution of portable instruments etc. may continue for a period that shall not exceed 48 hours to allow for troubleshooting and repair. In this case the. following , conditions must be mett x )
- 1. All reactor room activities shall be secured.
- 2. Reactor room ventilation shall be secured.
.m 307 A- ;
6 n
l O 3. During normal working hours surveys shall be done on a two hour cycle, l
- 4. After normal working hours the VIAR facility shall be secured and entry permitted only to those personnel required to repair the af fected equipment. - At no time shall activities take place which could create a situation where radiation or contamination levels could increase above the normal shutdown levels.
16.3.9.2 Bases To provide early and orderly evacuation of the reactor facility and reactor building in order to minimize radioactive hazards to the operating personnel and general public. 16.3.10 Fuel And Fuel Handling And Storage These specifications apply to the arrangement of fuel in core and in storage, as well as the handling of fuel elements, and they also establish the maximum core loading and storage loading for reactivity control purposes. Additionally, these limitations establish fuel handling criteria with regard to radiological safety considerations. 16.3.10.1 Specifications A) The maximum fuel loading shall consist of 12 full fuel elements , each consisting of 12 fuel plates containing enriched uranium and clad with high purity aluminum.
- B) Fuel element loading and distribution in the core shall comply with fuel handling procedures.
I C) Fuel elements exhibiting release of fission products due to cladding rupture shall, upon positive identification, be perma-nently removed from the core. Fission product contamination of the primary coolant water shall be treated as evidence of a fuel element failure. D) The reactor shall not be operated if there is evidence of a fuel element failure. I E) All irradiated .(hot) fuel shall be moved and handled in accordance with approved procedures. In addition, all hot fuel movements shall be conducted, with the reactor shut down, by a minimum staff of three persons which shall include a licensed senior operator and a health physicist. The staff members shall monitor the operation using appropriate portable radiation monitoring equipment and all fixed radiation monitoring equiment listed in section 16.3.7.1. with the i O. c eci f =* ert arr t it -
'l l
308
F) All ?sel movements shall be entered in the reactor conversational log. G) Fuel elements or fueled devices shall be stored and handled out of core geometry such that k,gg is less than 0.70 under optimum conditions of moderation and reflection using light water. H) Two fuel storage pits shall % used for storing fuel elements or fueled devices and shall te capable of storing 16 fully loaded fuel elements each.
- 1) A fuel transfer cask shall be utilized to transfer single hot fuel elements to and from the core and storage pit locations.
K) Hot fuel handling shall not commence earlier than 96 hours after the last reactor shutdown. L) Irradiated fuel shall be stored in the west pit unless specifically permitted otherwise by RSC. M) A maximum of one full irradiated fuel element shall be allowed in transition at any one time. 16.3.10.2 Bases The fuel loading is based on the present fuel configuration. The reactor systems do not have adequate engineering safeguards to allow operation with a detectable inventory of fission products in the primary coolant. The fuel is stored in a safe configuration and shall be handled so as to prevent any violation of radiological contamination and exposure limits. Commencing fuel handling operations 96 hours after the last reactor shutdown allows for a large part of the core activity to decay and thus decreases the exposure and safety hazards to personnel involved. v 309
+
~- . . - -
16.4.0 SURVEILLANCE REQUIREMENTS l The requirements listed in this section prescribe the frequency and scope of tests used to determine the performance of the specifications ! listed in section 16.3, Limiting Conditions for Operation, in order to periodically verify satisfactory performance of all equipment required for safe operation. 16.4.1 General Requirements And Surveillance Intervals 16.4.1.1 Surveillance Intervals Allowable surveillance intervals shall not exceed the following: A) Biennial (interval shall not exceed 28 months) B) Annual (interval shall not exceed 15 months) C) Semiannual (interval shall not exceed 7 months) D) Quarterly (interval shall not exceed 4 months) E) Monthly (interval sh&ll not exceed 6 weeks) () The maximum intervals are to provide for operational flexibility and not to reduce frequency. Established frequencies shall be maintained over the long term. Surveillance requirecents specified in this section, (with the exception of those specifically tequired for safety when the reactor is shutdown; see section 16.4.13 shutdown surveillance require-ments) may be deferred if during the specified surveillance period, the reactor has not been brought critical or is maintained in modes II, III, or IV, extending beyond the specified surveillance period. However, the
- surveillance requirements must be met prior to subsequent start up of the reactor.
16.4.1.2 General Requirements I A) Whenever an unscheduled shutdown occurs--not due to a power loss--an evaluation shall be made to determine whether a safety limit was exceeded. This shall be noted in the unscheduled Jhutdown log. B) In the event a LSSS actuation occurs, and is not due to equip-ment failure, an evaluation shall take place. This evaluation shall consist of a re-evaluation of the affected LCO by the reactor staff and RSC to determine if the initial safety anal-ysis was correct and to set criteria that prevent a reoccurance of the actuation. O 310
I C) A reactor start-up is considered to have started at the time the dump valve is closed and filling of the core tanks commences. 16.4.2 Reactivity And Core Parameters 16.4.2.1 Excess reactivity shall be determined annually or whenever section 16.4.2.3 (Contol rod reactivity worths) is performed. 16.4.2.2 Shutdown margin shall be determined annuelly or whenever section 16.4.2.3 (Control rod reactivity worthe) is performed. 16.4.2.3 Control rod reactivity worths shall be determined whenever: A) A control rod is removed from the core; or, B) A control rod is replaced; or, C) Operation requires a re-evaluation of core physics parameters; or, O => i first. 117 er - c tr. whica occ r-16.4.2.4 control rod reactivity insertion rates shall be required annually or whenever Section 16.4.2.3 (control rod reactivity worths) is performed. 16.4.2.5 Experiment reactivity worths shall be estimated or 4 measured, as appropriate, before or during the first start-up, prior to conducting that experiment. 16.4.2.6 Coefficients of reactivity (VOID and TEMPERATURE) A) Void coefficient of reactivity shall be measured during initial reactor criticality or when core physics parameters require re-evaluation. B) Temperature coefficient of reactivity shall be measured biennially -or when core physics parameters require re-evaluation. 16.4.2.7 The ability of the neutron source to produce an indication of 3 cps or greater on the source range O G 311
b instrument shall be checked prior to each start-up. 16.4.3 Reactor Control And Safety 16.4.3.1 Control rod d.op times shall be measured semi-annually or whenever: A) A control rod is removed from the core temporarily; 4 B) A new control rod is installed in the core; or, C) Modifications or repairs are done on the control rod or its drive mechanism. 16.4.3.2 Control Rod drive mechanisms shall be tested and inspected semi annually, whenever modifications or repairs have taken j place. 16.4.3.3 Control rod position indication and member operability shall be checked during each reactor start-up. 16.4.3.4 Reactor start-up interlock circuitry shall be tested semi-annually, fo11owing modification or repair. 16.4.3.5 Regulating rod automatic ' control circuitry shall be j tested semi annually. following modification or repair. 4 16.4.3.6 Channel calit, rations, where applicable, shall be 1 performed annually or following modification or repair, on all functions
- or channels listed in Table 16-6. >
'16.4.3.7 A channel operability test of the power, intermediate and appropriate source range channel (s) shall be performed prior to each reactor start-up unless the latest shutdown occurred less than 4 hours before the reactor start-up.
16.4.3.8 A channel operability test shall be performed on the following instruments / circuits prior to each reactor start up: l. A) All radiation monitoring instruments B) Keithley reactor power instrument (s) C) Primary coolant conductivity instrumenti D) Annunciator instrument panel (indication only). 16.4.3.9 Scram channel operability checks shall be performed 1 quarterly, at the beginning of each startup on a rotating basis, 'or 4 following repair or modification. i. 312
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p TABLE 16-6 Channel Calibration Requirements , l N Channel / Function Type Calibration t i
-1 Source Range Channel #1 Drawer Calibration ,
k .' c
- 2 Source Range Channel #2 Drawer Calibration !
3 Intermediate Range Channel Drawer Calibration i ! t ! 4 Power Range Channel #1 Drawer Calibration ! l
- 5 Power Range Channel #2 Drawer Calibration I
', - 6 Keithley Ammeter Channel.41 Drawer Calibration } 7 Keithley Ammeter' Channel #2 Drawer Calibration I . 8 Keithley Channels # 1 and # 2 Reactor Power Flux Calibration 3 i ; Thermal Power Calibration j 9 Power Range Channels # 1 and # 2 i i I !. O . I ! 4 . e ) i i t 4 4 kg e d 1' i i s 3 i i l l M
,I i
t- ' 1 , 4' t ! [O 313 ]
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O 16.4.3.10 Thermal power verification shall be required annually or following any modification or repair to the measuring channel (s) or any situation which requires re-evaluation of core physics parameters. 16.4.3.11 Calculated thermal power and indicated reactor power shall be compared hourly during each reactor run when power is above 10 kW with " steady state" conditions present for an hour. { 16.4.4 Coolant Systems 16.4.4.1 Primary Coolant System , A) Primary conductivity cell and circuit calibration shall be performed annually or following any modification or repair. B) Make-up water domineraliser bed shall be replaced when the system can no longer maintain primary coolant water conductiv-ity specifications. C) Dump valve operability check shall be performed prior to each reactor startup or following any modificacion or repair. D) Primary temperature instrument calibration shall be done semi-annually or following any modification or repair O =) Primar, coetant f1ow ca11bration sha11 be done semi-annua 11y er following any repair or modification. F) Primary coolant pump in-service inspection shall be done annu-ally. G) Primary to secondary heat exchanger in-service inspection shall be done annually. 16.4.4.2 Secondary Coolant System A) Cooling tower inservice inspection shall be done semi-annually. B) Secondary throttle -valve position indication calibration shal1 be done semi-annually or following any modification or repair. C) Secondary coolant temperature calibration shall be done semi-annually or following any modification or repair. ' D) Secondary coolant flow calibration shall be done semi annually or following any modification or repair. E) Secondary coolant pump in-service inspection shall be done annually. 314
- - F) Optimization of secondary coolant flow by adjustment and recording of tower throttle valves shall be performed semi-annually, or as nacessary.
16.4.4.3 Emergency Core Cooling System NOTE: Sections A, and C are included in anticipation of a future power upgrade and are not applicable for a maximum power of 100 kW. A) Boron injection tank water inventory shall be checked prior to each start-up to ensure minimum water inventory is maintained. , B) ECCS flow calibration shall be performed semi-annually or following any modification or repairs to the ECCS. C) Boric acid inventory and location shall be checked prior to each reactor start-up. , D) ECCS coolant pump shall be tested for operability ac,nthly or following any modification or repairs. i E) ECCS coolant isolation valve shall be tested for operability monthly or following any modification or repairs. F) ECCS coolant pump in-service inspection shall be done annually. O o> sccS cee1 ant isotacion va1ve in-service inseectien sha11 be done annually. H) ECCS valve alignment shall be verified prior to each reactor s tar t-up. 16.4.5 Confinement } A functional test of confinement boundaries shall be done annually or following any modification or repat6s. 16.4.6 Reactor Room Ventilation A) A ventilation integrity check shall be done annually or follow-ing any modification or repairs. B) A test of the automatic ventilation securing circuitry shall be done semi annually. C) A reactor room negative pressure check'shall be performed prior to each reactor start-up. D) An in-service inspection on reactor room ventilation fans shall be done annually. 315
l O 16.4.7 Emergency Power A functional transfer and load test shall be done quarterly or following any repairs or modification. A commercial power outage, during normal working hours, with proper operation longer than 15 minutes, may cour*. as a successful test. 16.4.8 Radiation Monitoring And Effluents 16.4.8.1 Radiation Monitoring A) An operability check of all fixed and portable radiation moni-toring equipment, not including personnel dosimetry, shall be done prior to each reactor start-up. B) The reactor room area, building, stack, primary coolant fission product, and air particulate fission product monitors shall be calibrated semi-annually. This calibration will check the capability to automatically secure reactor room ventilation and the proper functioning of the key reset feature. 16.4.8.2 Effluents i A) Argon-41 stact, monitoring shall be done semi-annually, while the reactor is at power. l B) Samples of primary and secondary coolant shall be analyzed quarterly, for gross 8 y activities. C) Argon-41 monitoring in unrestricted areas shall be done quarterly. 16.4.9 Experiments Specific. surveillance requirements shall be established during the review and approval process as specified in 16.6.3, Experimental Review and Approval, and as such are not specifically detailed here. 16.4.10 Building Evacuation System Building evacuation drills shall be performed quarterly. These drills may be performed as part of the emergency plan drills or exer-cises, or as part of regularly scheduled maintenance (i.e., following channel calibration). 16.4e11 Fuel And Fuel Handling And Storage A) The physical integrity of all control rods shall be inspected annually. O 316
- O- B) Irradiated fuel shall be inspected biennially in a random pattern. At least two elements shall be inspected.
- C) The fuel transfer cask operation and integrity shall be checked
- prior to. irradiated fuel handling operations.
1 ,
. D) Applicable fuel storage pits shall be inspected for water or
]- other. foreign materials prior to fuel handling operations. E) Fuel handling tools and procedures shall be reviewed for ade-quacy prior to irradiated fuel handling operations. The 4 assignment of responsibilities and training simulation shall be performed prior to irradiated fuel handling operations. 16.4.12 Reactor Components And Structure ! A) The reactor room bridge crane shall be inspected annually. B) The primary shield tank shall be inspected biennially. If excessive corrosion or other damage is apparent, corrective action will be taken prior to reactor operation above 10 watts.
- C) The pneumatic rabbit system shall be inspected and leak checked l i
quarterly or prior to use, whichever is greater; or following modification or repair. O 16.4.13 Shutdown Surveillance Requirements l t
, The following surveillance requirements shall not be deferred due i
to reactor status and as such are required to be performed on ' the
! intervals listed.
t-l 16.4.13.1 Reactor Room Ventilation A) A test of the automatic ventilation securing capability shall i be done semi-annually.
- 16.4.13.2 Emergency Power L
A) A functional transfer and load test shall be done -quarterly or j following any repairs or modification. A commercial power
- outage, during _ normal working hours, with proper operation
, longer than 15 minutes, may count as a successful test. 16.4.13.3 Radiation Monitoring
, i A) The reactor room Area, Building. and Stack monitors shall be !
i - calibrated semi-annually. This calibration will check the i j capability to automatically secure reactor room ventilation and ' l the proper functioning of the key reset feature. 317 f
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O- 16.4.13.4 Building Evacuation System i A) Building evacuation drills shall be performed quarterly. These j drills may be performed as : part of emergency plan drills or j exercises, or as part of regularly scheduled maintenance (i.e., following channel calibrations). 16.4.13.5 Fuel, And Fuel Handling A) The physical integrity of all control rod mechanisms shall be inspected annually. B) Irradiated fuel shall be inspected biennially in a random pattern. At least two elements shall be inspected. 4 l O a j f i i I O 4 318
I
~h 16.5.0 DESIGN FEATURES These features are described in order to assure that major alter-ations to safety related equipment or components are not performed prior to all appropriate safety reviews.
16.5.1 Site The Virginia Tech Argonaut Reactor (VTAR) is located in Robeson Hall (Reactor Building) of the Virginia Polytechnic Institute and State University. The campus is located in the town of Blacksburg, Virginia, in Montgomery County. Robeson Hall is situated in the Northwest corner of the main ca: pus. Buildings located within a 500-foot radius include Davidt on, Williams, Derring, Cowgill, and Pamplin Halls. 16.5.2 Reactor Facility The reactor room shall be of reinforced concrete construction with approximate dimensions of 40 feet by 40 feet and 22 feet floor to ceil-ing. A viewing window is provided on the first floor of the reactor building near the control console. 16.5.2.1 Restricted Areas The reactor room (room 10), the adjacent rooms . (room 6, and 8.), O# room 106 and the building roof are restricted areas. These spaces shall be utilized and posted in such a manner as to conform with 10 CFR 20. 16.5.2.2 Reactor Room Ventilation The reactor room ventilation shall be discharged abovoi the roof level of Robeson Hall by two exhaust fans. The main exhaust fan shall take its suction from the . reactor room ventilation duct' while the booster exhaust fan, mounted directly above the exhaust outlet of the main fan, shall mix and dilute the main exhaust and boost the exhaust higher into the atmosphere for~ better dispersion. All gases which may cause a hazard from pneumatic Rabbit System neutron activation analysis shall be discharged via the reactor room ventilation system. 16.5.2.3 Reactor Bridge Crane The reactor bridge crane shall not be operated during reactor operation in a manner that could pose a threat to the control rods, experimental facilities, or reactor core. Crane operat'on (during; reactor operation) shall be performed only by a ' licensed reactor oper-ator or senior operator. O 319
- v. .- . - - . _ - - - . - - -- . - .- . . - -- .- - -. =_ - . -
i pt 16.5.3 Reactor Coolant Systems 16.5.3.1 Primary Coolant Sys tem, 1 g Core cooling is - affected by forced input, gravity return flow of m demineralized water f rom parallel piped core tanks to a storage (dump) tank, located below ground level in a process pit that provides for N-16 I gamma decay and acts as the primary coolant water inventory reservoir. Water is then pumped back to the core tank inlet through the primary y side of a heat exchanger where heat is transferred to the secondary
- h ., cooling system. . A purification / deionization loop is used to maintain
' ' ' ~
- primary water purity. This purification loop circulates approximately 3 gpm of primary water drawn - from the discharge of the primary coolant
- - pump, through a heat exchanger, filter and mixed ion-exchange resin bed, !
and returned to the dump tank. The duinp tank is vented to the reactor room ventilation duct to facilitate Ar--41 removal. A 6-inch dump-inlet i line leads from the dump tank to the bottom of the core tanks. . Coolant j- is drained (dumped) from the , reactor core tanks by de-energizing an electro-magnetic clutch which allows the spring opened dump valve to i open, thus by passing the core tanks in the coolant loop. Primary flow is measured by a flow detector placed in the inlet line of the- primary ' to secondary heat exchanger and gives flow indication at the control ! console. Temperature sensors in -the 2-inch core inlet and 4-inch core ! outlet (overflow) piping allow the core temperature differential to be , measured. Both local and' remote (indication at the control console) , temperature indications are available. A primary conductivity probe, j - indicating primary conductivity at the control console, is installed in ; i the purification loop upstream of the filter and de-ionizer. I ' Annunciators at the control console shall provide indication of primary coolant system abnormalities. - 16.5.3.2 Secondary Coolant System { Reactor heat transferred through the primary to secondary haat j exchanger is dissipated to .the atmosphere using a cooling tower on the i \, reactor building roof. To minimize . corrosion, chemistry control is { being used. To prevent water fron ' entering the secondary system should the primary to secondary integrity be broken, the' static pressure of the l 3 secondary is maintained higher than that of the primary. All secondary coolant piping is 6-inch in diameter. Secondary coolant flow is v measured by a flow detector located in the secondary ' inlet to the i ; primary ' to secondary heat exchanger, with indication at the control 3' console. Secondary temperature detectors, giving indication at the ! control console, are located in the inlet and outlet lines from the primary to secondary heat exchanger which allow the temperature differ-ential of the heat exchanger - and cooling tower to be measured. An + additonal temperature detector, which also gives indication at the
- . control console, is located ~ in the cooling tower pan. Annunciators at i the control console shall provide indication of secondary coolant system
^ abnormalities. L O j 320 $ T ~ . - . - . - - - . - . - _ - - - - . - - - - - --- ,. . - - _ . _ -
l l l 16.5.3.3 Emergency Core Cooling System t As a back-up in the event of a casualty situation an emergency core cooling system (ECCS) is provided. The ECCS is not necessary for any reactor power levels less than 100 kW and for this licensing will only be used in a training and testing capacity, but because a power upgrade is anticipated it is addressed. Should a casualty occur in which emerg-i ency cooling is needed, the ECCS is capable of shutting down (via Boron Injection) the nuclear chain reaction and supplying a means of decay heat removal. Reactor heat is transferred from the core using primary l coolant uater and dissipated via an ECCS heat exchanger, using city water as the secondary coolant. To afford reactor shutdown capability a boron injection tank (BIT) is installed and feeds into the inlet of the ECCS pump. ECCS flow is measured and indicated at the console. The same temperature monitors used for the primary coolant system are available for use with the ECCS due to their location in the primary piping. A storage container for boric acid is near the BIT for uses requiring the ECCS reactor shutdown capability. An annunciator shall give visual and audible indicatiro that the ECCS is in use at the control console. 16.5.4 Reactor Core And Fuel 16.5.4.1 Reactor Core The two reactor core tanks are open to the atmosphere, imbedded in O a reactor grade graphit.e lattice with the approximate dimensions of 3-2/3 feet by 4-2/3 feet with a height of 4 feet. The core tanks are 4 separated by approximately 18 inches of graphite and are 2 feet apart on centerline. The core shall be covered by concrete and steel top and bottom operating and shutdown closures at all times when_ fuel is in the 4 core except when fuel handling operatione are being performed. During fuel handling operations the shutdown closures must be in place. After removal of the top and bottom perating closures the reactor shall not be left unattended while fuel is still in the reactor unless the fuel element plugs. in the bottom closures are secured in such a way as to - prevent their removal and that of ' the fuel elements. When the core is devoid of fuel the bottom closures shall be maintained in place except
~
when incore maintenance or testing is being . performed. The core shall not be left unattended when any bottom closures are removed. In the event special measurements are being performed such as flux profile determinations at power levels less than or equal to 10 watts operation with only:the-shutdown closures shall be allowed with RSC~and Radiation Safety approval. The reactor. core shall contain 12 fuel elements, 6 to each core ! tank, and 'each fuel element shall contain 12 fuel plates such that all j fuel positions are occupied. Dummy, quarter, and half load ' plates may ; be used to adjust core loading. j l 321 e l
l l J 16.5.4.2 Reactor Fuel Fuel elements shall be of the general MTR type, with fuel plates clad with aluminum on each side and containing uranium fuel enriched to approximately 93%. The fuel matrix may be fabricated by alloying high purity aluminum uranium or by the powder metallurgy method where in the starting ingredients (aluminum-uranium) are in fine powder form. Fuel elements shall conform to the following nominal specifications: overall size 5.55 in x 3 in x 26 in (excluding lifting attachment) clad thickness: 0.020 in plate thickness: 0.080 in water channel width: 0.400 in number of plates: 12.0 per element , Fuel Content: 22.0 gm, U-235 per plate plate attachment: bolted or pinned, utilizing spacers. Upon receipt from 'the fuel vendor, all . fuel elements shall be visually inspected and the accompanying quality control documents checked for compliance with specifications. Each new fuel element shall be inspected for damage and flow obstructions prior to it:artion into the core. 16.5.4.3 Fuel Transfers
- Fuel transfers shall be done with reactor shut down for 96 hours or longer, by a minimum staff of three persons. Actual fuel handling i
i shall be performed by a senior reactor operator. All fuel transfers shall be conducted using appropriate radiation monitoring and shall be entered in the Resctor Conversational Log. Fuel transfers will use a Fuel Transfer Cask with sufficient shielding such that radiation levels are as low as reasonably achievable and in compliance with 10 CFR 20. The Radiation Monitoring system shall be fully operable during any fuel transfer, the primary coolant monitor may be excluded from the requirement. During fuel transfers only licensed operators shall operate the reactor room bridge crane.
- 16.5.5 Fissionable Material Storage Two fuel storage pits shall be located in the Reactor Room. Each shall be capable of storing 12 fully loaded fuel elements. Each fuel element shall be stored in a separate cylindrical hole within the stor-age pit and each shall have adequate shielding. Fuel elements, includ-ing fueled experiments, shall be stored and handled- (out of the core) in a geometry such that, under optimum conditions of moderation and reflec-tion, K is less than 0.70 (using light water). A maximum of 13 fully loaded ,fgplates will be allowed in each separate storage hole.
10 J
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r 16.6.0 ADMINISTRATIVE CONTROLS 16.6.1 Organization 16.6.1.1 Structure The organization for the management and operation of the VTAR facility shall include the structure shown in figure 16-2. The job titles are shown for illustration and may vary. Five levels of author-ity are provided as follows: A. Level 1 - Individuals responsible for the reactor facility's licenses, charters and site administration. B. Level 2 - Individuals responsible for reactor facility manage-ment. C. Level 3 - Individuals responsible for reactor operation, and supervision of day-to-day facility activities. D. Level 4 - Senior operating staff (class B and class A reactor operators). . E. Level 5 - Ceneral operating staff (reactor operator trainees and administrative employues). F. The VTAR Reactor Safety Committee is appointed by the Vice-President for Administration and Operations and reports to the Chairman of the Univeristy Radiation Safety Committee. The Chairman of the University Radiation Safety Committee reports ' to the Vice-President for Administration and Operations. Radiation Safety Office personnel report to the Director of Health and Safety Programs (S.H.P.). 16.6.1.2 Responsibility The responsibility for the safe operation of the reactor facility shall be with the chain of command established in Figure 6-2. The facility shall be under direct supervision of the Reactor Supervisor or
- t. Licensed Senior Reactor Operator designated -by him to be in direct control. The Reactor Supervisor shall be responsible to the Director of the Nuclear Reactor Laboratory (NRL). In all matters pertaining to reactor operation and these technical specifications, the Reactor Super- l visor and NRL Director shall be responsible to the VTAR Raactor Safety Committee. For administrative matters the NRL Director shall be responsible to the Department Head of Mechanical Engineering.
l i i 323 i l
President VPI&SU Level 1 Provost l Vice-President for Administration and Operations 1 I Dean l - College of Engineering Radiation Safety ---- Director i Committee S.H.P, Department Head Level 2 Mechanical Engineering Reactor Radioisotope I-~ Safety Committee l Committee Director l l l Nuclear Reactor ---{ l g Laboratory (NRL) l Audit l l l_ _ _ Sub- l l Committee 1 I o - Reactor -- l l l
- ~ ~ ~
Radiation Supervis r Safety Office Level 3 (NRL) - - - - - - - - - - - - - - - - - - - Operating Level 4 Personnel Operator Administrative Level 5 Trainees Personnel VTAR ORGANIZATIONAL STRUCTURE Figure 1G - 2 l t Ot
- 'L) 324 i
i
r t 16.6.1.3 Staffing A. The minimum staffing when the reactor is in Modes I or 11 shall be: (1)- A certified reactor operator at the control console (operator in charge, 0.I.C.). (2) A second person (designated as reactor assistant) present at the facility complex able to carry out orders given by the 0.I.C. * (3) A designated class A reactor operator shall be readily available on ' call. "Readily available on call" means an individual who (A) is able to be rapidly contacted by phone or other communications, and (B) is capable of getting to the reactor facility in a reasonable amount of time (e.g.,'30 minutes or within a 15 mile radius). B. A list of facility personnel by name and telephone number shall be readily available for use by the OIC. The list shall include: (1) Reactor Staff personnel s (2) Radiation Safety personnel 4 (3) Other operations personnel (e.g. police, fire and medical-personnel) C. Events requiring the direction of a class A reactor operator: ' (1) All fuel or control rod relocations within the reactor core region;.or, (2) Installation or relocation of any experiment; or, (3) Recovery from unplanned shutdowns; or, (4) Any condition not in conformance with normal conditions. l 16.6.1.4 Selection And Training Of Personnel L The' selection, training and requalification of operations personnel shall meet or: exceed the. requirements of American National Standards for
- Selection and training of Personnel for Research Reactors, ANSI /ANS 15.4-Sections 4-6 and Code of Federal Regulation .part 10, Chapter -55, and 10 1'
CFR 55 Appendix A. In all case's for selection of a reactor operator trainee the ultimate decision shall be with the Reactor Supervisor. y ;
- j y-325-1
16.6.1.5 Audit And Review 1 A method for the. independent review and audit, of reactor ! operations and safety related administration shall be established to advise the reactor staff and improve the quality of operations. The ! review and audit of the VTAR are conducted by the Reactor Safety Committee (RSC) and the Reactor Safety Committee Audit Sub-committee. I 16.6.1.5.1 Composition And Qualifications l The RSC shall be composed of 9 voting members, including the Reactor Supervisor, NRL Director, Reactor Radiation Safety Officer, and l the Director S.H.P., (all ex-officio voting members), the chairman of the RSC, and other members having technical expertise in reactor - operations or design. One of the other members nust be from outside the university. The Audit Subcommittee shall consist of three persons designated by the RSC and must have background in the areas to be audited. J 16.6.1.5.2 Charter And Rules The review and Audit functions shall be conducted in accordance i with the VTAR Charter of the Reactor Safety Committee (Appendix 13A VTAR SAR). The following is an overview of the charter.
- O ci> Desi n tion - the name ef the committee is:
Committee, the abbreviated form is RSC. Reacter Safety (B) Accountability - the RSC reports to the Vice-President for Administration and Operations or in cases requiring special review to the University Radiation Safety Committee (abbreviated URSC). (C) Purpose - che VTAR RSC has been established with the authority to advis.a and regulate the safe operation of the Virginia Tech Argonaut Re:ctor, through the. discharge of the Review and Audit functions. 1 (D). Meetings (1) Meetings shall be called not less than once per calendar quarter--not~ to exceed four months and more frequently if necessary. (2) A quorum shall . consist of the chairman or alternate and one half the total members of the RSC.- (3) Minutes of the RSC meetings are recorded and furnished, in a timely manner,.to members of the RSC.
- O A
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(E) Review Function The RSC review shall include but not be limited to the following: (a) changes to procedures, equip-(1) All safety evaluations: l ment or systems, (b) tests or experiments conducted, to l . verify that such actions do not represent an unreviewed I safety question. 1 (2) Proposed changes to procedures, equipment or systems that change the original intent or use, or those that involve an unreviewed safety question. (3) Propoud tests or experiments which are significantly different,from previously approved tests or experiments or those that involve an unreviewed safety question. (4) Proposed changes in Technical Specifications or facility licenses. (5) Violations of applicable codes, regulations, orders, technical specifications, license requirements or of j- internal procedures or instructions having nuclear safety significanee. (6) Significant operating abnormalities, or deviations .from normal and expected performance of facility equipment that affect nuclear safety. (7) Events which require written reports to the NRC (8) Audit and operations reports (9) -Qualifications of prospective staff members (F) Audit Function The Audit function shall include selective (but comprehensive) examination of operating logs, records and other documents. . Where
- necessary, discussion with cognizant personnel shall take place.' in no case will personnel responsible for the items being audited, solely perform the audit. The following items shall be audited:
(1) Special Nuclear Material (SNM). Audited by the Reactor Supervisor, Reactor Radiation ' Safety Officer and one member of the Audit Subcommittee. (2) Reactor Operations, Administration, and Training; (3) Security and Emergency Planning; , -(4). Safeguards. 327
4 4' O (_j Any deficiencies found are forwarded to the operating staff and the RSC in a written report prior to the next regularly . scheduled meeting of the RSC. Deficiencies uncovered which directly affect reactor safety are reported to the chair of the RSC immediately. 16.6.2 Procedures The facility shall be operated and maintained in accordance with approved written procedures. All procedures and major revisions thereto shall be reviewed and approved by the VTAR RSC prior to becoming effec-
.tive. , 16.6.2.1 Substantial changes in procedures shall be made effective ,
only after approval by the VTAR RSC. New procedures shall be implemented only after approval by the RSC. . Previously approved procedures may be changed by using an interim procedure. This interim procedure shall have a safety analysis attached and. shall be reviewed by all available operators, and the Reactor Radiation Safety officer. Disapproval by any reviewing person shall . be reason for not implementing the interim procedure prior to RSC review. After favorable review, the interim procedure may be implemented by the Reactor- Supervisor. I 16.6.2.2 The following types of written procedures shall be main-tained: i A. Normal Operating Procedures (OP) (1) Reactor Start up, operations and shut down. These procedures shall include applicable check-off lists and instructions. (2) Routine Experiments Procedures (a) Approach to critical, Reactivity worth and rate measurements, and incore flux profiles. (b) Temperature coefficient measurements for fuel, moderator, reflector and void. (c) Administrative controls for operation and use of experi-mental irradiation facilities. (3) Infrequent Operations Procedures
'(a) Fuel loading, handling, and transfers (b) Fuel inspections 4
(4) Surveillance. tests required by these. technical specifications. O 328
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- 3. Maintenance Procedures (MP)
(1) Routine maintenance of equipment and systems having an effect on reactor safety. (2) Calibrations required by these Technical Specifications. C. Emergency Procedures (EP) (1) operator actions required in the event of a specific emergency or malfunction of the reactor and/or associated systems. (2) operator actions required in the event of fire or high radiation levels in the facility, including guidance as to when the procedure is to be initiated. D. Security Procedures (SP) Physical security and emergency plan implementation applicable to a class B operator. E. Radiation Procedures (RP) (1) Personal Radiation protection and guidelines consistant with applicable regulations. i O (2) Radioactive material transfers, handling, and storage. F. Administrative Procedures (AP) (1) Procedural guides (2) Reporting requirements and reactor run authorizations (3). Other state and local guidelines (4) Training (5) Audit and Review ,
-l 16.6.3 Experiments Review And Approval '
16.6.3.1 Experiment review and approval shall be conducted as specified under section 16.3.8 - Limitations On Experiments of these technical specifications. 16.6.3.2 The experiment review and approval shall assure compliance- with the requirements of the license, technical
- specifications and applicable regulations and shall be documented.-
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I i 4 16.'6.3.3 Significant changes to previously approved experiments shall be made effective only af ter approval by the VTAR RSC. Minor j changes that -do .not. significantly alter the experiment may be implemented af ter approval by the Reactor Supervisor, Reactor Radiation Safety Officer and the Nuclear Reactor Laboratory Director. L 16.6.3.4 Approved experiments shall be carried out in accordance with established approved procedures. 16.6.4- Required Actions 16.6.4.1 Actions to be taken in case of Safety Limit violations: (a) The reactor will be shut down and reactor operations shall not i resume until authorized by the NRC.- (b) The safety limit violation shall be reported to the Reactor Supervisor and the chair of the' RSC. P I (c) The safety limit violation shall be reported to the NRC and . shall include the following: (1) applicable circumstances leading to the violation including (when known): (a) the cause
'O (b) the contributing factors.
(2) Effect of the violation upon reactor components, systems { or structures and on the health and safety of personnel , and the public.. i (3) _ Corrective actions.to be taken to prevent reoccurrence. (d) Applicable' actions as set forth in Section 4.1.2 of these Technical Specifications.- The report shall be . reviewed by the RSC and' any follow-up report shall be submitted to the NRC when authorization is sought to resume operation of the reactor. ; 16.6.4.2 . Action to be taken in the event of an occurrence of type identified in section 16.6.5.2..
- A. Reactor conditions" shall be returned to normal or the reactor
~
shutdown. If ' it is necessary to shut down' the reactor to correct-.the
- occurrence, operations .will not resume until authorized by, the Reactor Supervisor and NRL. Director. +- _
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C. Occurrences shall be reviewed by the VTAR RSC at the next
- regularly scheduled meeting.
16.6.5 Reports
.In addition to the requirements of applicable regulations, reports shall be made to the NRC as follows:
16.6.5.1 Operating Reports Routine annual reports covering the activities of the reactor facility during the previous calendar year shall be submitted to the NRC and RSC within three months following the end of each prescribed year. The prescribed year ends December 31 for the VTAR. Each annual operating report shall include: A. . A narrative summary of reactor operating experience including 4 the energy produced by the reactor and the hours the reactor was crit-j ical. B. .The unscheduled shut-downs including, where applicable, actions taken to preclude reoccurrence. i O C. procedures. A tabu 1ation ef ma3 er chanses in the reactor faci 11ty and-i D. A tabulation of new- tests and experiments 'that are , significantly different from those performed previously -and are not described in the Safety Analysis -Report, including conclusions that no unreviewed safety questions were involved. + E. A tabulation of major preventive and corrective maintenance operations having safety significance, such as scram times and equipment , failures. F. Any facility staff changes. a + G. A review of the Safety Analysis Report including an update of any sections which were affected during the past year. lH. A summary of the nature and amount of radioactive effluent's released or discharged to environs .beyond the effective control of the facility operators as determined at or before .the point of'such release or discharge.- The summary shall include, to the extent practicable, an estimate- of. individual radionuclides - present in the effluent. If the , estimated average release af ter dilution or diffusion is . less than 25
- percent oflthe concentration l allowed or; recommended, a statement to this effect is sufficient.-
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- c. I. A summarized result of environmental surveys performed outside
, /- the facility.
. J. .A aumatry of exposures received by facility personnel and I visitors whera such exposures are greater than 25 percent of that allowed or rectenended. The annual report shall be submitted to the Director, Division of Licensing, USNRC, Washington, DC 20555 and .to the Director, NRC Region 11, Inspection and Enforcement Office, Atlanta, Georgia 30303. 16.6.5.2 Special' Reports l There shall be a report not later than the following working day by _ telephone and confirmed in writing to the NRC, to be followed by a written report that. describes the circumstances of the event within 14 , days of-any of the following: A. Release of radioactivity from the site above allowed limits. B. Violation of safety limits (see 16.6.4.1). C. Any of the following: (1) Operation with actual safety-system settings for required systems less conservative than the Limiting Safety-System Settings specified in these technical specifications. (2) Operation in violation of Limiting Conditions for Operation established in these Technical Specifications unless prompt remedial actial is taken. (3) A reactor safety system component malfunction which renders the reactor safety system incapable of performing its intended safety function unless the malfunction or condition is discovered during maintenance tests or periods of reactor shutdowns.- (Note: Where components or systems -are provided -in addition to those ' required - by technical specifications, the failure of the extra j components or systems is not considered reportable i provided that the minimum number of components or systems , specified or required perform their intended reactor [ safety function.) (4) An unanticipated or uncontrolled change in reactivity j greater . than 0.30% k/k. Reactor trips resulting from a known cause are excluded. l .(5) Abnormal or significant degradation in reactor' fuel and/or-cladding, coolant boundary, or confinement boundary, where O 332= 1 -
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l i I applicable, which could result in exceeding prescribed radiation exposure limits of personal and/or environment. (6) An observed inadequacy in the implementation of I administrative or procedural controls such that the l inadequacy causes or could cause the existence or {. development of an unsafe condition with regard to reactor
- operations.
(7) A violation of these technical' specifications or the } facility license. t 16.6.6 Records 4 Records of the following activities shall be maintained and ] retained for the periods specified below. The records may be in the
- form of logs, data sheets, microfiche or other suitable forms. The j
required information may be contained in singic, or multiple ricords, or j a combination L ereof. Recorder charts showing operating parameters of i the reactor (i.e., power level, temperature, etc.) for unscheduled { shutdown and significant unplanned transients shall be maintained for a } minimum of two years. 4
- 16.6.6.1 Records To Be Maintained For A Period Of At Least Five Years O A. Normal reactor fac .lity operations, including but not limited to the following:
) 1) Completed pre-startup, start up, power change and shut-3 J down checkouts. l 2) Installation or removal of fuel e.lements, control rods, or i J experiments that could affect core reactivity. t i 3) Installation or removal of jumpers, special tags or i notices, or other temporary changes to reactor safety ] circuitry. l- 4) Rod worth measurements and other reactivity measurements. j ^ B. Principle maintenance operations. C. Surveillance activities required by technical specifications D. Experiments performed with the reactor. This requirement may be satisfied by the normal operations log books plus 1) records of radioactive material transferred from the ' facility as f required by the license, and 2) records required by the RSC for performance of new or special experiments or both as described , in 16.6.3 and 16.6.1.5 parts e and f. 333 i
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I !O E. Changes to operating procedures. 16.6.6.2 Records To Be Retained For A: Least One Training Cycle Retraining and requalification of ce.rtified operations personnel. (Records of the most recent complete cycle shall be maintained at all times the individual is employed.) 16.6.6.3 Records To Be Retained For The Life Of The Facility NOTE: Applicable annual reports, if they contain all of the required information,' may be used as records in this section. A. Gaseous and liquid radioactive effluents released to the environs. B. Off-site environmental monitoring surveys required by technical specifications. C. Radiation exposures for all personnel monitored as per 10 CFR 20. D. Up-dated as built drawings of the facility. O S. xechiner7 *ister7 recera ef tr nsient or ever cie et cretes for major components of the facility. F. Records of reviews performed for changes made to procedures or equipment, or reviews of tests and experiments pursuant to 10 CFR 50.59. (See ANS-15.3-74 Appendix) G. Records of meetings of the RSC, RSC Audit Sub-committee and Reactor Operations Staff. H. Special records comprising:
- 1. Final Safety Analysis Report;
- 2. Technical Specifications;
- 3. Environmental Assessment Statements; and,
- 4. Decommissioning Plans (see ANS-15.10 Draf t of 10-79 part 9.0 Reports and documentation).
I. Summary records of training and qualifications of members of facility staff. J. Fuel inventories, receipts and shipments 334
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9 K.- Facility radiation and contamination surveys L. Reportable . occurrences (the special report shall satisfy this
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- 17. QUALITY ASSURANCE At the. time of construction of the VTAR facility, and at other points in time where significant modifications were made to the reactor facility as documented in license amendments and other records of system changes, the necessary assurances of quality in the design, procurement, construction, installation and operation of the f acility were obtained l'
and records kept by those responsible for VTAR safety. Anytime in the
- . future that significant modifications are considered for the facility, j quality assurances are required. Since the VTAR is a small facility, j there is no separate quality assurance division. However, the various
- requirements for quality assurance are and will continue to be met by an effective system of overviews.
The general requirements for establishing and executing a quality assurance program for the design, construction, testing, modification and maintenance of research reactors are included in ANSI N402-1976. ' This standard provides a method acceptable to the NRC for complying with the requirements of 10 CFR 50.34 and will be used as a guide for all future design, construction, and testing connected with significant 4 modifications to the reactor facility. Testing and maintenance of the existing VTAR f acilities continue to have their quality assured as in
~
the past. Maintenance and testing records are kept in accordance with previously established procedures which have been accepted by the NRC.- 10 CFR 50.34 also requires that each applicant for a license to j operate a facility include in the FSAR a description of the admini- , scrative controls to be used to assure safe operation of said facility. This description is contained in Chapter 13 " Conduct of Operations." This administrative organization is considered adequate to assure - the safety of operation from the design stages to testing and operation of a major modification as well as operation of the facility in its present capacity or in any future capacity as approved by the NRC. 4 Chapter 14 also addresses requirements for the establishment of a Startup and Test Committee to review all test data, documentation, etc. relating to major system riodifications to verify that these changes meet l their design requirements following installation, prior to criticality, ' and prior to full power operations. O 336 e , . . - - + --.--%., + . _ .m-g ge ,T ?--MP - d- 4 -P- - -
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l O l REFERENCES Chapter 2 2.1 Algermissen, S. T.,; Perkins, D. M.; Thenhaus, P. C.; Honson, S. L.; Bender, B. L. "Probabilistic Estintes of Maximum Accelera- 1 tion and Velocity in Rock in the Contiguous United States." U.S. ! Department of the Interior, Geological Survey. 1982
- 2. 2 Bollinger, G. A. " Seismic Hazard in Virginia." Virginia !
Minerals.24(4). Charlottesville, Virginia: Division of Mineral Resources; November 1978. 2.3 Commonwealth of Virginia Radiological- Emergency Response Plan - Peacetime Disasters. Richmond, Virginia; June 1983. 2.4 Hewitt, D. A. " Proposed Accelerator Site." Memo to B. Johnson. Blacksburg, Virginia: Virginia Polytechnic Institute and State University, Department of Geological Sciences; 13 August 1981. 2.5 Langley, T. M. ; Marter, W. L. "The Savannah River Plant Site." E. 1. DuPont DeNemours and Company, Aiken, South Carolina: Savannah River Laboratory; September 1973. O 2.6 Mattatese . - > Smit *. G. S. i Simetatien ef eartwe axe Effects on the UCLA Reactor Using Structural Vibrators." Los Angeles, California: Depsrtment of Engineering, University of California at Los Angeles; October 1966. 2.7 Robeson, A.; Hahn, T. M. Jr. " Application for Construction Permit." Blacksburg, Virginia: Virginia Polytechnic Institute and State University, Department of Physics; 5 February 1959. 2.8 Title 10, Chapter 1, Code of Federal Regulations - Energy, Part 100 Section lla. U.S. Nuclear Regulatory Commission; September 1982. 2.9 U.S. Department of Commerce. Climatology of the United States, Number 20-44. Washington,. D.C.: National Oceanic and Atmospheric Administration; October 1971. Chapter 4 4.1 Diaz, N. J.; Vernetson, W. G. Safety Analysis Report. Gainesville, Florida: University of Florida, Department of Nuclear Engineering Sciences; January 1981. 4 a 337
4.2 Florian, R. J. " Thermal Hydraulic Safety Analysis of the VPI&SU Research and Training Reactor." Blacksburg, Virginia: Virginia Polytechnic Institute and State University, Department of Mechanical Engineering; March 1983. 4.3 Record Book UTR-10 VPI. Mountainview, California: Advaced Technology Laboratories; December 1958. 4.4 Safety Evaluation Report. los Angeles, California: University of California at Los Angeles; June 1981. 4.5 Stam, Ephraim. " Performance Characteristics of the VPI Training and Research Reactor." Blacksburg, Virginia: Nuclear Science and Engineering Croup, Virginia Polytechnic Institute and State University; September 1961. Master's Thesis.
- 4. 6 Tuley, K. D. "The Power Excursion Safety Analysis of the VPI&SU Reactor - 500 kW Model." Blacksburg, Virginia: Nuclear Science and Engineering Group, Virginia Polytechnic Institute an'd State University; August 1976. Master's Thesis.
Chapter 5 5.1 Alfa-Laval, Inc. Plate Heat Exchanger, Installation and Maintenance Manual, No. IM70149-E4, Reg 3510, 8005. 1976; avail-
' N able from HVAC Equipment Division, Alfa-Laval, Inc., 2115
, Linwood Avenue, Fort Lee, New Jersey, 07024.
5.2 Allis-Chalmers. Installation, Operation and Maintenance Manual for Type CS0 Model F-4 Pumps. Cincinnati, Ohio. 5.3 Crane Company. Installation, operation and Maintenance Manual (Rev. 11) for Series G, Type GC Pumps. 1978; available from Crane Company, Chempump Division, Warrington Industrial Park, Warrington, Pennsylvania, 18976. 5.4 Mapco Controls Company. Model 9000 Flow Meter Installation and Maintenance Manual, Publication Number ESD-611. December 1981; available from Mapco Controls Company, 11391 E. Tecumseh Street, P.O. Box 21418, Tulsa, Oklahoma, 74121.
- 5. 5 Marley Company. Installation and Maintenance Manual for Model 8805A Cooling Tcwer. 1980; available from The Marley Cooling Tower Company, 5800 Foxridge Drive, Mission, Kansas, 66202.
5.6 Omega Engineering, Inc. Maintenance Manual for Model 49 and 50 Temperature Indicators. 1974; available from Omega Engineering, Inc., One Omega Drive, Box 4047, Stamford, Connecticut, 06907. A U 338 i r
5.7 Omega Engineering, Inc. Temperature Measurement Handbook. 1983; available from Omega Engineering, Inc., One Omega Drive, Box 4047, Stamford, Connecticut, 06907. 5.8 Signet Scientific. Instruction Manual for Mk.577 Digital Flometer. 1980; available from Signet Scientific, 3401 Aerojet Avenue, El Monte, California, 91734. 5.9 Victoreen, Inc. Instruction Manual for G-M Area Monitoring Systems, Model 855 Series. 1979; Part No. 855-10-1, available from Victoreen, Inc., 10101 Woodland Avenue, Cleveland, Ohio, 44104. Chapter 6 6.1 Florian, R. J. " Thermal Hydraulic Safety Analysis of the VPI&SU Research and Training Reactor." Blacksburg, Virginia: Virginia Polytechnic Institute and State University, Department of Mechanical Engineering; March 1983. 6.2 Harrison Radiator Division, General Motors Corporation. Catalog HE-500. Lockport, New York: General Motors Corporation. (ECCS Heat Exchanger).
- 6. 3 Lamarsh, J. R. Introduction to Nuclear Engineering. 2nd ed.
O Reading, Massachusetts: 1983. Addison-Wesley Publishing Company; Chapter 7 7.1 Bailey Meter Company. Nuclear Instrumentation System Maintenance and Repair Manual. Wickliffe, Ohio: Bailey Meter Company; 1961.
- 7. 2 Keithley Instruments, Inc. Instruction Manual for Models 416,417 High Speed Picoammeters. 1976; available .from Keithley Instruments, Inc., 28775 Aurora Road, Cleveland, Ohio, 44139.
7.3 Mapco Controls Company. Model 9000 Flow Meter Installation and Maintenance Manual, Publication Number ESD-611. December 1981; available from Mapco Controls Company, 11391 E. Tecumseh Street, P.O. Bcx 21418, Tulsa, Oklahoma, 74121. 7.4 Signet .S cientific. Instruction Manual for Mk.577 Digital Flometer. 1980; available from Signet Scientific, 3401 Aerojet Avenue, El Monte, California, 91734. ; i' 7. 5 Victoreen, Inc. Instruction Manual for G-M Area Monitoring Systems, Model 855 Series. 1979; Part No. 855-10-1, available from Victoreen, Inc., 10101 Woodland Avenue, Cleveland, Ohio, l , 44104. 1 . I t ! 339 l
O Chapter 8 ) l 8.1 Advance Conversion Devices Company. Installation, Maintenance and Repair Manual for Model U74-2 Uninterruptible Power Supply. 1982; available from Advance Conversion Devices Company, 109 , Eighth Street, Passaic, New Jersey, 07055. 8.2 National Fire Protection Association. National Electric Code 1981. ANSI /NFPA 70. Chapter 9 9.1 American Machine and Foundry Company. Maintenance and Repair Manual for Model SD-7-4066 Master / Slave Manipulator ARM. York, Pennsylvania: AMF Atomics Division; 1964. 9.2 Robins and Meyers. Operating and Maintenance Manual for Model FSDiB Electric Hoist. Springfield, Ohio: Robins and Meyers. Oserhead Crane. Chapter 11 11.1 ANSI /ANS(1977). Design Objective for the Monitoring of Systems
- Controlling Research Reactor Effluents, Report 15.12-1977. New O ver*= imeric tie 1 St a ra 1n tic tei 1977-11.2 Cember, Herman. Introduction to Health Physics. 2nd ed. New York
- Pergamon Press, Inc. ; 1983.
, 11.3 National Oceanic and Atmospheric Administration. Airport Climatological Summary. Roanoke, Virginia: Woodrum Airport, NOAA; 1982. Chapter 12 12.1 Curtner, A.; Parkinson, T. F. Application for a Class 104 License for the Virginia- Polytechnic Institute and State University Nuclear Research Reactor Facility. Blacksburg, Virginia: Nuclear Science and Engineering Group; 1980. 12.2 Onega, R. J.; Curtner,--.A. Reactor Operator Training Manual, Nuclear Research Reactor Facility, VPI&SU. Blacksburg, Virginia: Nuclear Science and Engineering Group; 1978. 12.3 Safety and Health Programs. Radiation Safety Manual. Blacksburg, Virginia: Department of Safety and Health Programs, Virginial Polytechnic Institute and State University; 1980. O 340
0 12.4 U.S. Nu. lear Regulatory Commission. Rules and Regulations (Title 10, Chapter 1, Code of Federal Regulations, Part 20). 1980.
. Chapter 13 13.1 ANSI /ANS 1974. Records and Reports for Research Reactors.
15.3. Arcerican National Standards Institute; 1974 { 13.2 ANSI /ANS 1977. Selection and Training of Personnel for Research Reactors. 15.4 American National Standards Institute; 1977. 13.3 Diaz, N. J.: Vernetson, W. G. Safety Analysis Report. Gainesville, Florida: University of Florida, Department of Nuclear Engineering Sciences; January 1981. 13.4 Title 10, Chapter ' 1, Code of Federal Regulations - Energy, Part 55, Appendix A. U. S . Nuclear Regulatory Commission; September 1982.- 13.5 VTAR Facility Staff. Emergency Plan for the VTAR. Blacksburg, Virginia: Virginia Polytechnic Institute and State University Nuclear Reactor Laboratory; May 1984. j Chapter 15 O 15.1 imerican Natienat Standards 1nstitete. inst / INS 15.15-1978
" Criteria for the Reactor. Safety Systems of Research Reactors."
LaGrange, Illinois: American Nuclear Society; 1978. 15.2 American Nuclear Society. ANS 15.7-1977 (N379), " Guide for Research Reactor Site Evaluation." LaGrange, Illinois: American Nuclear Society; 1977. 15.3 Advanced Technology Laboratories. Hazards Analysis, UTR-10 Standard Model, ATL-137. Mountain View, California: Advanced Technology Laboratories; October 1959. 15.4 Advanced Technology' Laboratories. Record Book, UTR-10, VPI. Mountain View, California: Advanced Technology Laboratories; December 1958. 15.5 Cort, C. E. Fuel Temperatures in an Argonaut Reactor Core Following a Hypothetical Design Basis Accident. Ios Alamos, New Mexico:. Los Alamos National Laboratory; June 1981. NUREG/CR-2198. 15.6 Diaz, N.. J.: Vernetson, W. G. Safety Analysis Report. Gainesville, Florida: University of Florida, Department -of 4 Nuclear Engineering Sciences; January 1981. O: , 1 341
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