LEU PSAR 6 DEC 2017 TABLE OF CONTENTS TABLE OF CONTENTS .................................................................................................... a ••••••• i 1.0 MIT Research Reactor·······························································"······························· 1-1 1.1 Introduction .......................................................................................................... 1-1 1.2 Summary and Conclusions of Principal Safety Considerations .............................. 1-1 1.2.1 Consequences from Operation and Use ............................................................. 1-1 1.2.2 Safety Considerations on Choice of Site, Fue~ and Power Level.. ..................... 1-2 1.2.3 Inherent Safety Features ................................................................................... 1-3 1.2.4 Design Features for Safe Operation and Shutdown............................................ 1-4 1.2.5 Potential Accidents ........................................................................................... 1-5 1.3 General Description of the Facility ........................................................................ 1-6 1.4 Shared Facilities and Equipment.. ....................................................................... 1-10 1.5 Comparison with Similar Facilities ..................................................................... 1-11 1.6 Summary of Operation ......................................................................................... 1-11 1.7 Nuclear Waste Policy Act of1982 ...................................................................... 1-11 1.8 Facility Modifications and History ...................................................................... 1-12 References .................................................................................................................. 1-13 2.0 Site Characteristics .... :............................................................................................... 2-1 3.0 Design of Structures, Systems, and Components ...................................................... 3-l 3.1 Design Criteria! ..................................................................... ·............................... 3-1 3.1.1 Prevention of Release of Radioactive Material.. ................................................ 3-1 3.5 Systems and Components ..................................................................................... 3-l 4.0 Reactor Description .................................................................................................... 4-1 4.1 Summary Description ........................................................................................... 4-1 4.2 Reactor Core ................................... : ..................................................................... 4-3 4.2.1 Reactor Fuel ..................................................................................................... 4-4 4.2.5 Core Support Structure ..................................................................................... 4-7 4.3 Reactor Tanks ........................................................................................... ;........... 4-7 4.3.1 Light-Water Core Tank ..................................................................................... 4-7 4.5 Nuclear Design ..................................................................................................... 4-8 Low Enriched Uranium (LEU) Conversion Preliminary Safety Analysis Report for the MIT Research Reactor (MITR) 1 LEU PSAR 6 DEC 2017 4.5.1 Normal Operating Conditions .......................................................................... .4-9 4.5.2 Reactor Core Physics Parameters ................................................................... .4-18 4.5.3 Operating Limits ............................................................................................. 4-19 4.6 Thermal-hydraulic Design .................................................................................. 4-22 4.6.1 Design Basis ·························································································:·········4-22 4.6.2 Major Correlations Used in the Thermal-Hydraulic Limit Calculations .......... .4-25 4.6.3 Reactor Power Deposition and Core Flow Distribution .................................. .4-30 4.6.4 Engineering Hot Channel Factors and Uncertainty Treatments ....................... .4-32 4.6.5 Thermal-Hydraulic Limits .............................................................................. 4-35 4.6.6 Derivation of the Safety Limits ...................................................................... .4-36 4.6.7 Calculation of the Limiting Safety System Settings ........................................ .4-41 4.6.8 Refueling Considerations ................................................................................ 4-44 References .................................................................................................................. 4-46 5.0 Reactor Coolant Systems ........................................................................................... 5-l 5.2 Primary Coolant System ....................................................................................... 5-1 5.2.1 Main Flow System .......................................................................................... .-.5-1 6.0 Engineered Safety Features ....................................................................................... 6-1 6.4 Emergency Core Cooling System .......................................................................... 6-1 References .................................................................................................................... 6-4 7.0 Instrumentation and Control Systems ....................................................................... 7-1 7.2 Design oflnstrumentation and Control System ..................................................... 7-1 7.2.2 Design Basis Requirements ............................................................................... 7-1 7.4 Reactor Protection System .................................................................................... 7-3 7.4.1 Nuclear Safety System ...................................................................................... 7-3 7.4.2 Non-Nuclear Safety System .............................................................................. 7-5 7.6 Control Console Display Instruments .................................................................... 7-7 7.6.2 Channel 8 ......................................................................................................... 7-7 8.0 Electrical Power Systems ........................................................................................... 8-1 9.0 Auxiliary Systems ....................................................................................................... 9-1 Low Enriched Uranium (LEU) Conversion Preliminary Safety Analysis Report for the MIT Research Reactor (MITR) · 11 LEU PSAR 6 DEC 2017 10.0 Experimental Facilities and Utilization ................................................................... 10-1 11.0 Radiation Protection and Waste Management ....................................................... 11-1 11.2 Radioactive Waste Management ......................................................................... 11-1 11.2.3 Release of Radioactive Waste ......................................................................... 11-1 12.0 Conduct of Operations ............................................................................................. 12-1 13.0 Accident Analysis ............................................................._ ........................................ 13-1 13.1 Accident-Initialing Events and Scenarios ............................................................ 13-1 13.1.1 Maximum Hypothetical Accident ................................................................... 13-1 13.1.2 Insertion of Excess Reactivity ......................................................................... 13-2 13.1.3 Loss of Primary Coolant .................. :.............................................................. 13-2 13.1.4 Loss of Primary Coolant Flow ........................................................................ 13-2 13.1.5 Mishandling or Malfunction ofFuel.. .............................................................. 13-3 13 .1. 6 Experiment Malfunction ................................................................................. 13-3 13.1.7 External Events ............................................................................................... 13-3 13.1.8 Mishandling or Malfunction ofEquipment ..................................................... 13-3 13.2 Accident Analysis and Determination of Consequences ...................................... 13-3 13.2.1 Maximum Hypothetical Accident ................................................................... 13-4 13.2.2 Insertion of Excess Reactivity ....................................................................... 13-12 13.2.3 Loss of Primary Coolant ............................................................................... 13-14 13.2.4 Loss of Primary Coolant Flow ...................................................................... 13-16 13.2.5 Mishandling or Malfunction ofFuel. ............................................................. 13-18 13.2.6 Experiment Malfunction ..............................................................................