ML12156A196

University of Texas at Austin - Redacted Supplement to Application Containing Table of Contents, Chapter 4 and Chapter 9 of the Safety Analysis Report (TAC No. ME7694)

ML12156A196
Person / Time
Site:	University of Texas at Austin
Issue date:	01/17/2012
From:	Biegalski S University of Texas at Austin
To:	Lising A J Document Control Desk, Division of Policy and Rulemaking
Lising A J
References
TAC ME7694
Download: ML12156A196 (97)

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UNIVERSITY OF TEXAS AT AUSTIN RESEARCH REACTOR LICENSE NO. R-129 DOCKET NO. 50-602 UNIVERSITY OF TEXAS AT AUSTIN LICENSE RENEWAL APPLICATION JANUARY 17, 2012 REDACTED VERSION* SECURITY-RELATED INFORMATION REMOVED *REDACTED TEXT AND FIGURES BLACKED OUT OR DENOTED BY BRACKETS DepanmL aot o i""hanical EngineeringTHE UNIVERSITY OF TEKAS AT AUSTINNuclearf .'5i*eerinig waaing taboratory AtArin, "7Txas 787585.I 2-232-5370 -FAX 512-471- -589- htp,'/wu'A me. nrexas.edul/.- nel/ATTN: Document Control Desk,U.S. Nuclear Regulatory Commission,Washington, DC 20555-0001Allan Jason LisingProject ManagerDivision of Policy and RulemakingResearch and Test Reactors Licensing BranchJanuary 17, 2012SUBJECT: Docket No. 50-602, Information Supplementing Request for Renewal of Facility OperatingLicense R-129 (TAC ME 7694)REFERENCE: (1) ML110040316(2) Letter, 12/12/2011 Docket No. 50-602, Request for Renewal of Facility OperatingLicense R-129Sir:In accordance with USNRC direction (ADAMS ML110040316), a request for renewal of Facility OperatingLicense R-129 (Docket 50-602) was submitted on 12/12/2011. The attached material provides minoreditorial corrections and clarification of three chapters previously submitted of the Safety AnalysisReport and the Technical Specifications. An additional item is included to support review of theproposed Technical Specifications. In summary:* Table of Contents (reflecting updates)* Chapter 4, additional figures are provided to better describe the control rod drive mechanisms,and the section on thermal hydraulic analysis was substantially augmented.* Chapter 9, operation of the auxiliary purge system and the confinement isolation was revised.* Chapter 12, the responsibilities of the Senior Reactor Operator was rewritten to emphasize therole of the Supervisor in reactor operations.* Technical Specifications, editorial changes and various improvements were made.* Technical Specifications review material: a tabulation of the current Technical Specifications wasprepared with a comparison to the proposed, new Technical Specifications.Your attention in this matter is greatly appreciated,I declare under penalty of perjury that the foregoing is true and correct.

0ITHE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 0*:. l 01/2012SAFETY ANALYSIS REPORT 000Table of ContentsSection Page1. THE FACILITY 1-11.1 Introduction 1-11.2 Summary and conclusions on principle safety considerations 1-21.3 General description of the facility 1-3A. Site 1-3B. Building 1-3C. Reactor 1-3C.1 Reactor Core. 1-5C.2 Reactor Reflector. 1-5D. Reactor Control. 1-6E. Experiment Facilities. 1-6E.1 Upper Grid Plate 7L and 3L Facilities 1-6E.2 Central Thimble 1-6E.3 Rotary Specimen Rack (RSR) 1-6E.4 Pneumatic Tubes 1-7E.5 Beam Port Facilities 1-7E.5 (1) Beam Port 1 (BP1) 1-7E.5 (2) Beam Port 2 (BP2) 1-8E.5 (3) Beam Port 3 (BP3) 1-9E.5 (4) Beam Port 4 (BP4) 1-10E.5 (5) Beam Port 5 (BPS) 1-10F Other Experiment and Research Facilities 1-101.4 Overview of shared facilities and equipment 1-101.4.3 Reference the other facilities operating history, safety and reliability 1-101.5 Summary of operations 1-121.6 Compliance with NWPA of 1982 1-121.7 Facility history & modifications 1-132.0 SITE DESCRIPTION 2-12.1 GENERAL LOCATION AND AREA 2-12.2 POPULATION AND EMPLOYMENT 2-72.3 CLIMATOLOGY 2-112.4 GEOLOGY 2-142.5 SEISMOLOGY 2-222.6 HYDROLOGY 2-222.7 HISTORICAL 2-273.0 DESIGN OF SYSTEMS, STRUCTURES AND COMPONENTS 3-13.1 Design Criteria for Structures, Systems and Components for Safe Reactor Operation 3-23.1.1 Fuel Moderator Elements 3-2 SAFETY ANALYSIS REPORT, TABLE OF CONTENTS 01/2012Table of ContentsSection Page3.1.2 Control Rods 3-33.1.3 Core and structural Support 3-43.1.4 Pool and Pool Support Systems 3-43.1.5 Biological Shielding 3-43.1.6 NETL Building/Reactor Bay 3-5A. Building 3-6B. Reactor Bay 3-73.1.7 Ventilation Systems 3-73.1.8 Instruments and Controls 3-83.1.9 Sumps and Drains 3-83.2 Meteorological Damage 3-93.3 Water Damage 3-93.4 Seismic Damage 3-10A. Core and structural Support 3-10B. Pool and pool cooling 3-10C. Building 3-104.0 Reactor 4-14.1 Summary description 4-14.2 Reactor Core 4-14.2.1 Reactor Fuel 4-2A. Fuel matrix 4-2(1) Fabrication 4-3(2) Physical Properties 4-4(3) Operational Properties 4-7(4) Neutronic Properties 4-7(5) Fuel Morphology & Outgassing 4-8(6) Zr water reaction 4-9(7) Mechanical Effects 4-9(8) Fission Product Release 4-10B. Cladding 4-104.2.2 Control Rods and Drive Mechanisms 4-11A. Control Rods 4-13B. Standard Control Rod Drives 4-16C. Transient Control Rod Drive 4-17D. Control Functions 4-19E. Evaluation of the Control Rod System 4-204.2.3 Neutron Moderator and Reflector (Core Structure) 4-20A. Upper grid plate 4-20B. Reflector 4-23ii THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR *Ilk, 01/2012SAFETY ANALYSIS REPORT I 1/01Table of ContentsSection Page(1) Radial Reflector 4-23(2) Graphite Rods. 4-24(3) Axial Reflector 4-24C. Lower grid plate 4-244.2.4 Neutron Startup Source 4-264.2.5 Core support structure 4-26A. Core Support Platform 4-26B. Safety plate 4-274.3 Reactor Pool 4-284.4 Biological Shield 4-304.5 Nuclear Design 4-324.5.1 Normal Operating Conditions 4-324.5.2 Nominal Reactivity Worth Values 4-334.5.3 Reactor Core Physics 4-32A. Reference Calculations 4-34B. Prompt Negative Temperature Coefficient 4-354.5.4 Operating Limits 4-39A. Core Peaking Factors 4-39B. Power distribution within a Fuel Element. 4-40C. Power per rod 4-414.6 Core Reactivity 4-454.7 Thermal Hydraulic Design 4-474.7.1 Heat Transfer Model 4-484.7.2 Results 4-51Appendix 4.1, PULSING THERMAL RESPONSE 4.1-15.0 REACTOR COOLANT SYSTEMS 5-15.1 Summary Description 5-15.2 Reactor Pool 5-15.2.1 Heat Load 5-25.2.2 Pool Fabrication 5-35.2.3 Beam Ports 5-35.3 Pool Cooling System 5-45.3.1 Reactor Pool 5-45.3.2 Pool Heat Exchanger 5-55.3.3 Secondary Cooling 5-105.3.4 Control System 5-105.4 Primary Cleanup System 5-115.5 Makeup Water System 5-125.6 Cooling System Instruments and Controls 5-13iii SAFETY ANALYSIS REPORT, TABLE OF CONTENTS 01/2012Table of ContentsSection Page6.0 ENGINEERED SAFEGUARD FEATURES 6-16.1 References 6-17.0 INSTRUMENTATION AND CONTROL SYSTEM 7-17.1 DESIGN BASES 7-17.1.1. NM-1000 Neutron Channel 7-37.1.2. NP-1000 Power Safety Channel 7-57.1.3. Reactor Control Console 7-67.1.4. Reactor Operating Modes 7-77.1.5. Reactor Scram and Shutdown System 7-117.1.6. Logic Functions 7-127.1.7 Mechanical Hardware 7-137.2 DESIGN EVALUATION 7-148.0 ELECTRIC POWER SYSTEMS 8-19.0 AUXILIARY SYSTEMS 9-19.1 Confinement System ...... 9-19.2 HVAC (Normal Operations) 9-19.2.1 Design basis 9-29.2.2 System description 9-39.2.3 Operational analysis and safety function 9-49.2.4 Instruments and Controls 9-69.2.5 Technical Specifications, bases, testing and surveillances 9-89.3 Auxiliary Purge System 9-89.3.1 Design basis 9-89.3.2 System description 9-89.3.3 Operational Analysis and Safety Function 9-99.3.4 Instruments and controls 9-99.3.5 Technical Specifications, bases, testing and surveillances 9-109.4 Fuel storage and handling 9-109.4.1 Design basis 9-109.4.2 System description 9-109.4.3 Operational analysis and safety function 9-129.4.4 Instruments and controls 9-129.4.5 Technical Specifications, bases, testing and surveillances 9-129.5 Fire protection systems 9-139.5.1 Design basis 9-139.5.2 System description 9-139.5.3 Operational analysis and safety function 9-14iv THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR o I 01/2012SAFETY ANALYSIS REPORT 0401/ 2012Table of ContentsSection Page9.5.4 Instruments and controls 9-159.5.5 Technical Specifications, bases, testing and surveillances 9-159.5 Communications systems 9-159.5.1 Design basis 9-159.5.2 System description 9-159.5.4 Instruments and controls 9-169.5.5 Technical Specifications, bases, testing and surveillances 9-169.6 Control, storage, use of byproduct material (including labs) 9-169.6.1 Design basis 9-169.6.2 System description (drawings, tables) 9-169.6.3 Operational analysis and safety function 9-179.6.4 Instruments and controls 9-179.6.5 Technical Specifications, bases, testing and surveillances 9-179.7 Control and storage of reusable components 9-179.7.1 Design basis 9-179.7.2 System description 9-179.7.3 Operational analysis and safety function 9-179.7.4 Instruments and controls 9-179.7.5 Technical Specifications, bases, testing and surveillances 9-189.8 Compressed gas systems 9-189.8.1 Design basis 9-189.8.2 System description 9-189.8.3 Operational analysis and safety function 9-189.8.4 Instruments and controls 9-199.8.5 Technical Specifications, bases, testing and surveillances 9-1910.0 EXPERIMENTAL FACILTIES AND UTILIZATION 10-110.1 Summary Description 10-110.2 In-Core Facilities10-310.2.1 Central Thimb;e (In-Core Facility) 10-4A. DESCRIPTION 10-4B. DESIGN & SPECIFICATIONS 10-5C. REACTIVITY 10-6D. RADIOLOGICAL ASSESSMENT 10-6E. INSTRUMENTATION 10-7F. PHYSICAL RESTRAINTS, SHIELDS, OR BEAM CATCHERS 10-7G. OPERATING CHARACTERISTICS 10-7H. SAFETY ASSESSMENT 10-810.2.2 Fuel Element Positions (In-Core Facilities) 10-8V SAFETY ANALYSIS REPORT, TABLE OF CONTENTS 01/2012Table of ContentsSection Page10.2.2.1 Pneumatic Sample Transit System 10-8A. DESCRIPTION. 10-8B. DESIGN & SPECIFICATIONS. 10-9C. REACTIVITY 10-10D. RADIOLOGICAL ASSEMENT 10-11E. INSTRUMENTATION 10-11F. PHYSICAL RETRAINTS, SHIELDS, OR BEAM CATCHERS10-12G. OPERATING CHARACTERISTICS10-12H. SAFETY ASSESSMENT 10-1210.2.2.2 Three Element Irradiator 10-13A. DESCRIPTION.10-13B. DESIGN & SPECIFICATIONS.10-13B (1) Upper and Lower Grid Plate Modifications.10-13B (2) Alignment Frame.10-14B (3) Three Element Facility Canister.10-14C. REACTIVITY 10-16C (1) Reactivity Calculation 10-17C (2) Reactivity Measurements10-18D. RADIOLOGICAL ASSESSMENT 10-18E. INSTRUMENTATION 10-19F. PHYSICAL RESTRAINTS, SHIELDS, or BEAM CATCHERS10-19G. OPERATING CHARACTERISTICS10-19H. SAFETY ASSESSMENT 10-19H (1) Cooling 10-19H (2) Temperature 10-20H (3) Pressure 10-21H (4) LOCA potential 10-2210.2.2.3 6/7 Element Irradiator 10-22A. DESCRIPTION 10-22B. DESIGN AND SPESIFICATIONS,10-22C. REACTIVITY.10-23D. RADIOLOGICAL ASSESSMENT 10-23E. INSTRUMENTATION 10-23F. PHYSICAL RESTRAINTS, SHIELDS OR BEAM CATCHERS10-24G. OPERATING CHARACTERISTICS10-24H. SAFETY ASSESSMENT 10-24H (1) Temperature (Fuel)10-24H (2) Temperature (Lead)10-24H (3) Pressure (irradiation Can)10-24H (4) Pressure (Lead Sleeve) 10-25vi THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 0#-0, 01/2012SAFETY ANALYSIS REPORTTable of ContentsSection PageH (5) Mass10-25H (6) Structural 10-2510.2.3 Rotary Specimen Rack 10-26A. DESCPIPTION 10-26B. DESIGN SPEC!,FiC1,1A*ATiN5 10-26C. REACTIVITY 10-28D. RADIOLOGICAL ASSESSMENT 10-28E. INSTRUMENTATION 10-29F. PHYSICAL RESTRAINTS, SHIELDS OR BEAM CATCHERS10-29G. OPERATING CHARACTERISTICS10-29H. SAFETY ASSESMENT 10-2910.3 Beam Ports10-29A. DESCRIPTION 10-29B. DESIGN AND SPECIFICATIONS10-30C. REACTIVITY 10-31D. RADIOLOGICAL ASSESSMENT 10-31E. INSTRUMENTATION 10-31F. PHYSICAL RESTRAINTS, SHIELDS, OR BEAM CATCHERS10-31G. OPERATING CHARACTERISTICS10-33H. SAFETY ASSESSMENT 10-3310.4 Cold Neutron Source 10-34A. DESCRIPTION 10-34B. DESIGN AND SPECIFICATIONS10-34C. REACTIVITY 10-37D. RADIOLOGICAL 10-37E. INSTRUMENTATION 10-37F. PHYSICAL RESTRAINTS, SHIELDS, OR BEAM CATCHERS10-39G. OPERATING CHARACTERISTICS10-39H. SAFETY ANALYSIS 10-4010.5 Non-reactor experiment facilities 10-4110.5.1 Neutron generator room 10-4110.5.2 Subcritical assembly 10-4210.5.3 Laboratories 10-4210.5.3.1 Radiochemistry laboratory 10-4210.5.3.2 Neuron Activation Analysis Laboratory 10-4310.5.3.3 Radiation detection laboratory 10-4310.5.3.4 Sample preparation laboratory 10-4310.5.3.5 General purpose laboratory 10-4310.6 Experiment Review 10-43vii SAFETY ANALYSIS REPORT, TABLE OF CONTENTS 01/2012Table of ContentsSection Page11 Radiation Protection and Waste Management 11-111.1 Radiation Protection 11-111.1.1 Radiation Sources11-111.1.1.1 Airborne Radiation Sources11-111.1.1.1.1 Production of Ar-41 in the Reactor Room 11-111.1.1.1.2 Radiological Impact of Ar-41 Outside the 11-2Operations Boundary11.1.1.2 Liquid Radioactive Sources11-311.1.1.2.1 Radioactivity in the Primary Coolant 11-311.1.1.2.2 N-16 Radiation Dose Rates from Primary 11-4Coolant11.1.1.3 Solid Radioactive Sources11-411.1.1.3.1 Shielding Logic 11-611.1.2 Radiation Protection Program 11-611.1.2.1 Management and Administration 11-711.1.2.1.1 Level 1 Personnel 11-711.1.2.1.2 Level 2 Personnel 11-711.1.2.1.3 Level 3 Personnel 11-911.1.2.1.4 Level 4 Personnel 11-1011.1.2.1.5 Other Facility Staff 11-1111.1.2.2 Health Physic Procedures and Document Control 11-1111.1.2.3 Radiation Protection Training 11-1111.1.2.4 Audits of the Radiation Protection Program 11-1311.1.2.5 Health Physics Records and Record Keeping 11-1311.1.3 ALARA Program 11-1311.1.4 Radiation Monitoring and Surveying 11-1411.1.4.1 Monitoring for Radiation Levels and 11-14Contamination11.1.4.2 Radiation Monitoring Equipment 11-1511.1.4.3 Instrument Calibration 11-1511.1.5 Radiation Exposure Control and Dosimetry 11-1611.1.5.1 Shielding 11-1611.1.5.2 Containment 11-1611.1.5.3 Entry Control 11-1711.1.5.4 Personal Protective Equipment 11-1711.1.5.5 Representative Annual Radiation Doses 11-1711.1.5.6 Personnel Dosimetry Devices 11-1811.1.6 Contamination Control 11-1811.1.7 Environmental Monitoring 11-1911.2 Radioactive Waste Management 11-1911.2.1 Radioactive Waste Management Program 11-20viii THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 01/2012SAFETY ANALYSIS REPORT Ru01/201tTable of ContentsSection Page11.2.2 Radioactive Waste Controls 11-2011.2.2.1 Gaseous Waste 11-2011.2.2.2 Liquid Waste 11-21112.2.3 Solid Waste 11-2111 2.2.4 Mixed Waste 11-2111.2.2.5 Decommissioning Waste 11-2111.2.3 Release of Radioactive Waste 11-2212 Conduct of Operations12-112.1 Organization 12-112.1.1 Structure 12-112.1.1.1 University Administration 12-112.1.1.2 NETL Facility Administration 12-112.1.2 Responsibility 12-312.1.2.1 Executive Vice President and Provost 12-312.1.2.2 Vice President for University Operation 12-312.1.2.3 Associate Vice President of Campus Safety And 12-3Security12.1.2.4 Director of Nuc!ear Engineering Teaching 12-3Laboratory12.1,2.5 Associate Director of Nuclear Engineering 12-3Teaching Labor3tory12.1.2.6 Reactor Oversight Committee 12-412.1.2.7 Radiation Safety Officer 12-412.1.2.8 Radiation Safety Committee 12-412.1.2.9 Reactor Supervisor 12-412.1.2.10 Health Physicist 12-612.1.2.11 Laboratcry Manager.12-612.1.2.12 Reactor Operators12-612.1.2.13 Technical Support 12-612.1.2.14 Radiological Controls Technicians12-612.1.2.15 Laboratory Assistants12-712.1.3 Staffing 12-712.1.4 Selection and Training of Personnel 12-812.1.4.1 Qualifications12-812.1.4.2 Job Descriptions12-812.1.4.2.1 Facility Director 12-812.1.4.2.2 Associate Director 12-912.1.4.2.3 Reactor Supervisor 12-912.1.4.2.4 Health Physicist 12-9ix SAFETY ANALYSIS REPORT, TABLE OF CONTENTS 01/2012Table of ContentsSection Page12.1.4.2.5 Laboratory Manager 12-912.1.4.2.6 Reactor Operators12-912.1.4.2.7 Technical Support 12-912.1.4.2.8 Radiological Controls Technician 12-1012.1.4.2.9 Laboratory Assistants 12-1012.1.5 Radiation Safety 12-1012.2 Review and Audit Activities 12-1012.2.1 Composition and Qualifications 12-1012.2.2 Charter and Rules 12-1112.2.3 Review Function 12-1112.2.4 Audit Function 12-1212.3 Procedures 12-1212.4 Required Actions 12-1312.4.1 Safety Limit Violation 12-1312.4.2 Release of Radioactivity 12-1412.4.3 Other Reportable Occurrences 12-1412.5 Reports 12-1412.5.1 Operating Reports 12-1512.5.2 Other or Special Reports 12-1512.6 Records 12-1612.6.1 Lifetime Records 12-1612.6.2 Five Year Period 12-1612.6.3 One Training Cycle 12-1712.7 Emergency Planning 12-1712.8 Security Planning 12-1712.9 Quality Assurance 12-1712.10 Operator Requalification 12-1812.11 Startup Program 12-1912.12 Environmental Report 12-1913.0 ACCIDENT ANALYSIS13-113.1 Notation and Fuel Properties13-113.2 Accident Initiating Events and Scenarios13-213.3 Maximum Hypothetical Accidents, Single Element Failure in Air 13-513.3.1 Assumptions13-513.3.2 Analysis 13-6A. Radionuclide Inventory Buildup and Decay, Theory 13-7B. Fission Product Inventory Calculations 13-7C. Fission Product release 13-10D. ALl Consequence Analysis13-11x THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR :, -01/2012SAFETY ANALYSIS REPORTI0/22Table of ContentsSection PageE. DAC Consequence Analysis13-14F. Effluent release Ccnsequence Analysis13-17F (1) Atmospheric Dispersion 13-18F (2) CASE I 13-19F (3) CASE II 13-20F (3) Source Term Release Rate 13-2213.3.3 Results and Conclusions 13-2413.4 Insertion of Excess Reactivity 13-2513.4.1 Initial Conditions, Assumptions, and Approximations 13-2513.4.2 Computational Model for Power Excursions 13-2613.4.3 Results and Conclusions 13-3013.5 Loss of Reactor Coolant Accident 13-3213.5.1 Initial Conditions, Assumptions, and Approximations 13-3413.5.2 Heat Transfer to Air 13-34A. Buoyancy Forces13-35B. Friction Losses13-35C. Losses from Flow Restrictions 13-3513.5.7 Radiation Levels from the Unccvered Core 13-3913.5.8 Results and Conclusions 13-4313.6 Loss of Coolant Flow 13-4313.6.1 Initialing Events and Scenarios 13-4313.6.2 Accident Analysis and Determination of Consequences 13-4313.7 Mishandling or Malfunction of Fuel 13-4413.7.1 Initiating Events and Scenarios 13-4413.7.2 Analysis 13-4413.8 Experiment Malfunction 13-4413.8.1 Accident Initiating Events and Scenarios 13-4413.8.2. Analysis and Determination of Consequences13-45A. Administrative Controls13-45B. Reactivity Considerations13-45C. Fueled Experiment Fission Product Inventory 13-46D. Explosives 13-4713.9 Loss of Normal Electric Power 13-4913.9.1 Initiating Events and Scenarios 13-4913.9.2 Accident Analysis and Determination of Consequences 13-4913.10 External Events 13-4913.10.1 Accident Initiating Events and Scenarios 13-4913.10.2 Accident Analysis and Determination of Consequences 13-5013.11 Experiment Mishandling or Malfunction 13-5013.11.1 Initiating Events and Scenarios 13-50xi SAFETY ANALYSIS REPORT, TABLE OF CONTENTS 01/2012Table of ContentsSection Page13.11.2 Accident Analysis and Determination of Consequences 13-50Appendix 13.1, T-6 DEPLETION ANALYSIS INPUT FILE FOR SCALE CALCULATION 13.1-1Appendix 13.2, ORIGEN ARP INPUT 13.2-1Appendix 13.3, MCNP INPUT FOR LOCA DOSES 13.3-115.0 FINANCIAL QUALIFICATIONS15-115.1 Financial Ability to Operate a Nuclear Research Reactor 15-115.2 Financial Ability to Decommission the Facility 15-115.3 Bibliography 15-1Appendix 15.1, STATUTES AND EXCERPTS REGARDING UT 15.1-1Appendix 15.2, FIVE-YEAR OPERATING COST ESTIMATE 15.2-1Appendix 15.3, Letter of Intent, Ultimate Decommissioning 15.3-1Appendix 15.4, DECOMMISSIONING COST ESTIMATE 15.4-1APPENDIX 15.5, FUELS ASSISTANCE CONTRACT 15.5-1xii THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 0*" 01/2012SAFETY ANALYSIS REPORT ?** REMLIST OF FIGURES PageFigure 1.1, UT TRIGA Mark II Nuclear Research Reactor 1-4Figure 1.2, Core and Support Structure Details 1-5Figure 1.3, Beam Ports 1-8Figure 1.4A, Days of Operation per Year 1-12Figure 1.4B, Burnup per Year 1-12Figure 2.1, STATE OF TEXAS COUNTIES 2-2Figure 2.2, TRAVIS COUNTY 2-3Figure 2.3, CITY OF AUSTIN 2-4Figure 2.4, JJ PICKLE RESEARCH CAMPUS 2-5Figure 2.5, LAND USAGE AROUND JJ PICKLE RESEARCH CAMPUS, 2007 2-6Figure 2.6, 2009 ZIP CODE BOUNDARIES 2-10Figure 2.7, AUSTIN CLIMATOLOGY DATA 2-11Figure 2.8, AUSTIN WIND ROSE DATA 2-12Figure 2.9, TROPICAL STORM PATHS WITHIN 50 NAUTICAL MILES OF AUSTIN, TEXAS (ALL 2-21RECORDED HURRICANES RATED H1 AND UP)Figure 2.10, TROPICAL STORM PATHS WITHIN 50 NAUTICAL MILES OF AUSTIN, TEXAS (ALL 2-21RECORDED STORMS RATED TROP OR SUBTROP)Figure 2.11, BALCONES FAULT ZONE 2-23Figure 2.12, TEXAS EARTHQUAKE DATA 2-24Figure 2.13, TEXAS EARTHQUAKE DATA 2-25Figure 2.14, LOCAL WATER AQUIFERS 2-26Figure 2.15, RESEARCH CAMPUS AREA 1940 2-27Figure 2.16, PICLKE RESEARCH CAMPUS 1960 2-28Figure 2.17, BALCONES RESEARCH CENTER 1990 2-29Figure 4.1: H/Zr Phase Diagram 4-6Figure 4.2A, Zr-H Transport Cross Section & TRIGA Thermal Neutron Spectra 4-7Figure 4.2B, Fuel Temperature Coefficient of Reactivity 4-7Figure 4.3, Thermal Pressurization in Fuel and Hydriding Ratios 4-9Figure 4.4A, Temperature and Cladding Strength for 0.2% Yield 4-11Figure 4.4B, Temperature, Cladding Strength, and Stress 4-12Figure 4.5, Lower Gird Plate Control Rod Positions 4-14Figure 4.6, Standard Control Rod Configuration 4-15Figure 4.7, Standard/Stepper Motor Control Rod Drive 4-16Figure 4.8, Transient Rod Drive 4-18Figure 4.9a, UT TRIGA Core 4-21Figure 4.9b, Core Top View 4-21Figure 4.10a, 6/7-Element Facility Grid 4-22Figure 4.10b, Upper Grid Plate Cut-out for 6/7-Element Grid 4-22Figure 4.1la, Reflector Top Assembly 4-23Figure 4.11b, Reflector Bottom Assembly 4-23xiii SAFETY ANALYSIS REPORT, TABLE OF CONTENTS 01/2012LIST OF FIGURES PageFigure 4.12b, Graphite Reflector Through port Detail 4-23Figure 4.12c, Graphite Reflector, Radial & Piercing-Beam Ports 4-23Figure 4.13a, Tangential Beam Port Insert 4-23Figure 4.13b, Radial Beam Port insert 4-23Figure 4.13c, Inner Shroud Surface 4-24Figure 4.14, Reflector Component and Assembly Views 4-25Figure 4.15, Fuel Element Adapter 4-26Figure 4.16, Core Support Views 4-27Figure 4.17, Core and Support Structure Views 4-27Figure 4.18, Safety Plate 4-28Figure 4.19a, Pool 4-29Figure 4.19b, Side View 4-29Figure 4.19c, Top View 4-29Figure 4.20, Biological Shielding, Base Dimensions 4-31Figure 4.21, Reactivity Loss with Power 4-34Figure 4.22, Radial Variation of Power Within a TRIGA Fuel Rod. (Data Points from Monte 4-41Carlo Calculations [Ahrens 1999a])Figure 4.23, Critical Heat Flux Ratio (Bernath and Biasi Correlations) 4-44Figure 4.24, Core Power, 45 kW Hot Element 4-45Figure 4.25, Power Coefficient of Reactivity 4-46Figure 4.25: Unit Cell Fuel Element Model 4-50Figure 4.26a, Unit Cell Temperature Distribution (10.5 kW) 4-55Figure 4.26b, Unit Cell Temperature Distribution (22.5 kW) 4-56Figure 4.27, Single Rod Flow Cooling Flow Rate versus Power Level 49°C 6.5 Pool, 4-56Figure 4.28, Comparison of Calculated and Observed Fuel Temperatures 4-58Figure 5.1A, Pool Fabrication 5-4Figure 5.1B, Cross Section 5-4Figure 5.C, Beam Orientation 5-4Figure 5.2, Pool Cooling System 5-4Figure 5.3, Pool Cleanup System 5-11Figure 5.4, Cooling and Cleanup Instrumentation 5-13FIGURE 7.1, CONTROL SYSTEM BLOCK DIAGRAM 7-3Figure 7.2, NEUTRON CHANNEL OPERATING RANGES 7-4Figure 7.3, Auxiliary Display Panel 7-5Figure 7.3, LAYOUT OF THE REACTOR CONTROL CONSOLE 7-6Figure 7.4, CONSOLE CONTROL PANELS 7-8Figure 7.5, TYPICAL VDEO DISPLAY DATA 7-9Figure 7.6, ROD CONTROL PANEL 7-9Figure 7.7, LOGIC DIAGRAM FOR CONTROL SYSTEM 7-13Figure 9.1, Conceptual Diagram of the Reactor Bay HVAC System 9-2Figure 9.2A, Main Reactor Bay HVAC System 9-3Figure 9.2B, Main Reactor Bay HVAC Control System Control 9-4xiv THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR *a** 01/2012SAFETY ANALYSIS REPORT 00FLIST OF FIGURES PageFigure 9.3, Confinement System Ventilation Contrcos 9-7Figure 9.4A, Purge Air System 9-8Figure 9.4B, Purge Air Controls 9-8Figure 9.5A, Storage Well 9-11Figure 9.5b, Fuel Storage Closure 9-11Figure 10.1, Core Grid Plate Design and Dimensions 10-3Figure 10.2, Reactor Core Diagram 10-4Figure 10.3, Central Thimble Union Assembly 10-5Figure 10.4, Three Element Irradiator 10-16Figure 10.5, Rotary Specimen Rack Diagram 10-28Figure 10.6, Rotary Specimen Rack Raceway Geometry 10-28Figure 10.7, Rotary Specimen Rack Rotation Control Box 10-28Figure 10.8, Beam Port Layout 10-30Figure 10.9, A1230 Cryomech Cryorefrigerator and Cold Head 10-35Figure 10.10, Cryomech Cold-Head and Vacuum Box 10-36Figure 10.11, TCNS Vacuum Jacket and Other Instruments (units in cm) 10-36Figure 10.12, Silicone Diode and Heater Relative to Cold-Head 10-37Figure 10.13, Neon and Mesitylene Handling System with Pressure Transducers 10-38Figure 10.14, Shielding around TCNS Facility 10-40Figure 10.15, Thermo MP 320 Neutron Generator at NETL 10-41Figure 10.16, Subcritical Assemblies 10-42Figure 12.1, University Administration 12-2Figure 12.2, NETL Facility Administration 12-2Figure 13.1, Ratio of Radionuclide Inventory to ALl 13-13Figure 13.2, Ratio of Radionuclide Concentration to 10CFR 20 DAC Values 13-14Figure 13.3, FUEL Temperature and Pulsed Reactivity 13-35Figure 13.4A, Pulse Measurements 13-31Figure 13.4B, Fuel Temperature and Peak Pulse Power 13-31Figure 13.5A, Cooling Time 13-37Figure 13.5B, Cooling Time and Power Density 13-38Figure 13.6, Core Model 13-41Figure 13.7A, Bay Model Top View 13-41Figure 13.7B, Bay Model Cross Section 13-41Figure 13.8A, Building Model 13-42Figure 13.8B, MCNP Side View 13-42Figure 13.8C, Top View 13-42xv SAFETY ANALYSIS REPORT, TABLE OF CONTENTS 01/2012Table PageTable 1.1, SHUTDOWN OR DECOMMISSIONED U.S. TRIGA REACTORS 1-10Table 1.2, U.S. OPERATING RESEARCH REACTORS USING TRIGA FUEL 1-10Table 2.1, AUSTIN AND TRAVIS COUNTY POPULATION TRENDS 2-8Table 2.2, TRAVIS COUNTY 2009 AUSTIN POPULATION DENSITY DISTRIBUTION BY ZIP CODE 2-9Table 2.3, 1982 METEOROLOGICAL DATA FOR AUSTIN TEXAS 2-15Table 2.4, HISTORICAL METEOROLOGICAL DATA FOR AUSTIN TEXAS 2-16Table 2.5, HISTORICAL METEOROLOGICAL DATA FOR AUSTIN TEXAS 2-17Table 2.6, HISTORICAL METEOROLOGICAL DATA FOR AUSTIN TEXAS 2-18Table 2.7, HISTORICAL METEOROLOGICAL DATA FOR AUSTIN TEXAS 2-19Table 2.8, TRAVIS COUNTY TORNADO FREQUENCIES 2-20Table 2.9 GROUND WATER ACTIVITY 2-26Table 3.1, SSC Vulnerability 3-2Table 4.1, TRIGA Fuel Properties 4-3Table 4.2, Physical Properties of High-Hydrogen U-ZrH 4-4Table 4.3, U-ZrH Volumetric Specific Heat Capacity (Cp) 4-6Table 4.4, Summary of Control Rod Design Parameters 4-13Table 4.5, Control Rod Information 4-15Table 4.6, Summary of Reactor SCRAMs 4-19Table 4.7, Summary of Control Rod Interlocks 4-19Table 4.8, Upper Grid Plate Penetrations 4-21Table 4.9, Displaced Fuel Spaces 4-22Table 4.10, Lower Grid Plate Penetrations 4-25Table 4.11, Reactor Coolant System Design Summary 4-28Table 4.12, Significant Shielding and Pool Levels 4-32Table 4.13, Control Rod Worth 4-33Table 4.14, Reactivity Values 4-33Table 4.15, GA-4361 Calculation Model 4-35Table 4.16, Selected TRIGA II Nuclear Properties 4-35Table 4.17, UT TRIGA Data 4-36Table 4.18, Critical Heat Flux ratio, Bernath Correlation 4-43Table 4.19, Core Power, 45 kW Hot Element 4-44Table 4.20, Reactivity Limits 4-46Table 4.21, Limiting Core reactivity 4-47Table 4.22, Thermodynamic Values 4-49Table 4.23, Hydrostatic Pressure 4-51Table 4.24, Coolant Temperature for 49°C 6.5 m Pool 4-51Table 4.25a, Outer Cladding Temperature (°C) for 49°C and 6.5 m Pool 4-52Table 4.25b, Inner Cladding Temperature (°C) for 49"C and 6.5 m Pool 4-53Table 4.26a, Heat Flux (Nodes 1-9) 49°C 6.5 Pool, 4-53Table 4.26b, Heat Flux (Nodes 10-15) 49°C 6.5 Pool 4-54Table 4.27, Peak Fuel Centerline Line Temperature (K) 49°C 6.5 Pool, 4-54Table 4-28, Coolant Flow for 1100 kW Operation 4-57Table 4-29, Observed Fuel Temperatures 4-57Table 4-30, Fuel Temperature Comparison 4-58Table 5.1, Reactor Coolant System design Summary 5-2Table 5.2, Heat Exchanger, Heat Transfer and Hydraulic Parameters 5-9xvi THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 0 j 01/2012Table 9.1, Typical Confinement Vent & Purge Parameters 9-4Table 9.2, Reactor Ventilation System Modes 9-5Table 10.1: Composition of Al 6061 1.0-6Table 10.2: Activation Products in Central Thimble 6061 Aluminum Alloy after 60 Year 10-7IrradiationTable 10.3 Characteristic Dimension of UT-TRIGA PTS 10-10Table 10.4: Activation of Pneumatic Transit System Cadmium Liner 10-11Table 10.5: Flux Measurements in Pneumatic Transit System zt 100 kW 10-12Table 10.6: Activity of Three Element Irradiator Cd Liner 10-19Table 10.7: Rotary Specimen Rack Gears 10-27Table 10.8: Items to be Addressed in Safety Analysis for Experiments 10-44Table 11.1, Representative Solid Radioactive Sources 11-5Table 11.2, Representative Radiation Detection Instrumentation 11-15Table 11.3, Representative Occupational Exposures 11-17Table 13.1. Neutronic Properties of TRIGA Mkll ZrH1.6 Fue! Elements. 13-1Table 13.2, Dimensions of TRIGA Mkll ZrH1.6 Fuel Elementsl 13-1Table 13.3, Thermal and Mechanical Properties of TRIGA Mkll ZrH1.6 Fuel Elements and 13-2Type 304 Stainless Steel CladdingTable 13.4, UT TRIGA Core-Conditions Basis for Calculations 13-2Table 13.5, Relevant IOCFR20 Appendix B Values 13-5Table 13.6, SCALE T-6 Sequence Continuous Burnup Parameters 13-8Table 13.7A, 1 MTU Gaseous Fission Product Inventory for 3.5 kW Case (Ci) 13-8Table 13.7B, 1 MTU Particulate Fission Product Inventory (Ci) 13-9Table 13.8A. Gaseous Fission product Release from Single Element (lVCi) 13-10Table 13.8B. Particulate Fission Product Release from Single Element 13-11Table 13.9A, Fraction of Gaseous Fission Product Inventory to 10CFR20 ALl 13-12Table 13.9B, Fraction of Particulate Fission Product Inventory to IOCFR20 ALl 13-12Table 13.10A, Fraction of Instantaneous Gaseous Fission Product Inventory to 1OCFR20 13-14DAC[1]Table 13.10B, Fraction of Instantaneous Particulate Fission Product Inventory to 10CFR20 13-15DAC [1]Table 13.11, DAC Ratios for All Cases 13-16Table 13.12, Reactor Bay Atmosphere Following MHA Compared to Effluent Limit 13-17Table 13.13: BRIGGS URBAN DISPERSION PARAMETERS 13-18Table 13.14, Calculated ?/Q Values 13-21Table 13.15, Reactor Bay Atmosphere Following MHA Compared to Effluent Limit 13-21Table 13.16, Calculated Plume Meander Factor (M) for < 6 m s-1 Winds 13-21Table 13.17, Minimum Dispersion Parameters by Stability Class 13-22Table 13.18, Minimum ?/Q by Stability Class 13-22Table 13.19, Effluent Limit Ratio to Release Concentrations 13-23Table 13.20, Low Power Pulsed Reactivity Response 13-28Table 13.21, Initial Power 880 kW Pulsed Reactivity Response 13-30Table 13.22, Gamma Source Term 13-39Table 13.23, Height/Thickness Dimensions of Unit Cell 13-40Table 13.24, Unit Cell Areas 13-40Table 13.25, Material Characterization 13-40Table 13.26, Post LOCA Doses 13-42xvii SAFETY ANALYSIS REPORT, TABLE OF CONTENTS 01/2012Table 13.27, Calculations Supporting Limits on Fueled Experiments 13-46Table 13.28, Material Strengths 13-48Table 13.29, Container Diameter to Thickness Ratio 13-49xviii THE UNIVERSITY OF TEXAS TRIGA 11 RESEARCH REACTOR 01/2012SAFETY ANALYSIS REPORT, CHAPTER 4 i ET0/4.0 ReactorThis chapter will discuss the reactor core (fuel, control rods, reflector and core support, neutronsource, core structure), reactor pool, biological shielding, nuclear design (normal operatingconditions, and operating limits), and thermal hydraulic design.4.1 Summary descriptionThe University of Texas Nuclear Engineering Teaching Laboratory (NETL) is home to a GeneralAtomics' TRIGA Mark II research reactor. This installation follows 25 years (1963-1988) ofsuccessful operation of a TRIGA reactor at Taylor Hall on the main campus.The basic TRIGA design uses U-ZrH1.6 fuel clad with stainless steel in natural water convectioncooling mode during operation, with a maximum decay heat that can be removed by naturalconvection of either water or air. The reactor is located in an open pool of purified, light waterthat serves as a heat sink during operations at power. Nuclear properties and characteristicscontrol heat generation; thermodynamic characteristics of the fuel and the coolant control heatremoval and temperature response. Maximum fuel temperature is the principle designconstraint. Solubility of hydrogen in the fuel matrix varies with temperature. Consequently,operation at high power levels (i.e., elevated fuel temperature) can cause hydrogen to evolveinto space around the fuel matrix; the hydrogen at elevated temperature can generate pressureinside the cladding. Temperature that produces stress greater than the yield strength for thestainless steel cladding is lower than temperature which leads to phase change or melts U-ZrH1.6.TRIGA fuel has a very strong prompt negative fuel temperature coefficient. Fuel massexceeding critical loading (i.e., excess reactivity) is required to compensate for the negative fueltemperature coefficient, as well as potential experiments, fission product poisons, and fuelburnup. There are several major experiment facilities that could affect core reactivity, asdescribed in Chapter 10. Experiment program requirements vary widely; limits are imposed onthe reactivity effects of experiments. The amount of excess reactivity determines themaximum possible power, and therefore the maximum possible fuel temperature.4.2 Reactor CoreThe University of Texas at Austin TRIGA II reactor core is configured in a hexagonal prismvolume bounded by aluminum plates at the upper and lower surfaces (grid plates), andsurrounded by a cylinder of graphite (aluminum clad) acting as a neutron reflector. Sections ofthe reflector are cut away to support experimental facilities, including beam ports and arotating specimen rack. The core assembly is supported by structural aluminum, and includesan aluminum plate that serves to limit downward travel of control elements.Page 4-1 CHAPTER 4: REACTOR01/20124.2.1 Reactor FuelThe TRIGA fuel system was'developed around the concept of inherent safety, with fuel andcladding designed to withstand all credible envikonmental and radiation conditions during itslifetime at the reactor site. A TRIGA fuel element consists of (A) a central fueled regioncontaining fuel matrix, bounded by an axial reflector and (B) stainless steel end caps at the topand bottom in a stainless steel envelope (cladding sealed by end cap assemblies).Design constraints limit internal fuel element pressure as a function of fuel and claddingtemperature to prevent cladding rupture. The fuel lattice structure that comprises the NETLTRIGA reactor core contains integral inlet and outlet cooling channels in a geometry which,combined with the thermo-physical properties of the fuel element, assure natural convection isadequate to limit maximum steady state operating temperature. ; The TRIGA fuel matrixexhibits a large prompt negative temperature coefficient ofý,reactivity. The maximum fueltemperature resulting from sudden insertion of all available excess reactivity would causepower excursion to terminate before any core damage is possible. Limits on core lattice excessreactivity and individual fuel element temperature therefore are interrelated. The maximumpossible TRIGA fuel fission product inventory is limited by fissionable material loading. Themaximum TRIGA fuel decay heat produced by fission product inventory can be removed bynatural convection in air or water.Handling, transport, and storage of TRIGA fuel elements at the NETL, fresh and irradiated, aredescribed in Chapter 9, Auxiliary Systems.A. Fuel matrixA TRIGA fuel element consists of a central fueled region containing fuel matrix, bounded by anaxial reflector (with a molybdenum disk as a protective interface between the fuel and thelower graphite/axial reflector,- and stainless steel end caps-at the top and bottom with astainless steel cladding.The basic safety limit for the TRIGA reactor system is the fuel temperature; this applies forboth the steady-state and pulse mode of operation. Twe, limiting temperatures are of interest,depending on the type of, TRIGA fuel used. The TRIGA fuel which is considered lowhydride, that with an H/Zr ratio of less than ..5, has a lower temperature !imit than fuel with ahigher H/Zr ratio. Fig. 4.1. indicates that the higher hydride compositions are single phase and.are not subject to the large volume changes associated with the phase transformations atapproximately 530'C in the lower hydrides. Also, it has been noted' that the higher hydrideslack any significant thermal diffusion of hydrogen. These two facts preclude concomitantvolume changes. The important properties of delta phasetU-ZrH are given in Table 4.1.'GA-3618, Thermal Migration of Hydrogen in Uranium-Zirconium Alloys, Marten U. et. Al., GeneralDynamics, General Atomics Division (1962)Page 4-2 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 4R~eETII01/2012Graphite dummy elements may be used to fill grid positions in the core. The dummy elementsare of the same general dimensions and construction as the fuel-moderator elements. They areclad in aluminum and have a graphite length of Table 4.1, TRIGA Fuel PropertiesPropertyMark IIIDimensionsOutside diameter, Do = 2roInside diameter, Dj= 2riOverall lengthLength of fuel zone, LLength of graphite axial reflectorsEnd fixtures and claddingCladding thicknessBurnable poisonsUranium contentWeight percent U235U enrichment percent235U contentPhysical properties of fuel excluding claddingH/Zr atomic ratioThermal conductivity (W cm-' K-1)Heat capacity [T >00C] (J cm3 K')Mechanical properties of delta phase U-ZrI-I0Elastic modulus at 20'CElastic modulus at 650'CUltimate tensile strength (to 650'C)Compressive strength (20'C)Compressive yield (20'C)(1) FabricationA uranium loaded zirconium hydride was found to produce desired moderating characteristicsand acceptably low parasitic neutron absorption with strong temperature feedback and highheat capacity. Feedstock of between (or recycled material) are cast2in controlled atmosphere, high-temperature induction furnace. Fuel element castings are machined to cylinders of approximately 5 inches in length. A centerhole is drilled the length of the cylinder. Additional machining is required for fuel meat to be2 TRIGA International: A New TRIGA Fuel Fabrication Facility at CERCA -Gerard Ilarbormier, Jean-Claude Ottone, CFIRCA,Proceedings of the 1997 TRTR Annual meetingPage 4-3 CHAPTER 4: REACTOR 01/2012fabricated into instrumented fuel assemblies (IFEs, described below) and fuel elementfollowers. The cylinders are heated in a high temperature electric furnace with a hydrogenatmosphere. The exterior and center surface: exposed to hydrogen induces the cylindrical fuelmeat to hydride, with a target Zr:H ratio of 1'.6; :.A pure zirconium filler rod is placed in thecenter hole to maintain nearly uniform thermo-hydraulic properties. Each TRIGA fuel elementcontains three of these machined pieces.Instrumented elements- have three chromel-alumel thermocouples.embedded to about from the centerline of the fuel,;one at the; axial center plane, and one each at above and: below the center plane. Thermocouple leadout wires pass through a sealin:the upper end fixture, and a leadout tube provides a watertight conduit carrying the leadoutwires above the water surface in the reactor tank.Followers are machined to an outer radius of 1.25 in. (0.318 m) and 1.35 in. (0.0343 m) for thetransient rod (air filled follower) and the standard rods (fuel fo!lowers) respectively.(2) Physical PropertiesThe zirconium-hydrogen system is essentially a simple eutectoid, with at least four separatehydride phases. The delta and epsilon phases are respectively face-centered cubic and face-centered tetragonal hydride phases. The two phase delta + epsilon region exists betweenZrHI.64 and ZrH1.74 at room temperature, and closes at ZrH1.7 at 455°C. From 455°C to about10500C, the delta phase is supported by a broadening range of H/Zr ratios. Other importantproperties observed for the delta phase U-ZrH are listed in Table 4.2.The ratio of Zr-H plays a significant role ir determining physicai properties. The H:ZR materialhas a cubic structure in the delta-phase at ratios greater than 1.4. in lower H:Zr ratios (< 1.5) aphase change occurs at about 955°F (535°C) with large dens:ity differences between the phasesleading to potential for deformation (swelling, and cracking). For hydrogen to zirconium atomratios greater. than 1.5, the matrix is single phase (delta or epsilon) and does not exhibit phaseseparation with ,thermal cycling; Thermal diffusion of hydrogen is rrtinimal in higher ratios aswell, minimizing potential for dceformaticrn from evolutiion of hydrogen gas. Any hydrogen gasis in equilibrium with: the matrix, substantially retained by the cladding, Losses through thecladding from hydrogen migration are about 1% for cladding temperature about 9300F (5000C).Table 4.2, Physical Properties of High-Hydrogen U-ZrHProperty Temperature Value UnitsThermal Conductivity, 93°C -650°C 0.22 W cm20°C 9.1x106 psiElastic Modulus 6500C 6.0x106 psiUltimate Tensile Strength 20°C 2.4 x104 psiCompressive Strength 20°C 6.0 x104 psiCompressive Yield 20'C 3.5 x104 psiHeat of Formation 298°C 37.75 kcal g-molrPage 4-4 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 00o°NlTi 01/2012SAFETY ANALYSIS REPORT, CHAPTER 4 000o nE tAt ratios greater than 1.6 there can be a shift to higher density tetragonal. Higher hydridecompositions are single phase. and are not subject to the large volume changes associated withthe phase transformations at approximately 530TC as in :the: lower hydrides. The stabilityextends from the minimum on the scale (OTC).to the maximum onwthe:scale (950TC), indicatingno volume changes from morphology which might stress cladding occur for a target ratio of 1.6other than thermal expansion. Significantly, zirconium hydrides at these ratios lack anysignificant thermal diffusion of hydrogen under isothermal conditions' Undernon-isothermalconditions, hydrogen migrates from high temperature regions to low temperature regions, withequilibrium disassociation pressures lower after redistribution. Hydrogen' dissociates slightlyfrom the fuel matrix at high temperatures, and is re-absorbed :into the matrix at lowertemperatures, with the equilibrium hydrogen dissociation pressure a function of both thecomposition and temperature. The equilibrium hydrogen dissociation pressure is governed bythe composition and temperature. For ZrH1.6, the equilibrium hydrogen pressure is oneatmosphere at about 7600C. Hydrogen dissociation pressures of hydrides are similar in alloysup to about 75 weight per cent uranium. For the delta and epsilon phases, dimensionalchanges from hydrogen migration are not significant. In the delta .phase, equilibriumdisassociation pressures are related by:Klog pK +With:P pressure (atm)T= temperature (K)K1= -3.8415 + 38.6433-X -34.2639.X + 9.282122X3K2= -31.2981 + 23.5741,X -.6.0280.X2X= hydrogen to zirconium atom ratioAt a ratio of 1.7 the equilibrium disassociation pressure corresponds to a temperature of about1400F (300°C). The density of ZrH. decreases as hydrogen ratio increases; from low ratios tothe delta phase (H:Zr of 1.5) the density change is high' with little cha&ngefor further increases.Massively. hydrided bulk density' is-reported to be about 2% Ylower than x~ray diffractionanalysis. For TRIGA-fuel with aZr:H ratio of 1:1.6, the uranium density,.volumefractioh, andweight fraction are related by:WUp(A)=0.177-0.125. WUandWU= 0.177-pu (A)1 + 0. 125.-pu (A)Page 4-5 CHAPTER 4: REACTORI01/2012CHPE,: ECO 0121po, (A) = 19.07. V)J(A)wherepu(A)= Uranium densityWU -Uraniuri Weight fractionV1= volume'fraaction of uranium in the U-ZrH1.6alloyThermal conductivity determined from short-pulse heating techniques. Using thermaldiffusivity values, density, and specific heat the thermal conductivity of uranium zirconium witha Zr:H ratio of 1:1.6 is 0.042 +/-+0.002 ca[-1 s-5 cm 'C -.Volumetric specific heat is a function of temperature and composition. Table 4.3 lists values forvariations in Zr:H and w% U based on a O°C reference, showing variation less than 10%.Table 4.3, U-ZrH Volumetric Specific Heat Capacity (Cp)ZrH W% U Value UnitsU-ZrH1.6 8.5 2.04 + 4.17x10 W s .cmr3U-ZrH.720 2.17 + 4.36x10 W s -cr3850wP600450I I hF0@4.8 ImI1ZwI I I I I I -- -02 04 06 08 to 12HYDROGEN CONTENT (Di ZraFigure 4.1: H/Zr Phase Diagram0 L. 16ILBPage 4-6:

THE UNIVERSITY OF TEXAS TRIGA 11 RESEARCH REACTOR 01/2012SAFETY ANALYSIS REPORT, CHAPTER 4(3) Operational PropertiesThe neutronic properties of ZrH are the primary motivation for incorporation in TRIGA fueldevelopment. The morphology of ZrH, in particular hydrogen diffusion in the material, imposeslimits during operation. Ultimately, personnel exposure related to TRIGA fuel is limited duringnormal operations and abnormal events by retaining fission products in the fuel elements. It iswell known that zirconium can undergo a reaction with water that releases hydrogen, withsubsequent potential for a mixture that can be detonated. Such a reaction has the potential-,torelease a large fraction of fission product inventory of affected fuel elements, but is not likelygiven characteristics of operation and properties of the fuel matrix. Fuel element changes occurduring operation from thermal stress, which can affect fuel performance. Fuel claddingprevents migration of fission products for the fuel element, but in the absence of cladding it isnot likely that all fission products will escape the fuel meat. Finally, thermal effects related tofuel matrix from steady state and pulsing operations are considered.(4) Neutronic Properties A large fraction of neutron moderation occurs through interactionswith hydrogen in the fuel matrix. The zirconium hydride structure has a profound effect onneutron scattering at low energies because of zirconium-hydrogen binding, with distinct latticeenergy levels of 0.13 eV and about 0.25 eV found in scattering experiments. Thermal neutronsthat interact wi~th.hydrogen in the lattice (where neutron energy is below the lattice energies)therefore have potential to gain energy. Because the fission cross section has 1/v dependencein the thermal range, increasing thermal neutron energy decreases fission probability. If fueltemperature increases, thermal excitation creates more of these relatively high-efnergy latticecenters as indicated in Fig. 4.2a. When the rate of fission is high enough to create elevated fueltemperatures, the elevated fuel temperatures decrease the rate of fission. This phenomenon isresponsible for an extremely, high feedback of negative reactivity from fuel temperatureillustrated in Fig. 4.2b. Maximum possible fuel temperature and maximum theoretical powerlevel are therefore a function of the amount of fuel in the reactor.I00USTAINLESS STEEL CLAD-12 8.5 WT-% U-ZrHj.6O CORE" 00 400-C-a a260401ba20' L 00.01 0.... ........... .l.0NEUTRON ENERGY (eVI0 200 400 600 800 1000 /POO'TEMPERATURE tIC)Figure 4.2A, Zr-H Transport Cross Section & TRIGA Figure 4.2B, Fuel Temperature Coefficient ofThermal Neutron Spectra ReactivityPage 4-7 CHAPTER 4: REACTOR 1 01/2012(5) Fuel Morphology & Outgassing As noted previously, during fuel fabrication the ratio ofhydrogen to zirconium is enhanced by thermally induced diffusion in an atmosphere ofpressurized hydrogen. During reactor operation, temperature gradients influence hydrogendiffusivity to promote outgassing, bounded by temperature induced pressurization of thehydrogen in free volume of the cladding. Pressure inside the fuel element does not intrinsicallypose a challenge to fuel element integrity, and will be considered as part of claddingperformance in a later section. At a given temperature, higher H:Zr ratios (in the absence ofphase change) exhibit more pressure at a given temperature in a well behaved relationship,shown in Fig. 4.3. Thermal diffusion is accelerated at higher temperatures, but the expansion offree hydrogen gas at higher temperatures also produces more partial gas pressure in the freevolume of the element. Calculations performed with a higher mass fraction of uranium result in3an increase in the partial pressure of hydrogen by as much as a factor of four.The fuel rod diameter is on the order of the path length of neutron from generation toabsorption, and the mean free path for thermal neutrons within the fuel rod is not large.Consequently, a large fraction of power in a TRIGA fuel element is produced close to the outersurface of the fuel. Fuel rod temperature gradient during normal, steady-state operations ismonotonically decreasing from a peak at the center of the fuel rod. Routine power changesoccur at a rate that allows quasi-steady state thermal equilibrium, but pulsing operations donot. As a consequence, power distribution and development of temperature gradients insteady-state operations is fundamentally different compared to fast transient (pulsing)operations.In general, gas pressure during the transient of pulsing operations is expected to be less thanduring steady state. Diffusion rates are finite, and the diffusion coefficient for thermal diffusionof hydrogen in zirconium4 (ranging from 4x105 to 2x108 cm2 s-1, and requiring days toequilibrate) lags the time cons-cant for the temperature changes. The temperature gradientduring the transient peaks near the surface of the fuel rod rather than the center, and rapidlyvanishes as the system comes to equilibrium. Therefore thermal gradients in pulsing biashydrogen diffusion towards the center of the fuel rod with only a small region near the surfacehaving a gradient that promctes outgassing. Surface cooling from endothermic gas emissionlowers the surface temperature and therefore tends to iower the diffusion constant at the fuelrod surfaces. Re-absorption occurs where hydride surfaces are at relatively lowertemperatLres. There is evidence that low permeability oxide films on fuel surfaces retard masstransfer. Local heat transfer effects cause the surface temperature to be lower than that whichwould occur during adiabatic conditions.3 Journal of NUCLEAR SCIENCE and TECHNOLOGY, Vol. 37, No. 10, p. 887-892 (October 2000); Estimation ofHydrogen Redistribution in Zirconium Hydride under Temperature Gradient4 Congreso Internacional de Metalugia y Materiales, Primeras Jornadas Internacionales de Materiales Nucleares(19 al 23 de Octubre de 2009, Buenos Airesm Argentina; Some Peculiarities of Hydrogen Behavior and DelayedHydride Cracking in Zirconium Based Reactor Alloys, Shmakov, R.N. SinghPage 4-8 THE UNIVERSITY OF TEXAS TRIGA Im RESEARCH REACTOR 000 01/2012SAFETY ANALYSIS REPORT, CHAPTER 4 0O%031027-- Z .' 5/1,1/ o,,o~AoDATA FROM GA-8129AND NAA-SR-9374600 700 80o 900 ICo I 100 200 1300TEMPERATURE ( C)Figure 4.3, Thermal Pressurization in Fuel and Hydriding RatiosLong term operations with steady state fuel temperatures exceeding 750'C (1023*K) may havetime- and temperata ure-dependent fuel growth.5 Mechanisms contributing to the growth areidentified as fission recoils and gaseous fission products, strongly influenced by thermalgradients. Analysis of steady state operating fuel,temperatures is provided in section 4.6, withpulsing operations fuel temperatures in Appendix 4.1.(6) Zr water reaction Among th.e.chemical properties of U-ZrH: and ZrH, :the. reaction rate of thehydride with water is of particular interest. Since the hydriding reaction ýis exothermic, waterwill react more readily with zirconium than with zirconium hydride systems. Zirconium isfrequently used in contact with water in reactors, and the zirconium-waterreaction is not asafety hazard.Experiments carried out at GA Technologies show that.the zirconium hydride systems have arelatively low chemical reactivity with respect to water and air .These tests have involved the,quenching with water of both powders and solid specimens of.U-ZrH after~heating to as high as;850°C, and of solid U-Zr alloy after, heating to as high as 1200*C. Tests have also been made to'determine the extent to which fission.products are removed from the surfaces of thefuel'elements at room temperature. Results prove that, .because of the high resistance to leaching, alarge fraction of the fission products is retained in even completely unclad U-ZrH fuel..:(7) Mechanical Effects At room temperature the hydride is like ceramic and shows littleductility. However, at the elevated temperatures of interest for pulsing, the material is found tobe more ductile. The effect of very large thermal stress on hydride fuel bodies has beenobserved in hot cell observations to cause relatively widely spaced cracks which tend to be5General Atomics Technical Report E-117-8336NUREG/CR-2387 Credible Accidents for TRIGA and TRIGA Fueled Reactors, S. C. Hawley,S. C. and Kathren, R. L.,PNL-4208 (1982)Page 4-9I CHAPTER 4: REACTOR A 01/2012either radial or normal to the central axis and do not interfere with radial heat flow. Since thesegments tend to be orthogonal, their relative positions appear to be quite stable. Duringfabrication, a molybdenum disk is placed between the lowest fuel mass and the lower axial-graphite reflector, minimizing potential for interaction that might affect the graphite and causeposition changes in fuel meat that has developed surface imperfections. Anticipatedmechanical effects from operation of the reactor are not expected to create conditions thatchallenge fuel performance.(8) Fission Product Release Early in development of U-ZrHx fuel, experiments were performed7to determine the potential of the evolution of fission products from the fuel matrix. Zr-U-Halloy foils were irradiated in a materials test reactor and a post irradiation test conducted, withwater flowing across the surface of the foil to remove fission products for analysis. The testwas performed for 1 day and for 8 days with the total fractional fission product loss calculatedto be between 10-7 and 10-s from preferential leaching of radionuclides, with gasses evolvingfrom depths of 2.6 plm in the foil, and particulate from 22 A. Acceptable8 upper values forrelease fraction are 1.0 x 10-4 for noble gases and iodine contained within the fuel, and of1.0 x 10-6 for particulates (radionuclides other than noble gases and iodine). Experiments byGeneral Atomics [Simnad et al., 1976] indicate a value of 1.5 x 10-5 for noble gases, which is inSARs for other reactor facilities [NUREG-1390, 1990].B. CladdingThe fuel matrix is enveloped by a cylindrical 304 stainless steel shell,welded to stainless steel fittings at each end (end caps). The cladding is the principal barrier torelease of those fission products that migrate to escape the fuel matrix surface. As notedpreviously, the free hydrogen in the space within the fuel element pressurizes the interior ofthe fuel element when fuel temperature is elevated during reactor operations. Power levelsare acceptable if they do not result in temperatures that produce stress from the gas pressurethat challenges the integrity of the cladding. A cylinder is considered a thin shell if wallthickness is less than about 10% of the radius and the classic equation for hoop stress createdby internal pressure is:o= P.r/twhere:oe is the hoop stressP is internal pressurer is inside radiust is the wall thickness7 General Atomic report GA-655, Uranium-Zirconium-Hydride Fuel Elements, Merten, Stone, Wallace(1959)8 NUREG/CR-2387, op. cit.Page 4-10 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 4R~eETLI01/2012For stress is times the internal pressure. Fig.4.4A provides temperature dependent ultimate strength and the 0.2% yield, and Fig. 4.4Bshows where the hoop stress induced by the internal pressure intersects with ultimatestrength. This intersection corresponds to a fuel temperature of 950'C for claddingtemperatures greater than 500'C.cc103400500600 700800 900 1000 i'300TEMPERATURE (0C)Figure 4.4A, Temperature and Cladding Strength for 0.2% YieldPage ,4-11 CHAPTER 4: REACTORI01/2012Therefore, if fuel and cladding temperature remains below 950°C with claddingtemperatures greater than 500°C, the stainless- steel cladding will not fail fromoverpressure. For cladding temperatures less than 500TC, hydrogen pressure frompeak fuel temperature of 1150TC would not produce a stress in the clad in excess of itsultimate strength. The limiting fuel temperature and pressure is therefore the designbasis for the UT TRIGA fuel. TRIGA fuel with a hydrogen to zirconium ratio of at least1.65 has been pulsed to temperatures of about 1150TC without damage to the clad9.1OS.ULTIM~ATE 5TRENGTH 304. SS1: S T R E NZ r H 1 .6 51-io2.10 2_-;500 600 700 800 900 1000 1100TEMPERATURE (*C)Figure 4.4B, Temperature, Cladding Strength, and Stress9 "Annual Core Pulse Reactor," General Dynamics, General Atomics Division report GACD 6977 (Supplement 2),Dee. J. B., et. A].Page 4-12 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR N0ETL 01/2012SAFETY ANALYSIS REPORT, CHAPTER 4 0.00 _There are several reasons why the gas pressure should be less for the transient conditionsthan the equilibrium condition values would predict. For example, the gas diffusion rates arefinite; surface cooling is believed to be caused by endothermic gas emission which tends tolower the diffusion constant at the surface. Reabsbr ption takes placelconcurrently on thecooler hydride surfaces away from the hot spot. There is evidence for a low permeabilityoxide film on the fuel surface. Some iocal, heat transfer does take place during the pulse timeto cause a less than adiabatic true surface temperature.4.2.2 Control Rods and Drive MechanismsThe control rods and drive mechanisms consist of (A) control rods, (B) standard, (or stepper)control rod drives, (C) transient rod drives, (D) control functions, and (E) system operation. TheUT TRIGA reactor was installed with 4 control rods, three standard rods magnetically coupled tothe control rod drive, and one pulse rod pneumatically coupled to the control rod drive. One ofthe standard rods, the regulating rod, is capable of being either automatically controlled withinstrumentation and control systems described in Chapter 7 or manually from the reactorcontrol console. The other control rods are manually shimmed. Principle design parametersfor the control rods are provided in Table 4.4.A. Control RodsThe standard/stepper control rods (regulating and shim) are sealed 304 stainless steel tubesapproximately 43 in. (109 cm) long by 1.35 in. (3.43 cm) in diameter in which the uppermost 6.5in. (16. 5 cm) section is an air void, followed by 15 in. (38.1 cm) of a neutron absorber, solidboron carbide. Standard control rods have a fuel follower attached so that as the control rod iswithdrawn from the core the water channel is filled with a fuel element as illustrated in Fig. 4.6.The fuel follower, 15 in. (0.381 cm) of U-ZrH1.6 fuel, is immediately below the neutron absorberof the standard control rods. The bottom 6.5 in. (16.5 cm) of the standard control rod is an airvoid. Table 4.4 summarizes control rod design parameters.Table 4.4, Summary of Control Rod Design ParametersCladdingMaterial Aluminum SS 304OD 1.25 in. 3.18 cm 1.35 in. 3.43 cmLength 36.75 in.. 93.35 cm 43.13 in. 109.5 cm ...Wall thickness 0.028 in. 0.071 cm 0.02 in. 0.051 cmPoison SectionMaterial Boron CarbideOD 1.19 in. 3.02 cm 1.31 in. 3.32 cmLength 15 in. 38.1 cm 14.25 in. 36.20 cmFollower SectionMaterial Air U-ZrH.16OD 1.25 in. 3.18 cm 1.31 in 3.34 cmLength 20.88 in. 53.02 cmPage 4-13 CHAPTER 4: REACTOR 1 01/2012The transient (also called safety-transient or pulse) rod is a sealed, 36.75 in. (93.35 cm) long by1.25 in. (3.18 cm) diameter tube containing boron in graphite as a neutron absorber. Below theabsorber is an air filled follower section. The absorber section is 15 in. (38.1 cm) long and thefollower is 20.88 in. (53.02 cm) long. The transient rod passes through the core in a perforatedaluminum guide tube. The tube receives its support from the safety plate and its lateralpositioning from both grid plates. It extends approximately 10 in. (25.4 cm) above the top gridplate. Water passage through the tube is provided by a large number of holes distributedevenly over its length. A locking device is built into the lower end of the assembly.Control rods are withdrawn out of the core through the upper grid plate; when fully insertedthe followers extend down through the lower grid plate. All fuel element position penetrationsin the upper grid plate are identical; the lower grid plate (an excerpt in Fig. 4.5, fully describedlater in Chapter 4) has a set of 11 penetrations in the C and D rings (shaded in gray and black inFig. 4.5, black representing the current configuration) with the same diameter as the upper gridplate. One of these penetrations in reserved for the central thimble (position Al) while theothers are available for use as control rod positions. A safety plate is mounted below the lowergrid plate as shown in Fig. 4.6, so that the control rod cannot exit the core region in thedownward direction.Figure 4.5, Lower Gird Plate Control Rod PositionsControl rod worth is principally a function of control rod dimensions and location, experimentfacilities in the core, with lessor influence by fuel and control rod burnup. Estimated controlrod from the 1991 preliminary safety analysis report is provide in Table 4.5, along with theworth of each control rod as measured in June 2011. Sections of the control rod are separatedand secured by 1-inch magneform fittings.Page 4-14 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 4NETII01/2012Table 4.5, Control Rod InformatonRod Location Diameter Estimated,(1991). Current (2011)In. cm. %Ak/k $STransient Rod C Ring 1.25 3.18 2.1 3.00 3.10Regulating Rod C ring 1.35 3.43 2.6 3.71 2.82Shim 1 D ring 1.35 3.43 2.0 2.86 2.52Shim 2 D ring 1.35 3.43 2.0 2.86 3.07AirB4CU-ZrHAir-U °T.I JFLNK ,16.5 cm38.1 cm38.1 cm6.5cm c..Figure 4.6, Standard Control Rod ConfiguratioInA threaded fitting at the end of each control rod connects to a series of shafts that connect tocontrol rod drive mechanisms mounted' on a bridge that spans the reactor pool. The topsection of the connecting shafts for standard rods passes through a hole in the bottom of atube supported by the control rod drive housing. The tube is designed with slots that provide ahydraulic cushion for the rod during a scram, and also prevent the bottom of the control, rodfrom impacting the safety plate.The shaft is secured to a cylinder that rests on the bottom of the housing when the rod is fullyinserted. The top of the cylinder is secured to an iron core, engaged by an electromagnet forfail-safe control. The electromagnet is at the bottom of a small shaft controlled by the controlrod drive mechanism. When the electromagnet is energized, the iron core is coupled to thedrive unit.Page 4m15 CHAPTER 4: REACTORI01/2012CHPE :RACO 121The top section of the transient rod is connected to a single acting pneumatic cylinder whichoperates on a fixed piston, that couples the connecting rods to the drive. The transient roddrive is mounted on a steel frame that. bolts to the bridge. Any value from zero to a maximumof 15 in. (38.1. cm,):. of rod may:ibe withdrawn from the core; rod travel is limited byadministrative control-not to exceed to the maximum licensed step insertion of reactivity.B. Standard Control Rod DrivesThe rod drive mechanism for the standard rod drives is an electric stepping-motor-actuatedlinear drive equipped with a magnetic coupler and a positive feedback potentiometer. Astepping motor drives a pinion gear and a 10-turn potentiometer via a chain and pulley gearmechanism. The potentiometer is used to provide rod position information.~MAGNET WIRE CONDUIT-MAGNET DOWNADJUSTMENT SCREW'MAGNET DRAW TUBE -MOTOR BIAS ADJUSTMENTCENTER SWITCHROD DOWN LIMIT SWITCH-'.. Y_-,TIAOUNTING PLATEWIRE CONDUITPULL-RODRODMOTORPULL-RODDRAW TUBE*PULLLOCK.CONNECTING RODFigure 4.7, Standard/Stepper Motor Control Rod DrivePage 4-16 THE UNIVERSITY OF TEXAS TRIGA 11 RESEARCH REACTOR 000 o f 01/2012SAFETY ANALYSIS REPORT, CHAPTER 4 00_ IThe pinion gear engages a rack attached to the magnet draw tube. An electromagnet, attachedto the lower end of the draw tube, engages an iron 3rm-nature. The armature is screwed andpinned into the upper end of a connecting rod that terminates at its lower end in the controlrod. When the stepping motor is energized (via the rod' control UP/DOWN switch on thereactor control console), the pinion gear shaft rotates, thus raising the magnet draw tube, thearmature and the connecting rod will raise with the draw tube so that the control rod iswithdrawn from the reactor core. In the event of a reactor scram, the magnet' is de-energizedand the armature will be released. The connecting rod, the piston, and the control rod will thendrop, thus reinserting the control rod.Stepping motors operate on phase-switched direct current power. The motor shaft advances200 steps per revolution (1.8 degrees per' step). Since current is maintained -on the'motorwindings when the motor is not being stepped, a high holding torque is maintained. The torqueversus speed characteristic of a stepping motor is greatly dependent on the drive circuit used tostep the motor. To optimize the torque characteristic for the motor frame size, a TranslatorModule was selected to drive the stepping motor. This combination of stepping motor andtranslator module produces the optimum torque at the operating speeds of the control roddrives. Characteristic data for the drive indicate a possible travel rate of 33 ipm (1.40 cm/s).Measurements of the actual rate provide a speed of 27 ipm (1.14 cm/s)..C. Transient Control Rod DriveThe safety transient control rod*drive is operated with a pneumatics rod drive. Operation ofthe transient rod drive is controlled from'the reactor control console. The transient rod is ascrammable rod operated in both pulse and steady-state modes of reactor operation. Duringsteady state operation, the transient rod will function as an alternate safety rod with aircontinuously supplied to the rod. Rod position is thus controlled by'Operation of an electricmotor that positions the air drive cylinder. The position of the transient control rod and itsassociated reactivity worth will generally dictate removal of the rod as the first step of a startupfor steady-state operation. Rod withdrawal speed is about 28 ipm (1.E9cm/s).The transient rod drive is a single-acting pneumatic cylinder with its piston attached to thetransient rod through a connecting rod assembly., The piston rod passes through an air seal atthe lower end of the cylinder. Compressed air is supplied to the lower end of the cylinder froman accumulator tank when a three -way solenoid valve located in the piping between theaccumulator and cylinder is energized. The compressed air. drives the piston upward in thecylinder and causes the rapid withdrawal of the transient rod from the core. As the piston rises,the air trapped above it is pushed out through vents at the upper end of the cylinder. At theend of its travel, the piston strikes the anvil of an oil filled hydraulic shock absorber, which has aspring return, and which decelerates the piston at a controlled rate over its last 2 in. (5 cm.) oftravel. When the solenoid is de-energized, a solenoid valve cuts off the compressed air supplyand exhausts the pressure in the cylinder, thus allowing the piston to drop by gravity to itsoriginal position and restore the transient rod to a position fully inserted in the reactor core.Page 4-17 CHAPTER 4: REACTORI01/2012V EN .TVALVEFigure 4.8, Transient Rod DriveThe extent of transient rod withdrawal from the core during a pulse is determined by raising orlowering the de~coupled cylinder, thereby controlling the distance the piston travels when air isapplied. The cylinder has external threads running most of its length, which engage a series ofball bearings contained in a ball-nut mounted in the drive housing. As the ball-nut is rotated bya worm gear, the cylinder moves up or down depending onr!the direction of worm gear rotation.A ten-turn Potentiometer driven by the worm shaft provides a signal indicating the position ofthe cylinder and the distance the transient rod will be ejected from the core. Motor 'operationfor pneumatic cylinder positioning is controlled by a switch on the reactor control console. Themagnet power key switch on the control console power supply prevents unauthorized firing ofthe transient rod drive.Attached to and extending downward from the transient rod drive housing is the rod guidesupport, which serves several purposes. The air inlet connection near the bottom of thecylinder projects through a slot in the rod guide and prevents the cylinder from rotating.Attached to the lower end of the piston rod is a flanged connector that is attached to the rodassembly that moves the transient rod. The flanged connector stops the downward movementof the transient rod when the connector strikes the damp pad at the bottom of the rod guidesupport. A microswitch is mounted on the outside of the guide tube with its actuating leverPage 4-18 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 4* eNETII01/2012extending inward through a slot. When the transient rod is fully inserted in the reactor core,the flange connector engages the actuating lever of the microswitch and indicates on theinstrument console that the rod is in the core. In the case of the transient rod a scram signal de-energizes the solenoid valve which supplies the air required to hold the rod in a withdrawnposition and the rod drops into the core from the fu!! out position in less than 1 second.D. Control FunctionsInstrumentation and controls provide protective actions through the control rod system, asdescribed in Table 4.6. A trip signal from the reactor protection system or the reactor controlsystems will deenergize the electro magnets and the pulse rod air solenoid valve previouslydescribed which allows gravity to insert the control rods.Table 4.6, Summary of Reactor SCRAMsLimiting Trip SetpointMeasuring Channel Steady Pulse Actual SetpointStateSS -1050 (N PP/NP) 1080 NMMaximum thermal power 1100 kW 2000 MW Pl -1 910 NPPulse -1910 NPPPower Channel High power 110% 110%Detector High Voltage 80% 8C%High Fuel Temperature 550°CMagnet current lossManual ScramDAC and CSC watchdog timersIn addition, the reactor control system (described in Chapter 7) has interlocks to preventvarious conditions from developing. Table 4.7 is a summary of the functions.Table 4.7, Summary of Control Rod InterlocksINTERLOCK SETPOINT FUNCTION/PURPOSEInhibit standard rod motion if nuclear instrumentS2 startup channel reading is less than instrumentsensitivity/ensure nuclear instrument startup channelis operatingPulse Rod Interlock Pulse rod inserted Prevent applying power to pulse rod unless rodinserted/prevent inadvertent pulsePrevent withdrawal of more than 1 rod/LimitMuti dWithdrawal s , me maximum reactivity addition rate (does not apply inautomatic flux control)Prevent withdrawing standard control rods in pulsePulse Mode Interlock Mode switch in Hi Pulse mode.Pulse-Power Interlock 10 kW Prevent pulsing if power level is greater than 10 kWThese safety settings are conservative in the sense that if they are adhered to, the consequenceof normal or abnormal operation would be fuel and clad temperatures well below the safetylimits indicated in the reactor design bases. Because of the conservatism in these safetyPage 4-19 CHAPTER 4: REACTOR 01/2012settings, it is reasonable that at some later date less restrictive safety system settings could beassigned in conjunction -with upgrading of the reactor to operate at higher steady-state powerlevels or in the pulsing mode while using the same fuel and core configuration.Administrative limitations are imposed for the excess reactivity, transient conditions andcoolant water temperature as follows:1) Maximum core excess reactivity of 4.9% Ak/k ($7.00) with a shutdown margin of at least0.2% Ak/k ($0.29) with the most reactive control rod fully withdrawn,2) Maximum transient control rod worth of 2.8% Ak/k ($4.00) with a limit of 2.2% Ak/k($3.14) for any transient insertion, and3) Core inlet water temperature of 48.9°C.E. Evaluation of the Control Rod SystemThe reactivity worth and speed of travel for the control rods are adequate to allow completecontrol of the reactor system during operation from a shutdown condition to full power. TheTRIGA system does not rely on speed of control for reactor safety; scram times for the rods aremeasured periodically to monitor potential degradation of the control rod system. Theinherent shutdown mechanism (temperature feedback) of the TRIGA prevents unsafeexcursions and the control system is used only for the planned shutdown of the reactor and tocontrol the power level in steady state operation. A scram. does not challenge the controlintegrity or operation, or affect the integrity or operation of other~reactor systems.4.2.3 Neutron Moderator and Reflector (Core Structure)The UT TRIGA core is supported within a reflector assembFy. The reflector assembly supports(A) an upper grid plate, (B) core barrel and reflector, and (C) lower grid plate, shown in Fig.4.9a/b. The upper and lower grid plates provide alignment and support for the fuel elements.A. Upper grid plateThe upper grid plate provides alignment for fuel elements and control rods, and (in conjunctionwith the top fuel assembly) space for cooling flow. The.top grid. plate is fabricated from acircular aluminum plate 5/8 inches (1.59 cm.) thick and 21..6 in. (55.245 cm) diameter, anodizedto resist wear and corrosion. The top of the upper grid plate is 59 in. (150 cm.) above thebottom of the pool. diameter areestablished on a triangular pitch of 1.714 in. (4.35 cm), separated by radial fuel arraysintegrated on the same pitch, although the radial arrays do not extend to the edge of the core.The holes position the fuel-moderator and graphite dummy elements, the control rods andguide tubes, the pneumatic transfer tube, and the central thimble. Small 0.203 in. (8 mm) holesat various positions in the top grid plate permit insertion of wires or foils into the core to obtainPage 4-20 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR No T 0 01/2012SAFETY ANALYSIS REPORT, CHAPTER 4 0 NET_flux data. The flux probe holes are counter sunk/chamfered to (820) to 0.31 in. (11 mm). Thecenter fuel element position is reserved as an experimental facility. The outermost fuelpositions in the radial arrays are not fabricated for fuel insertion. Upper grid plate penetrationsare summarized in Table 4.8.01<Ito. 2W' /Figure.4.9a, UT TRIGA Core Figure 4.9b, Core Top ViewThe grid plate is supported by a ring welded to the top inside surface of the reflector container.The ring is fabricated with bosses-that hold alignment pins to engage and ;center the upper gridplate using % in. (0.953 cm) holes centered along each of the hexagonal faces of the G ring fuelpositions.Table 4.8, Upper Grid Plate Penetrations-Penetration Function..: Size "Fuel Elements -, 1.505 in. (3.8227 cm), diameter3Telement -.1.2 in. (3.048 cm). radius !6/7-Element 2.2 in. (5.588 cm) radiusUpper grid plate alignment 3/8 in. (0.9525 cm) diameterFlux probes 0.203 in. (0.5156 cm) diameter..Fuel positions are :indexed by !etters denoting a "ring" where elements are colfinear withrespect to the adjacent radial array fuei positions; A is the centralring position and G is furthestfrom the center. One radial array is used as a reference position, and the fuel positions rangefrom 1 at the index to the maximum value for the ring, except for the G ring. Since the verticesof the G ring are not used as fuel positions, index numbers for the G ring vertices are not used.Circular cutouts to replace fuel element positions are fabricated using two different. designs, 3-element fuel position facilities and 7-element fuel position facilities (6-element for the facilityencompassing the, central thimble since the central thimble does not contain fuel). Page 4-21 CHAPTER 4: REACTORI01/2012 The inserts mesh in slots milled in the circular grid platecutouts; engagement secures the insert. There are two locations fabricated for each design.The 6/7 element facilities permit specimen as large as 4.4 in, (11.8 cm) and the 3 elementfacilities permit specimen as large as 2.4 in. (6.1 cm).In addition to the experiment facilities that replace fuel positions, the current coreconfiguration reserves one position for a neutron source, one position for a pneumatic facility,and four positions for control rods. Table 4.9 summarizes fuel element positions displaced orpotentially displaced by core equipment. For control rods, only currently used positions areidentified; there are alternate positions useable for control rods.Table 4.9, Displaced Fuel SpacesFacility Core Location . Page 4-22 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 400 4 nETLI01/2012B. ReflectorThe core is surrounded by a graphite radial reflector for neutron economy. In addition,graphite cylinders are positioned within the fuel cladding above and below the active fuelregion.(1) Radial Reflector. The radial reflector is a 10.2 in. (25.91 cm) graphite ring with an innerdiameter of 21 % in. (54.93 cm) that is 21 13/16 in. (54.40 cm) tall, surrounded by aluminum.The reflector is fabricated in a top and bottom section. Lifting bosses are located on the surfaceof the top section (Fig. 4.9a), with flat welded plates tying the top and bottom sections to thelift points. Angle plate structures are welded on the outer perimeter as points to secure thepower level detectors. A 3 inch (7.62 cm.) wide well is fabricated in the top section (Fig. 4.11b),and blocks with threaded penetrations are welded at the inner perimeter of the well to allowsecuring the rotary specimen rack (an experimental assembly) in the well.Figure 4.11a, Reflector Top AssemblyFigure 4.11b, Reflector Bottom AssemblyThe lower radial reflector is constructed of graphite contained in a welded aluminum canister.The graphite is machined to accommodate three beam ports oriented radial from the center ofthe reactor core, with one "through port" (Fig. 4.12a) and a 10 in. (25.3 cm.) cylinder cut fromthe inner surface to allow a 3 inch wide experimental facility surrounding the core.1Figure 4.12a, Graphite Reflector, Through PortFigure 4.12b, Graphite Reflector Through port DetailPage 4-23 CHAPTER 4: REACTORI01/2012CHAPTER 4: REACTOR 01/2012------a--Figure 4.12c, Graphite Reflector, Radial & Piercing-Beam PortsThe through port has a rectangular water-filled cut-out between the core shroud and the beamport penetration (Fig. 4.12b). Aluminum canisters that mate with the beam ports are nested inthe reflector in two of the beam ports, one radial and one tangential (Fig. 4.12c, Fig. 4.13a/b).The third beam port (radial) penetrates the core shroud (Fig. 4.13c).Figure 4.13a, Tangential Beam Fort InsertFigure 4.13b, Radial Beam Port inertFigure 4.13c, Inner Shroud Surface(2) Graphite Rods. Graphite dummy elements may be uFAd to fP!. grid positions not filledby the fuel-moderator elemerts or other core compound, ,. They are of the same generaldimensions and construction aý. the fuel-moderator elements, !ýut are f;lled entirely withgraphite and are clad witi :m.(3) Axial Reflector. Graphite cylinders are placed above and below the fuel in the fuelelements. Fuel element construction was previously discussed.C. Lower grid plateThe lower grid plate (Fig. 4.14) provides alignment for fuel elements and control rods, and (inconjunction with the top fuel assembly) space for cooling flow. The lower (or bottom) gridPage 4-24 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 40E~I01/2012plate is fabricated from a circular aluminuLr. p*2t.e ý..75 inches (3.81 cm.), anodized to resistwear and corrosion. The top of the bottom piate is S.9 in. (25.19 cm.) above the bottom ofthe pool. The bottom grid plate is fabricated with fuel position penetrations and penetrationsmatching the flux probe holes on toe sarne, center 5s the upper grid plate, but also containspenetrations that support alignment cf the 3, G-, af:d 7 element facilities (Table 4.10). All but 11fuel penetrations in the lower grid plate are sarnfller than the diameter of the fuel element andchamfered to provide a surface supporting triflutes on the bottom of the fuel elementelements.Table 4.10, Lower Grid Plate PenetrationsPenetration Function SizeCentral thimble 1.505Control Rod 1.505Flux Hole Probes 8 mm3-Element Alignment 3/8 in.Lower grid plate alignmentLower Grid Plate SupportLower Grid Plate11Reflector Canister Bottom View Grid Plate in Core ShroudFigure 4.14, Reflector Component and Assembly ViewsTen lower grid plate penetrations are the same diameter as the penetration in the upper gridplate, providing clearance for the central thimble and control rods. Since only 4 controls rodsPage 4-25 CHAPTER 4: REACTOR01/2012are installed, unused control rod positions (i.e., large diameter holes) can be used for fuel withan adapter to support positioning the fuel above the lower grid plate (Fig. 4.15).Figure 4.15, Fuel Element Adapter4.2.4 Neutron Startup SourceThe reactor license permits the use of sealed neutron sources, including a is a standard sealed neutron source, encapsulated in stainless steel. The sourceis maintained in an aluminum-cylinder source holder of approximately the same dimensions as afuel element. The source holder is manufactured as upper and lower (threaded) sections. Thetop of the lower section is at the horizontal centerline of the core. A soft'aluminum ringprovides sealing against water leakage into the cavity. The sourceholder may be positioned in any one of the fuel positions defined by the upper and lower gridplates. The upper end fixture of the source holder is similar to that of the fuel element; thesource holder can be installed or removed with the fuel handling tool. In addition, theupper end fixture has a small hole through which one end of a stainless steel wire may beinserted to facilitate handling operation from the top of the tank.4.2.5 Core support structure.The core support structure includes (A) a platform supporting the reflector and core structure,and (B) a "safety plate" thaftprevents the control rods in a failure mode from falling out of thecore.A. Core Support PlatformThe reflector assembly rests on a platform (Fig. 4.16) constructed of structural angle 6061-T5aluminum with a 3 in. x 3 in. x % in. (7.62 cm x 7.62 cm x 0.953 cm) web. Aluminum 6061-T651plate is used for safety plate support pads (% in., 1.905 tmP), cross braces (% in., 0.953 cm.), andplatform support pads (Y/ in., 1.27 cm.). Angle aluminumr is inserted 9 iii. (22.86 cm)frorn twoedges to support the safety plate, with angle bracing on the edges perpendicular to the safetyplate supports.Page 4-26 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 4IMeETII01/2012Core Support Top ViewCore Support Side ViewCore Support Side ViewFigure 4.16, Core Support ViewsThe platform top surface is 30 X in. X 30 Y4 in., with the top surface 16 Y4 in. above the poolfloor. Surfaces are matte finished for uniform appearance with shot cleaning and peeningusing glass beads (MIL-STD -852).-h 1Core and Support Structure AssemblyCore and Support Assembly IsometricFigure 4.17, Core and Support Structure ViewsB. Safety plateThe safety plate (Fig 4.18) limits the distance that a control rod can fall to less than 17.44 in.(44.30 cm) below the top surface of the lower grid plate. The safety plate is an aluminum plateY2 in. (1.27 cm.) thick, 12 in. (30.48 cm) X 13.5 in. (34.29 cm), anodized to resist wear andcorrosion (MIL-A-8625 TYPE II, with exception that abrasive and corrosive testing not required).The top of the safety plate is 7.75 in. (3.05 cm.) above the bottom of the pool. As previouslydescribed, the bottom grid plate has a set of through-penetrations for optional placement ofcontrol rods. A special adapter is required to support fuel elements when these locations areused for fuel. The adapters have a central alignment pin that fits within holes in the safetyplate, and an offset keeper-pin that prevents the adapter from rotating around the central pin.Page 4-27 CHAPTER 4: REACTOR 01/2012Figure 4.18, Safety Plate4.3 Reactor PoolThe reactor pool is a 26 foot, 11.5 in. (8.2169 m) tall tank formed by the union of two half-cylinders with a radius of 6 1/2z feet separated by 6 Y feet (1.9812 m). The bottom of the poolis at the reactor bay floor level. The reactor core is centered on one: of the half-cylinders.Normal pool level is 8.179 (26.57 ft.) meters above the bottom of the pool, with a minimumlevel of 6.5 m (21.35 ft.) required for operations. Volume of water in pool (excluding thereflector, beam tubes and core-metal) is 40.57 mi3 and 32.50 m3 for the nominal andminimum-required levels. Table 4.11 summarizes reactor coolant system design.Table 4.11, Reactor Coolant System Design SummaryMaterial Aluminum plate (6061)Reactor Tank Thickness Y4 in. (0.635 cm)Coolant LinesCoolant PumpHeatExchangerTypical Heat ExchangerOperating ParametersVolume (maximum) 11000 gal (41.64 M3)Pipes .. Aluminum 6061Iron-Plastic Liner, 316 SSBall and StemFittings ..Aluminum (Victaulic)Type .CentrifugalMaterial Stainless SteelCapacity 250 gpm (15.8 Ips)Type Shell & TubeMaterials (shell) " Carbon steelMaterials (tubes) 304 stainless steelHeat Duty Flow Rate (shell), " Flow Rate (tubes) .-.. Tube Inlet ....I u e uOutlet Shell Inlet Shell OutletPage 4-28 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 00 T 01/2012SAFETY ANALYSIS REPORT, CHAPTER 4 _ 00 00The pool (Figs. 4.19a/b/c) is fabricated from sheets of 0.25 in. (0.635 cm) 6061 aluminum in 4vertical sections welded to a Y2 in. thick aluminum plate. Full penetration inspection wasperformed on tank components during fabrication, including 20% of the vertical seam welds,100% on the bottom welds (internal and external to the pool volume), and 100% on thebeam port weld external to the pool volume. A single floor centerline seam weld was used;a sealed channel was welded under the seam and instrumented through a /4 in. NPT threadedconnection to perform a leak test during fabrication. A 2 in. X 2 in. X Y4 in. (square) aluminumchannel was rolled and welded to the upper edge of the tank.Beam port penetrations are fabricated around the core to allow extractior of radiatiorbeams to support experiments. The beam ports are centered 90.2 cm (35 in.) above the poolfloor, 7.2 cm (2.83 in.) below the core centerline. The section of the beam ports that are anintegral part of the pool include an in-pool section, interface with the pool wall, and a section.extending outside of the pool.In pool sections are 6 in. (15.4 cm) in diameter, with a 0.635 cm (0.25 in.) wall thickness. Thein pool section for BP 1 and 5 is 6 in. (15 cm), while the remaining in-pool beam port sectionsare much longer. Supports (2 in. X 2 in. X Y4 in. aluminum angle bracket) are welded at thebottom of the pool and directly onto. BP 2, 3, and 4 because of the extended lengths. BP 2and 4 terminate at the outer surface of the reflector, while BP 3 extends into the reflector,terminating at the inner shroud. BP 2 terminates in an oblique cut, and extendsapproximately 43 cm (16.94 in.) into the pool with the support 12.7 cm (5 in.) from the in-core end. BP 3 extends 73 cm (28.75 in.) into the pool with the support 37.62 cm (14.8125in.) from the in-pool end. BP 4 extends 43 cm into the pool (16.94 in.) with the support 7.62cm (3 in.) from the in pool end. Beam port 1 and 5 are aligned in a single beam line. A flighttube inserted into BP 1/5 extends through the reflector near the core shroud; BP 1 and 5 areequipped with a bellows to seal a neutron flight-tube. Beam ports 2, 3, and 4 are sealed atthe in-pool end. BP 2 is tangential to the core shroud, offset 34.29 cm (13 /2 in.) from corecenter rotated 300 with respect to BP 3. Beam port 3 is 90' with respect to BP 1/5, aligned tothe center of the core. Alignment of BP 4 is through the core center, rotated 60' from BP 3.The beam port interface with the pool wall includes a reinforcing flange on the inner poolwall. The flange is 3/8 in. thick, 11 in. in diameter. The flange is welded on the outerPage 4-29 CHAPTER 4: REACTOR 01/2012diameter to the pool wall and on the inner diameter to the beam port tube.The beam ports extend approximately 15.24 cm (6 in.) outside of the area define by the poolwalls. A stainless steel (304) ring is machined for a slip fit over the extension. The ring iswelded to 6 5/8 in. diameter stainless steel pipe (SST 304W/ASTM 312) extending the flighttube for the beam port into the biological shielding.The floor of the pool has four welded pads for the core and support structure. As noted, thein-pool beam port supports are welded to the pool floor.Detection of potential pool leakage could occur in a number of ways.1. Pool water level is maintained approximately 8.1 m above the pool floor, andmonitored with an alarm on the control room console. A sudden decrease inpool water will create a condition that alerts the reactor operator at thecontrols.2. Losses to evaporation are compensated by makeup water. Makeup water usageis closely monitored, and changes in makeup requirements or increases inmakeup water that do not correspond to power level operation are a primarypool-leak indicator.3. French drains around. the reactor pool shielding ,foundation are collected in a.sump, and sampled periodically. Increases in radiation levels from the sump..(particularly tritium) could indicate pool leakage.,4.4 Biological ShieldPool water system and. shield structure (Fig. 4.20), design combine to control the effectiveradiation levels from, the, operation of the reactor.: One goal of the design is a radiologicalexposure constraint of 1 mrem/hour for accessible areas of the pool and shield system. Dose.levels assume a full power operation level of 1.500 megawatts (thermal). Radiation doses abovethe pool' and at specific penetrations into or through the shield may exceed the design goal. Thereference. case design is a solid structure without any system penetrations. Tank assembly is by shop fabrication. A protective layer of epoxy paint andbitumen coal tar pitch with paper provides a barrier between the aluminum pool tank and thereactor shield concrete.Page 4-30 THE UNIVERSITY OF TEXAS TRIGA 11 RESEARCH REACTOR N 0001/2012SAFETY ANALYSIS REPORT, CHAPTER 4 0.. thick foundation pad supports the reactor pool and shield structure.Standard weight concrete,' comprises the foundation pad. High densityconcrete, of Five beam tubes atthe level of thereactor -provide experimental access, to:reactor neutron and gamma radiations. Two of thetubes combine to penetrate the complete reactor pool and shield structure from one side tothe other side. Special design features of the beam tubes'are beam plugs, sliding lead shutters,bolted cover plates, and gasket seal for protection against reactor radiation and coolant leakagewhen the tubes are not in use. Beam port details are discussed in Chapter 10. A summary ofsignificant component elevations and control functions is provided in Table 4.12. .Page 4-31 CHAPTER 4: REACTORI01/2012CHAPTER 4: REACTOR 01/2012Parameter of InterestCONCRETE PADFLOORSAFETY PLATEGRID PLATECORE BOTTOMBEAM PORT CLCORE CLCORE TOPGRID PLATEMAIN LOWER SHIELDINGTRANSITIONAL CONCRETESHIFT TO HIGH DENSITY C(MIN CORE LEVEL (TS)VACUUM-BREAKERSLOW POOL LEVEL SCRAMLOW POOL LEVELLOW POOL LEVEL ALARMNORMAL POOL LEVELHIGH POOL LEVELHIGH POOOL LEVEL ALARMTOP OF TOP LEVELTable 4.12, Significant Shielding and Pool LevelsLevei Notes(meters)STEP )NCRETE o'1 1p4.5 Nuclear DesignThe characteristics and operating parameters of this reactor,ý have been calculated andextrapolated using experience and data obtained from existing TRIGA reactors as bench marksin evaluating the calculated data. There are several TRIGA systems with essentially the samecore and reflector relationship as this TRIGA so the values presented here are felt to beaccurate to within 5% but, of course, are influenced by specific core configuration details aswell as operational details. An operational core of 3 fuel followed controlrods, and one air followed control rod is to be arranged in 5 rings with a central, water filledhole. Dimension of the active fueled core, approximated as cylinder, 15 in. (cylinder radius is calculated as the average radius ofa hexagonal fuel array with 4.5.1 Normal Operating ConditionsReactivity worth of core components is generally determined by calculation and/or comparisonof the reactivity worth associated with the difference in the reactivity worth of control rodpositions in the critical condition, component installed and component removed. The 1992 UTSAR provided data indicating estimated worth of the control rods (Table 4.13). Control rodworth is influenced by core the experiment configuration, with significant impact from the largein core irradiation sites. Table 4.13 provides the worth of the control rods in the currentconfiguration (3 element facility in Eli, F13, and F14). Change in core configuration requirevalidation that control rod worth is not affected by the experiment facility, or re-establishmentPage 4-32 THE UNIVERSITY OF TEXAS TRIGA 11 RESEARCH REACTOR 00 A TI 01/2012SAFETY ANALYSIS REPORT, CHAPTER 4 ....of the control rod worth followed by verification that the limiting conditions for operation aremet.Table 4.13 Control Rod WorthReference Current (2011),Control Rd .Position Worth Position ; Worth.Transient rod C ring 2.1% Ak/k $3.00 C-1 $3.10Regulating rod C ring 2.6% Ak/k $3.71 C-7 $2.82Shim 1 D ring .2.0% Ak/k $2.86 D-14 $2.52Shim 2 D ring 2.0% Ak/k $2.86 D-6 $3.074.5.2 Nominal Reactivity Worth ValuesReactivity values for core components based on calculations and observations are provided inTable 4.14, with Technical Specifications values in bold face type. Current values are based onmeasurements; nominal values are calculations frOm indicated sources.'Table 4.14, Reactivity Values$"TS CURRENT NOMINALParameter LIMIT VALUE VALUE 10 Reactor Reference Data Notebook, Safety Analysis, report Table 4-5; SAR Table 4-6 indicates CT Fuel $0.90, CTVoid -$0.15, PNT Void -$0. 10, RSR void -0.20"3-Element Experiment Authorization12 Significant deviation from values in 3-Element Experiment Authorization (cf. E-Ring -$0.50 & D-Ring $0.95)Page 4-33 CHAPTER 4: REACTORI01/2012Table 4.14, Reactivity ValuesTS CURRENT NOMINALLIMIT VALUE VALUE " "4.5.3 Reactor Core PhysicsThe performance of the TRIGA core was evaluated by General Atomics, as described below.The basic parameter which allows the TRIGA reactor system to operate safely with large stepinsertions of reactivity is the prompt negative temperature coefficient (Fig. 4.21) associatedwith the TRIGA fuel and core design. This temperature coefficient allows a greater freedom insteady-state operation as the effect of incidental reactivity changes occurring from theexperimental devices in the core is greatly reduced.44. 03.0 i--, 2.0.5tIInI00 200 100 600 800 1000POWER (KW)Figure 4.21, Reactivity Loss with PowerPage 4-34 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 4SNETLI01/2012A. Reference CalculationsA reference calculation of neutron flux distribution across the core was performed by GeneralAtomics13.The calculations were accomplished on an .BM-7090 using General Atomics(diffusion theory based) codes GAMBLE and GAZE, and GAIV1-I. GAM-l is a fast neutron (usingP1 treatment), temperature dependent (using methods developed by Nordhiem) cross sectioncalculations for neutrons above 1 eV. GATHER-I was used to calculated cross sections below 1eV. Homogenization was accomplished by the transport theory code DSN for group-dependentdisadvantage factors (a second homogenization was accomplished for inhomogeneities in cellswith control rods). No attempt was made to account for spatial variations in coretemperatures. Basic core data for the calculations is provided in Table 4.15, with selectednuclear properties in Table 4.16. The model varies from the UT TRIGA reactor in specification ofcontrol rods, with one poison and three aluminum followers, where the UT TRIGA uses onealuminum and three poison followers; since this effects only the homogenization for twodiscrete cells, the results for core wide parameters is valid. UT TRIGA data is provided in Table4.17.Table 4.15, GA-4361 Calculation ModelRadius Area Volume volumeFractionCell Region in. cm crn2 cm3U-ZrH1.7 0.7175. 1.822 10.429 397.34 0.6308SS Cladding 0.7375 1.873 0.592 22.56 0.0358Water 0.9032 2.294 5.511 209.98 0.3334TOTAL na na 16.532 629.88 1.0000Table 4.16, Selected TRIGA II Nuclear PropertiesNumber of cells 80 91Fuel Temperature 23°C 200°C1 eV to 10 MeV1;a 0.00660 0.00675if 0.00135 0.00135Flux/watt 2.46x107 2.21x107p'll 0.9405 0.94810 to 1 eV1" 0.0873 0.0794if 0.0526 0.0472Flux/watt 1.11x107 1.08x107% of fissions 94.6 94.5Vave cm/s 2.73x105 2.94x10sEae eV 0.0391 0.0455NOTE 1: Resonance escape probability13 GA-4361, Calculated Fluxes and Cross Sections for the TRIGA Reactors, G. B. West. August 1963Page 4-35 CHAPTER 4: REACTOR 1 01/202Table 4.17, UT TRIGA DataCore ConfigurationRef Cold Clean Critical Loading : 64 elementsRef Operational Loading 90 elementsActulal'Initial CriticalityFuel' elemeht pitch 0.043536 cm* " Coolantlvolume to cell ratio .32.86%* .--, %.. , ' ... .Fuel Elem ents* Cladding, .. SS 304.Fuel matrix .J-ZrH..6Fuel Mass 2.5 kgUranium fraction 8.5%Enrichment 19.5%,Nuclear ParametersPrompt neutron lifetime,( f) 41 psEffective delayed neutron.fraction (13) 0 .007Prompt negative temperaturecoefficient (a) 1x104 Ak/k0CB. Prompt Negative Temperature Coefficient .GA Technologies, the designer of the reactor, has developed techniques to calculate thetemperature coefficient accurately and therefore predict the transient behavior of the reactor.This temperature coefficient arises primarily from a change in the disadvantage factor resultingfrom the heating of the uranium zirconium hydride fuel-moderator elements. The coefficient isprompt because the fuel is intimately mixed with a large portion of the moderator and thus fueland solid moderator temperatures rise simultaneously. A quantitative calculation of thetemperature coefficient requires knowledge of the energy dependent distribution of thermalneutron flux in'the reactor.The basic physical processes which occur when the fuel-'-moderator elements are heated can bedescribed as follows: the rise in temperature of the hydride increases the probability.that athermal neutr'on in the fuel element will gain energy from an excited state of an oscillatinghydrogen atom in the lattice. As the neutrons gain e.nergy-from the ZrH, their mean free path isincreased appreciably. Since the average chord length. in:the fuel element is comparable to amean free path, the probability of escape from the fuel element before capture.is increased. Inthe water the neutrons are rapidly thermalized so that the ,capture and escape probabilities arerelatively insensitive to the energy with which the neutron enters the water. The* heating of themoderator mixed with the fuel thus causes the spectrurn to h1airden more in the fuel than in thewater. As a result, there is a temperature dependent disadvantage factor for the unit cell in thecore which decreases the ratio of absorptions in the fuel to total cell absorptions as the fuelelement temperature is increased. This brings about a shift in the core neutron balance, giving aloss of reactivity.Page 4-36 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR l' Tl 01/2012SAFETY ANALYSIS REPORT, CHAPTER 4 n ÷The temperature coefficient then, depends on spatial variations of the thermal neutronspectrum over distances of the order of a mean free path with large changes of mean free pathoccurring because of the energy change in a single collision. A quantitative description of theseprocesses requires a knowledge of the differential slow neutron energy transfer cross section inwater and zirconium hydride, the energy dependence of the transport cross section ofhydrogen as bound in water and zirconium hydride, the energy dependence of the capture andfission cross sections of all relevant materials, and a multigroup transport theory reactordescription which allows for the coupling of groups by speeding up as well as by slowing down.Calculation work on the temperature coefficient made use of a group of codes developed by GATechnologies: GGC-3"4, GAZE-21s, and GAMBLE-516, as well as DTF-IV17, an Sn multigrouptransport code written at Los Alamos. Neutron cross sections for energies above thermal (>1eV) were generated by the GGC-3 code. In this code, fine group cross sections (-100 groups),stored on tape for all commonly used isotopes, are averaged over a space independent fluxderived by solution of the 81 equations for each discrete reactor region composition. This codeand its related cross-section library predict the age of each of the common moderatingmaterials to within a few percent of the experimentally determined values and use theresonance integral work of Adler, Hinman, and Nordhein to generate cross sections forresonance materials which are properly averaged over the region spectrum. Thermal crosssections were obtained in essentially the same manner using the GGC-3 code. However,scattering kernels were used to describe properly the interactions of the neutrons with thechemically bound moderator atoms. The bound hydrogen kernels used for hydrogen in thewater were generated by the THERMIDOR code18 using thermalization work of Nelkin19.Earlythermalization work by McReynolds et a120 on zirconium hydride has been greatly extended atGA Technologies21, and work by Parks resulted in the SUMMIT t251 code, which was used togenerate the kernels for hydrogen as bound in ZrH. These scattering models have been used topredict adequately the water and hydride (temperature dependent) spectra as measured at theGA Technologies linear accelerator as shown in section 4.2.1 (A).'4 General Atomics Report GA-7157, "Users and Programmer Manual for the GGC-3 Multigroup CrossSection Code," General Dynamics, General Atomic Division (1967)is General Atomics Report GA-3152 "GAZE-2: A One-Dimensional, Multigroup, Neutron Diffusion TheoryCode for the IBM-7090," Lenihan, S. R., 'General Dynamics, General Atomic Division (1962)16 General Atomrics Report GA-818, "GAMBLE-5 -A program for the Solution for the MultigroupNeutron-Diffusion Equations in Two Dimensions, with Arbitrary Group Scattering, for the UNIVAC-1108Computer," Dorsey, J. P. and R. Foreloch, General Dynamics, General Atomic Division (1967)17 USAEC ReportLA-3373, DTF-IV, A FORTRAN-IV Program for Solving the Multigroup Transport Equationwith Anisotropic Scatterings, Los Alamos Scientific Laboratory, new Mexico (1965)18 "THERIMIDOR- A FORTRAN II Code forCalculating the Nelkin Scattering Kernel for Bound Hydrogen (Amodification of Robespierre),"Gulf General Atomic, Inc. (unpublished data) Brown, H. D., Jr.19 "Scattering of Slow Neutrons by Water," Phys. Rev., 11, 741-746, Nelkin, M. S. (1960)20 "Neutron Thermalization by Chemically-Bound Hydrogen and Carbon," Proc. 2nd Intl. Conf. PeacefulUsed at Energy (A/Conf. 15/F/1540), Geneva, IAEA (1958)21 General Atomics Report GA-4490 Neutron Interactions in Zirconium Hydride, Whittenmore, W. L.,General Dynamics, General Atomic Division (1964)Page 4-37 CHAPTER 4: REACTOR 01/2012Qualitatively, the scattering of slow neutrons by zirconium hydride can be described by a modelin which the hydrogen atom motion is treated as an isotropic harmonic oscillator with energytransfer quantized -in multiples. of 70..14 eV. More precisely, the SUMMIT model uses afrequency spectrum with two branches, one for the optical modes for energy transfer with thebound proton, and the other for-the acoustical modes for energy transfer with the lattice as awhole. The optical modes are represented as a. broad frequency band centered at 0.14 CV, andwhose width is adjusted to fit the cross section data of Woods et al. 1281. The low frequencyacoustical modes are assumed to have a Debye spectrum with a cutoff of 0.02 eV and a weightdetermined by an effective mass of 360.This structure then allows a neutron to slow down by thetransition in energy units of 0.14 eVas long as its energy is above 0.14 eV. Below 0.14 eV the neutron can still lose energy by theinefficient process of exciting acoustic Debye type modes in which the hydrogen atoms move inphase with the zirconium atoms, which in turn move in phase with one another. These modestherefore, correspond to the motion of a group of atoms whose mass is much greater than thatof hydrogen, and indeed even greater than the mass of zirconium. Because of the largeeffective mass, these modes are very inefficient for thermalizing neutrons, but for neutronenergies below 0.14 eV they provide the only mechanism for neutron slowing down within theZrH. (In a TRIGA core, the water also provides for neutron thermalization below 0.14 eV.) Inaddition, in the ZrH it is possible for a neutron to gain one or more energy units of -0.14 eV inone or several scatterings, from excited Einstein oscillators. Since the number of excitedoscillators present in a ZrH lattice increases with temperature, this process of neutron speedingup is strongly temperature dependent and plays an important role in the behavior of ZrHmoderated reactors.Calculations of the temperature coefficient were done in the following steps:a. Multigroup cross sections were generated by the GGC-3 code for a homogenized unitcell. Separate cross-section sets were generated for each fuel element temperature byuse of the temperature dependent hydride kernels and Doppler broadening of the 238Uresonance integral to reflect the proper temperature. Water at room temperature wasused for all prompt coefficient calculations.b. A value for. k- was computed for each fuel element temperature by transport cellcalculations, using the P1 approximation. Comparisons have shown 54 and S8 results tobe nearly identical. Group dependent disadvantage factors defined as (Dg'/ (1gc (regioncell) were calculated for each cell region (fuel, clad, and water).c. The thermal group disadvantage factors were used as input for a second GGC-3calculation where cross sections for a homogenized core were generated which gave thesame neutron balance as the thermal group portion of the discrete cell calculation.Page 4-38 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR *g0 fl0 T I " 01/2012SAFETY ANALYSIS REPORT, CHAPTER 4 0 °° * .---Od. The cross sections for an equivalent honmogenized core were used in a full reactorcalculation to determine the contribution to the temperature coefficient due to theincreased leakage of therma! neutrons into <the, reflector with increasing hydridetemperature. This calculation requires severa! thermnalgrbups, butiransport effects areno longer of major concern. Thus, reactivity calculati6ns :s a fbiit-tion of fuei elementtemperature have been done on the entire reactor with the use of diffusion theorycodes.Results from the above calculations indicate that more than 50% of the temperature coefficientfor a standard TRIGA core comes from the temperature-dependent disadvantage factor or "celleffect", and ~20% each from .Doppler broadening of the 238U resonances and temperaturedependent leakage from the core. This produces a temperature coefficient of ~ -0.01%/°C,which is rather constant with temperature.Because of the prompt negative temperature coefficient a significant amount of reactivity .isneeded to overcome temperature and .allow the reactor to operate at the higher power levelsin steady-state operation. Fig. A.19 shows. the. relationship of reactor power level andassociated reactivity loss to achieve a given power level.4.5.4 Operating LimitsThe core-wide operating limits associated with nuclear design are based on spatialdistributionof neutron flux that determines*the local peak power production. Therefore (A)., the peakingfactors are required to determine (B) the limiting core configuration. Core reactivity limits (C)are established by Technical Specifications and used as a basis for evaluating performance andcapabilities. -..A. Core Peaking FactorsThe core.is generally modeled as aýrght cylinder. Neutron flux,.varies along the. axis of acylindrical reactor using periodic Bessel functions. Neutron flux varies: radially in a cylindricalreactor using period sine functions. The product of these two functions. provides a relationshipbetween average core power and the maximum power at a location within the core. Neutronflux and fission rate also varies significantly across the radius of a TRIGA fuel element; thecomplexities of the system do notlend themselves to reasonable analytic description.Core Radial Peaking Factor. Classically, the radial hot-channel factor for a cylindrical reactor(using R as the physical radius and Re as the physical radius and the extrapolation distance) isgiven 2by:22 Elements of Nuclear Reactor Design, 2nd Edition (1983), J. Weisman, Section 6.3Page 4-39 CHAPTER 4: REACTOR 01/20121 .202 * (R./[2.4048"(4 DlHowever, TRIGA fuel elements are on the order of a mean free path of thermal neutrons.Consequently, there is a significant change in thermal neutron flux across a fuel element.Calculated thermal neutron flux data23 indicates that the ratio of peak to average neutron flux(peaking factor) for TRIGA cores under a range of conditions (temperature, fuel type, water andgraphite reflection) has a small range of 1.36 to 1.40. Therefore, actual power produced in themost limiting actual case is 14% less than power calculated using the assumption.Core Axial Peaking Factor. The axial distribution of power in the hottest fuel element issinusoidal, with the peak power a factor of rn/2 times the average, and heat conduction radialonly. The axial factor for power produced within a fuel element is given by:g(z) 1.514"cos(',*zin which £ = L / 2 and f.., is the extrapolation length in graphite, namely, 0.0275 m. The valueused to calculate power in the limiting -location-within the fuel element is therefore 4% higherthan power calculated with the actual peaking factor. Actual power produced in the mostlimiting actual case is 4% less than power calculated using the assumption; therefore calculatedtemperatures will bound actual temperatures.Core Local Peaking Factor. The location on the fuel rod producing the most thermal power withthermal power distributed over N fuel rods is therefore:Pq N D, LB. Powerdistribution:within a Fuel Element. -The radial and axial distribution~of the power'within a.fuel.element is given byq .'(r, z)= qfL(r)g(z)in which r is measured from the vertical axis of the fuel element and z is measured along theaxis, from the center of the fuel element. The axial peaking factor follows from the previousassumption of the core axial peaking factor, but (since there is a significant flux depressionacross a TRIGA fuel element) distribution of power produced across the radius of the fuel theradial peaking factor requires a different approach than the previous radial peaking factor forthe core. The radial factor within a fuel element is given by:23 GA-4361, Calculated Fluxes and Cross Sections for TRIGA Reactors (8/14/1963), G.' B. WestPage 4-40 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 4000IIETI000I01/2012f a+cri+ er2I + br + dr2in which the parameters of the rational polynomial approximation Iare derived from flux-depression calculations for the TRIGA fuel24.Values for the coefficients are: a = 0:824461 b =-0.26315, c = -0.21869, d = -0.01726, and e = +0.04679. The fit is IElustrated an Fig. 4.20.1.31.21.1.0L0.900.800.700.0 0.20 0.40 0.60 0.80 1.0 1.2 1.4 1.6 1.8 2.Figure 4.22, Radial Variation of Power Within a TRIGA Fuel Rod.(Data Points from Monte Carlo Calculations [Ahrens 1999a])0C. Power per rodThe Bernath correlation25 calculates critical heat flux as:Q AIBO (Two -TB)Where the convection heat transfer coefficient for "burnout" condition is, determined by:hBo .10~990 (De2+ D) + SLOPE .VWith two possible values for the "SLOPE" term:(1) IF De< 0.1 ft.,24 Report KSUNE -Investigation of the Radial Variation of the Fission-Heat Source in a TRIGA Mark III FuelElement Using MCNP, Ahrens, C., Department of Mechanical and Nuclear Engineering, Kansas StateUniversity, Manhattan, Kansas (1999)25 ANL/RERTE/TM-07-01, Fundamental Approach to TRIGA Steady state Thermal-Hydraulic CHFAnalysisPage 4-41 CHAPTER 4: REACTOR 01/201248SLOPE = 0.SLOPE =9 9+ -D.eAnd the burnout wall temperature term is calculated:P VTwBo = 57- In(P) -P54 +P +15 42The CHF heat flux in is p.c.u./hr-ft , the heat transfer coefficient corresponding to the CHF in2p.c.u./hr-ft -C, is the wall temperature at which CHF occurs in °C, T is the local bulk coolantbtemperature in TC, D hydraulic diameter of the coolant passage in feet, D is the diameter of theeheater surface (heated perimeter divided by n) in feet, P is the pressure in psia, and V is thevelocity of the coolant in ft/s. Substituting equivalent terms into the CHF equation results in:[OIt10890. ( De, 8 P_ + i ,r(= 80 57-in(P)-54- -T[A80 De+ DiJDe* )~l~)~+5 4 BWhere A is the flow area and WP the wetted perimeter, hydraulic diameter is calculated:4-ADe=-WP(1) Wetted perimeter:WP =- r -D *,2(2) Flow area:A=PITCH2 f---r4 14 2 2)Page 4-42 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 40000 ETII00001/2012TRACE calculates CHFR using the Biasi correlation, however the more accepted CHF correlationfor TRIGA reactors is provided by Bernath26.TRACE calculations completed as described insection 4.6 provides thermal hydraulic parameters used to calculate critical heat flux using theBernath correlation (and the ratio of the heat flux to the critical heat flux, CHFR). The results ofcalculations using heat flux and temperature data for 490C water at .6.5 m level is provided inTable 4.18. The minimum CHFR versus power level is provided in Fig. 4.22. As illustrated, theCHFR values agree well and remain much greater than 2 at power levels up to 22.5 kW per unitcell.Table 4.18, Critical Heat Flux ratio, Bernath CorrelationkW1.53.04.56.07.59.010.512.013.515.016.518.019.521.022.51106.261.044.335.429.926.023.120.819.117.616.415.414.514.013.7293.553.438.630.925.822.319.817.816.315.013.913.012.211.811.5383.347.234.026.922.519.417.115.313.912.811.811.010.39.99.74 5 '74.9 68.642.2. 38.530.2 .27.623.6 21.619.7 17.916.9 15.314.8 13.413.2 12.011.9 10.810.9 9.9i0.0 9.19.3 8.48.6 7.88.3 7.58.1 7.3666.137.026.420.817.214.712.911.510.39.48.68.07.47.27.07 863.7 61.535.7: .34.425.5 24.420.0 19.2.16.5 15.814.1 :13.512.4, .11.811.0 10.59.9 9.49.0 Q 8.68.2 7.87.6 7.27.1 6.76.8 6.56.6 6.3963.535.525.319.716.3.13.912.110.89.78.88.07.46.86.66.41065.736.626.020.316.814.212.411.09.99.08.27.57.06.76.61167.237.226.220.316.614.C12.1.10.79.58.67.87.16.56.36.11272.539.727.821.317.414.612.510.99.78.67.87.06.46.26.01378.942.929.723.919.216.013.611.810.49.27.97.06.4*6.15.91486.8'46.832.224.619.616.213.811.810.39.18.07.16.46.15.91596.85i.935.526.721.317.614.812.610.99.58.37.36.56.16.026 ANL/RERTE/TM-07-01, op. cit.Page 4-43 CHAPTER 4: REACTOR[I01/2012CHAPTER 4: REACTOR 01/2012Critical Heat Flux Ratio046.0041.0036.0031.0026.00 -21.0016.0011.006.001.00 41.50III4.00 6.50 9.00 11.50 14.00 16.50 19.00 21.50Unit Cell Power (kW)Figure 4.23, Critical Heat Flux Ratio (Bernath and Biasi Correlations)Thermal hydraulic analysis using TRACE (section 4.6) demonstrates that a TRIGA fuel elementoperating at about 45 kW has a minimum critical heat flux ratio of 5.9 at a location about 86.7%of the distance of the heated length (38.1 cm) of the fuel. For a core of N fuel elements, thefuel element that produces the most power (PPEAKROD) is related to the core average power level(PAVE) by:PFi.AKROI) = PAVh.. KPFParametric variations including peaking factors from 1.3 to 2.0 and the number of fuel elementsfrom 85 to 100 are provided in Table 4.19 and Fig. 4.23. With a peaking factor of 2 and 85 fuelelements, a core at 1913 kW would produce 45 kW in the element producing the highestpower.Table 4.19, Core Power, 45 kW Hot ElementPeaking 85 90 100Factor1.3 2942 3115 34621.4 2732 2893 32141.5 2550 2700 30001.6 2391 2531 28131.7 2250 2382 26471.8 2125 2250 25001.9 2013 2132 23682 1913 2025 2250Page 4-44 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 40ETI01/2012Limiting Core Configuration(Fuel Element Peak Power 45 kW)34003150290)26500C-C) 24000U.. 21504 -..-,-+~190c) 41.3I1-37 1-44 1.51 1.58 1.65 1.72 179 1.86 1-93Peaking Factor-85 ELEMENTS -g 90ELEMENTS 1--IOELEMENTS2Figure 4.24, Core Power, 45 kW Hot ElementBased on the calculations, 85 fuel elements with a peaking factor of less than 2.0 provides alarge margin to thermal hydraulic limits.4.6 Core ReactivityAs noted in 4.5.1 (A), reactivity worth of material in the core is determined from differentialmeasurements of calibrated control rod worth positions. Verification that the coreconfiguration meets operating limits is similarly determined from the calibrated control rodpositions.As shown in Apoendix 4.1, the rapid fuel temperature response from a pulsed reactivityaddition terminates the power increase and causes the reactor to stabilize at a power levelcorresponding to the fuel temperature consistent with Fig. 4.21. Therefore limits on reactivityare based not on the peak pulse power level, but on the final equilibrium power levelassociated with the reactivity. A polynomial equation calculating the reactivity deficit based onFig. 4.24 with an R2 value of 0.99999 is: ý6k = -1.75340-"2P4 + 6.06670-10-9"P3-8.777401 0-6"P2 +8.45380-10-3"P- 0.072937An approximation of the power coefficient of reactivity from 100 kW to 1 MW is therefore:d6k = -7.01360-2.P3 + 1.82001-10--P2-1 .755488.10-6_P +8.45380-1 0-2dPPage 4-45 CHAPTER 4: REACTOR 01/2012Power Coefficient of Reactivity0.00S -- -....-OJA00 -.1W0 200 3W 400 5WD f00 700 am0 9W0 1000Power Level NkW)Figure 4.25, Power Coefficient of ReactivityTherefore a pulse rod worth limited to 2.8% Ak/k ($4.00) will prevent exceeding steady statepower level of 1.1 MW following a pulse using the total reactivity worth of the rod.A limit on pulsed reactivity addition of 2.8% Ak/k ($4.00) provides an adequate safety margin.Limiting the total experiment worth to 2.1% Ak/k ($3.00) provides additional safety margin inthe event of an inadvertent pulse from the removal of all experiments.Limiting an individual experiment to 1.75% Ak/k ($2.50) ensures that an inadvertent pulseoccurring from removal of the experiment at full power operations does not exceed limits.Limiting moveable experiments to less than 0.7% Ak/k ($1.00) will prevent an inadvertentpulsed reactivity addition leading to prompt critical condition.There appears to be a significant difference in response in the power level coefficientcomparing low power level data to high power level data; the prediction of the powercoefficient of reactivity beyond the range of 1000 kW using a simple polynomial fit is notsupported. Operating limits on core reactivity are provided in Table 4.20.Table 4.20, Reactivity Limits% Ak/k $Excess reactivity 4.9 7.00Shutdown margin[11 0.2 0.182Moveable experiment worth 0.7 1.00Single experiment worth 1.75 2.50Total experiment worth 2.10 3.00NOTE [1): most reactive rod fully withdrawn, moveable experiments in themost positive-reactive statePage 4-46 THE UNIVERSITY OF TEXAS TRIGA 11 RESEARCH REACTOR

• f 01/2012SAFETY ANALYSIS REPORT, CHAPTER 4 IBased on control rod worth values noted in Table 4.13 and calibration data from June 29, 2011,the ability of the control rods to meet the specified limits is demonstrated in Table 4.21. Whensignificant changes to the core configuration are made, verification that the core meetsrequirements is accomplished including evaluation that the control rod calibration is valid or re-establishing the control rod worth calibration.Table 4.21, Limiting Core reactivityReference Current (2011)Control RodPosition Worth Position WorthTransient rod C ring $3.00 C-1 $3.10Regulating rod C ring $3.71 C-7 $2.82Shim 1 D ring $2.86 D-14 $2.52Shim 2 D ring $2.86 D-6 $3.07Total Rod Worth $12.43 $11.51Critical Reactivity $5.43 $5.95LIMITING CURRENTExcess Reactivity $7.00 $5.56Shutdown Margin -$1.72 -$2.854.7 Thermal Hydraulic DesignThis section provides an independent assessment of the expected fuel and cladding thermalconditions, both steady-state and pulse-mode operations, with realiktic modeling of the fuel-cladding gap. Analysis is based on limiting conditions applied to 'a single fuel channel. Thecorrelation of the limiting channel to core average power is sued to validate maximumpermissible power level.Analysis of pulsed-mode behavior is provided in Appendix 4.1, a commonly cited analysis ofTRIGA fuel and cladding temperatures associated with pulsing operations. Analysis shows filmboiling is not expected, even during~or, after pulsing leading adiabatic fueltemperatures. The analysis addresses the case of a fuel element at-an-average temperatureimmediately following a pulse, then estimates cladding temperature and surface heat flux as afunction of time after the pulse. The analysis predicts that, if there is no gap resistancebetween cladding and fuel, film boiling can occur very shortly after a pulse and claddingtemperature can reach reaching 470'C. Mechanical stress to the cladding well is below theultimate tensile strength of the stainless steel at these temperatures. Through comparisonswith experimental results, the analysis concludes that an effective gap resistance of 450 Btu hr-1ft-2 OF- (2550 W m-2 K-) is representative of standard TRIGA fuel and, with that gap resistance,film boiling is not expected.Page 4-47 CHAPTER 4: REACTOR 01/2012Analysis of steady state conditions reveals maximum heat fluxes remain well below the criticalheat flux associated with departure from nucleate boiling. The heat transfer model isdiscussed, followed by the'results. .4.7.1 Heat Transfer ModelHeat generated in the fuel is conducted through the fuel matrix, transferred by convectionacross the gap between the fuel matrix and the cladding, conducted through the fuel cladding,and transferred by convection to the cooling water that flows through the core. Fuel centerlinetemperature can be calculated as the cooling water temperature increased by temperaturechanges through each physical element from the centerline of the fuel rod to the watercoolant.T,= T11 + ATb,+ AT + ATg + A T,Where7 , is the fuel centerline temperatureTh is the bulk water temperatureAIT,r is the difference in temperature between bulk water and fuel claddingAT, is the difference in temperature acrossATg is the difference in temperature across the gap betweenthe fuel andthe claddingAT,,, is the difference in temperature across the radius of the fuelA standard heat resistance model for this system is:h " k c r, 4 zh , 2 .k fWhereq" is the heat flux through the cladding surfaceh is-the convective heat transfer coefficient associated with the cooling waterro and ri are cladding inner and outer radiikc is the cladding thermal conductivityhg is the gap conductivitykf is the fuel thermal conductivityThermodynamic values are provided in Table 4.22, with the exception of the convective heattransfer coefficient associated with the cooling water. The gap conductivity of 2840 W m2 K-1(500 Btu h-' ft -2 F-') is taken from Appendix A. General Atomics reports that fuel conductivityPage 4-48 0THE UNIVERSITY OF TEXAS TRIGA 11 RESEARCH REACTOR

• fl 1 01/2012SAFETY ANALYSIS REPORT, CHAPTER 4 1over the range of interest has little temperature dependence. Cladding conductivity across thecladding is temperature dependent, with values quoted at 300 K, 400 1( and 600 K.Table 4.22: Thermodynamic ValuesParameter Symbol Value UnitsFuel conductivity kf 18 K m- __14.9 W m 'K (300 K)Clad conductivity kg 16.6 W m-4K' (400 K)19.8 W m1 Kz (600 K)Gap resistance h9 2840 W m-2 K-1Clad outer radius r0 0.018161 mFuel outer radius ri 0.018669 mActive fuel length Lf 0.381 mAxial peaking factor APF 71/2 N/AThe convective heat transfer coefficient is mode dependent and can only be determined incontext; TRACE was used to provide heat transfer coefficient data supporting the analysis. TheTRAC/RELAP Advanced Computational Engine (TRACE) code is the latest in a series of advanced,best-estimate reactor systems codes developed by the U.S. Nuclear Regulatory Commission foranalyzing transient and steady-state neutronic-thermal-hydraulic behavior in light waterreactors. It is the product of a long term effort to combine the capabilities of the NRC's fourmain systems codes (TRAC-P, TRAC-B, RELAP5 and RAMONA) into one modernizedcomputational tool.27The TRACE calculation models a unit cell composed of the area enclosed within a geometry unitdefined by fuel pitch. Flow through the unit cell is modeled as a pipe, with model elementsrepresented in Fig. 4.25. The UT TRIGA unit cell is an equilateral triangle, based on hexagonalgeometry. Three 30' segments of a fuel element fall within the unit cell, with calculations forheat generation corresponding to 1/2 of the element. For example, calculations assuming 10 kWfor the unit cell give indication of thermal response to an element output of 20 kW. The sectionof the fuel element that contains the fuel matrix (heated length) is modeled separately fromthe unheated lengths.The active length of the fuel element was modeled as a TRACE heat source 15 in. (38.1 cm) inlength, with the heat exiting through stainless steel cladding. Heat distribution was modeledas sinusoidal variation from a maximum at the center to a minimum modified at the end byextrapolation length of thermal neutrons in graphite. Data was calculated for 15 equallyspaced nodes across the span of the simulated fuel element (i.e., 0.0127, 0.0381, 0.0635,0.0889, 0.114, 0.140, 0.165, 0.191, 0.216, 0.241, 0.267, 0.292, 0.318, 0.343, and 0.368 m).2' https://www.nrcsnap.com/snap/-luRins/trace/index.isp 01/2012Page 4-49 CHAPTER 4: REACTOR 1 01/2012Flow entrance and exit has a more complex geometry, and is not modeled explicitly. Specialconsideration given to the thermal hydraulic characteristics of the fuel end fittings that act asan interface between the flow channel and the grid plates, and the expansion or contraction offlow as it passes into/out of the flow channel. The characteristic losses associated with theentrance and exit includes turbulence effects from sudden expansion and contraction imposedby the grid plates as well as changes in flow direction. These losses are understood in terms offractional values, or K factors.28 The analytic expression of K for expansions/contractions is:K=F1 l1L A21The flow path exiting the grid plate and entering the area below the cooling channel undergoesa 450 rotation, followed by another 450 rotation to direct flow along the cooling channel; theassociated K factors are 0.3429 for each turn. The flow area expands suddenly at the entrance,with a K factor approximately 1.0. The loss factor for the cooling channel entrance is therefore1.68. Similar calculations at the exit of the cooling channel yield a total loss factor of 1.18.IN") 1.714-Figure 4.25: Unit Cell Fuel Element ModelHydrostatic pressure is required for TRACE calculations. Pressures at the inlet and outlet to theunit cell were calculated from nominal values of pool level, differential pressure from theconfinement system HVAC, and local barometric pressure (Table 4.23). Normal pool level is 8.1m, with a minimum of 6.5 m. (required by Technical Specifications). Normal pool temperature isabout 20°C, with limiting temperature 48°C. Cooling flow enters the lower grid plate at 23.94in. (0.608 m) above the pool floor and exits the upper grid plate at 51.00 in (1.2954 m). Thereactor bay ventilation system operates at a slight vacuum (nominally 0.07 in. H20 or 17.44 Pa)reducing barometric pressure slightly. The average barometric pressure for the Austin area is28 Handbook of Hydraulics, 5 h Ed. New York: McGraw Hill, King, H.W. and Brater, E. F. (1963)29 http://www.westerndynamics.com/Download/friclossfittinzs.odf (01/2012)Page 4-50 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 4000 D 11ETLI0-001/2012reported to range from 28.88 in. to 30.09 in., with an average of the high and low range at29.485 in., or 99847.67 Pa.Table 4.23, Hydrostatic PressuresColumn Temp waterwater level height (m) (°C) density(M) (kg/m3)WaterPressure(Pa)TotalHydrostaticPressureCore top 8.1 6.805 20 998.23 66609.88 166440Core bottom 8.1 7.492 20 998.23 73338.06 173168Core top 6.5 5.205 48 988.56 50455.54 150285Core bottom 6.5 5.892 48 988.56 57118.74 1569494.7.2 ResultsThe TRACE model was used to calculate temperatures at each of the 15 nodes. The heat fluxand the temperature at the inner surface of the cladding were used to calculate the centerlinefuel temperature. The temperature data (including TRACE data and centerline fuel temperaturein is presented in Tables 4.24-4.27, and graphically (Fig. 4.26). Unit cell and total core flow wascalculated (Table 4.28). Finally, recent observations of fuel temperature from installedinstrumented fuel elements are compared to calculated fuel centerline temperatures in Table4.29.A. Water TemperatureLimiting thermal hydraulic design is based on system response with -the maximum permittedpool water temperature and the "minimum allowed pool water level. TRACE calculations wereperformed for a range of unit cell power production. Table& 4.24 provides the coolanttemperatures calculated by TRACE at each node.Table 4.24, Unit Cell Coolant Temperature (°C) for 490C 6.5 m Pool WaterUnit Ce:. Node(kW) 1 2 3 4 5 6. 7 8 91.5 , 503.0, 504.5 .506.0 517.5 519.0 ... 5110.5 5112.0 5213.5 5215.0 5216.5 5218.0 5250 5151 5352. 5452 5553 5554 5654 5755 5855 5955 5956 6056 6052. .5254, 5456 56.57., 5758 5959. 6060 6161 6262 6363 6464 6565* 66.52 53,. 53;55 55 5556'..57 57r8 58 5959 60 6060. 61 6262 62 6363 64 6464 65 6565 66 6766 67 6867 68 6953555759616264656667-697010 11 ..12 3.3 1.4, 1553 54 .55 56 57 .57,.56 57 59 60 .61- 6258,60 rV62 .64 65 6760 62 64 _64 68 7061 64 '67 67 72 7363 66 69 69 74 7764 68 71 72 77 7966 70 73 74 80 8267 71 75 77 82 8568 73 77 79 84 8769 74 79 83 86 9071 76 80 85 89 92Page 4151 CHAPTER 4: REACTOR 01/2012Table 4.24, Unit Cell Coolant Temperature (°C) for 49°C 6.5 m Pool WaterUnit Cell Node(kW) 1 2 3 4 5 6 7 8 9 10 11 12 13 14 1519.5 53 57 61 66 67 68 69 70 71 72 '77 82 87 91 9421.0. 53 57 ;61 66 67 68 69 70 71 72 78 83 87 91 9522.5 53 ,57 61 '66 67 68 69 70 71 72 78 83 87 91 95B. Fuel TemperatureTRACE calculations provide cladding temperatures directly. Given the cladding temperatures,the standard heat resistance model previously identified in 4.7.1 can be simplified to:claddnginner + q h 2where heat flux and cladding temperatures are calculated in TRACE (Table 4.25 and 4.26). Thegap and fuel physical dimensions and thermodynamic properties are constants; based on valuesin Table 4.22, the terms in parenthesis resolve to 5.39E-4 W m-2 k-1.About 6% of the coefficientis determined by the gap, and any error associated with gap conductivity is minimized.Table 4.25a, Outer Clad Temperature (°C) for 49°C and 6.5 m PoolUnit Cell NodekW 1 2 3 4 5 6 7 8 9 10 11 12 13 14 151.5 65 67 70 73 75 76 77 78 77 76 77 76 75 73 723.0 76 81 85 89 93 94 96 97 96 95 95 93 91 88 864.5 86 91 97 102 106 109 110 113 1I1 110 109 106 103 100 986.0 '94 100 -107 -114* 119 121 122 123 i22 122 121 118 114 Il 1077.5 102 I09 116 123 124 124 125 125 125 124 124 123 122 120 1169.0 109 117 123 125 125 125 126 126 126 125 125 125 124 123 12210.5 116 123 "125 '126 126 126 126 127 'i26 126 126 126 125 125 12412.0 122 125 126 126 127 127 127 127 127 127 127 126 126 125 12513.5 124 126" 126 i27 127 127 128 110 127 127 127 127 126 126 12515.0 125 126 127 127 128 128 128 128 128 128 128 127 127 126 12616.5 126 127 127 128 128 128 128 129 128 128 128 128 127 127 12618.0 127 127 128 128 129 129 129 129 129 129 128 ".128 128 127 12719.5 127 128 128 129 129 129 129 129 129 129 129 128 128 127 12721.0 127 128 128 129 129 129 130 130 129 129. 129 129 128 128 12722.5 128 128 129 129 130 130 130 130 130 130 i29 129 129 128 127Page 4-52 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 400 NETL01/2012Table 4.25b, Inner Clad Temperature (°C) for 49°C and 6.5 m PoolUnit Cell(kW)1.53.04.56.07.59.010.512.013.515.016.518.019.521.022.51667989981071151221301331351361381391401422342356368378388397404406408410411413414416417Node3 4 5 6 7 :..8 9 .10344 347 349 351 352 .353 352 351360 365 369 370 372 374 .372 371373 379 384 386 388 391 389 387385 392 397 400 401 402 402 400395 402 404 405 405 406 405 405403 405 407 407 408...408 408- 407405 408 409 410 410. 411 410 409110 410 411 412 412 413 412 411410 412 413 414 414 415 414 413412 413. 415 415 416 417 416 415413 415 417 417 418 419 418 417415 417 418 419 420 421 420 419416 418 420 421 422 422 422 421418 420 422 423 423 424 423 422419 421 423 424 425 426 425 42411 12 13 14351 350 349 348 346371 368 366 364 361386 383 380 377 374400 396 392 388 384404 403 401 398 394407 406 404 403 40140S 408 406 405 404;411 410 408 407 405413 411 410 408 407415 413 412 410 408416 415 413 411 410418 416 415 413 411420 418 416 414 412422 420 418 415 413423 421 '419 417 415Tabie 4.26a, Heat Flux (Nodes l-18) 49°C 6.5 Pool,Unit Cell Node(kW) 1 2 3 4 5 ..6 7 81.5 -2.72E4 -3.06E4 -3.40E4 -3.74E4 -4.08E4 -4.23E4 -4.38E4 -4.53E43.0 -5.44E4 -6.12E4 -6.80E4 -7.48E4 .-8.16E4 -8.46E4 -8.76E4 -9.06E44.5 -8.16E4 -9.18E4 -1.02E5 -1.12E5 -1.22E5 -1.27E5 -1.31E5 -1.36E56.0 -1.09E5 -1.22E5 -i.36E5 -1.50E5 -1.63E5 -1.69E5 -1.75E5 -1.81E57.5 -1.36E5 -1.53E5 -1.70E5 -1.87E5 -2.04E5, -2.11E5 -2.19E5 -2.27E59.0 -1.63E5 -1.84E5 -2.04E5 -2.24E5 -2.45E5 -2.54E5 -.2.63E5 -2.72E510.5 -1.90E5 -2.14E5 -2.38E5 -2.62E5 -2.86E5 -2.96E5 -3.06E5 -3.17E512.0 -2.18E5 -2,45E5 -2.72E5 -2.99E5 -3.26E5 -3,38E5 _-:3.50E5 -3.63E513.5 -2.45E5 -2,75E5 -3.06E5 -3.37E5 -3,67E5 -3.81E5 -3.94E5 -4.08E515.0 -2.72E5 -3.06E5 -3.40E5 -3.74E5 -4.08E5 -4.23E5 .-4.38E5 -4.53E516.5 -2.99E5 -3.37E5 -3.74E5 -4.11E5 -4.49E5 -4.65E5 -4.82E5 ý4.99E518.0 -3.26E5 -3.67E5 -4.08E5 -4.49E5 -4.89E5 .-5.07E5 -5.25E5 -5.44E519.5 -3.54E5 -3.98E5 -4.42E5 -4.86E5 -5.30E5 -5.50E5 -5.69E5 -5.89E521.0 -3.81E5 -4.28E5 -4.76E5 -5.23E5 -5.71E5 -5.92E5 -6.13E5 -6.34E522.5 -4.08E5: -4.59E5 -5.10E5 -5.61E5 -6.12E5 -6.34E5 -6.57E5 -6.80E5Page 4-53 CHAPTER 4: REACTORI01/2012Table 4.26b, Heat Flux (Nodes 8-15) 49°C 6.5 Pool,Unit Cell Node(kW)I 9 10 11 12 13 14 151.5 -4.38E4 -4.23E4 -4.08E4 -3.74E4 -3.40E4 -3.06E4 -2.72E43.0 -8.76E4 -8.46E4 -8.16E4 -7.48E4 -6.80E4 -6.12E4 -5.44E44.5 -1.31E5 -1.27E5 -1.22E5 -1.12E5 -1.02E5 -9.18E4 -8.16E46.0 -1.75E5 -1.69E5 -1.63E5 -1.50E5 -1.36E5 -1.22E5 -1.09E57.5 -2.19E5 -2.11E5 -2.04E5 -1.87E5 -1.70E5 -1.53E5 -1.36E59.0 -2.63E5 -2.54E5 -2.45E5 -2.24E5 -2.04E5 -1.84E5 -1.63E510.5 -3.06E5 -2.96E5 -2.86E5 -2.62E5 -2.38E5 -2.14E5 -1.90E512.0 -3.50E5 -3.38E5 -3.26E5 -2.99E5 -2.72E5 -2.45E5 -2.18E513.5 -3.94E5 -3.81E5 -3.67E5 -3.37E5 -3.06E5 -2.75E5 -2.45E515.0 -4.38E5 -4.23E5 -4.08E5 -3.74E5 -3.40E5 -3.06E5 -2.72E516.5 -4.82E5 -4.65E5 -4.49E5 -4.11E5 -3.74E5 -3.37E5 -2.99E518.0 -5.25E5 -5.07E5 -4.89E5 -4.49E5 -4.08E5 -3.67E5 -3.26E519.5 -5.69E5 -5.50E5 -5.30E5 -4.86E5 -4.42E5 -3.98E5 -3.54E521.0 -6.13E5 -5.92E5 -5.71E5 -5.23E5 -4.76E5 -4.28E5 -3.81E522.5 -6.57E5 -6.34E5 -6.12E5 -5.61E5 -5.10E5 -4.59E5 -4.08E5Calculation of maximum fuel temperature was performed using the standard heat resistancemodel (Table 4.27) modified to use the TRACE data as described above.Table 4.27, Fuel Centerline Temperatures (°C)Unit Cell Node(kW) 1 2 3 4 5 6 7 8 9 10 11 12 13 14 151.5 90 95 101 107 112 115 117 119 117 115 114 110 106 102 973.0 126 137 147 158 167 172 176 180 176 173 170 161 153 145 1364.5 161 175 190 205 218 225 231 237 231 226 221 209 197 185 1726.0 194 212 232 251 268 276 283 289 283 276 270 255 239 222 2077.5 226 249 272 294 311 318 325 333 325 318 311 295 278 260 2419.0 258 286 310 330 350 358 367 375 367 358 350 330 311 292 27210.5 290 319 343 366 388 398 407 417 407 397 388 366 343 321 29812.0 322 349 375 400 425 437 448 460 448 436 425 400 375 350 32513.5 349 378 407 436 464 477 489 502 489 476 463 436 407 378 35015.0 375 407 438 470 502 515 529 S43 529 515 501 470 438 407 37516.5 400 436 470 504 540 554 570 586 570 554 539 504 470 436 40018.0 425 464 502 540 577 593 610 627 610 593 576 539 501 463 42519.5 451 492 533 574 615 633 650 669 650 633 614 574 533 492 45121.0 476 520 565 608 652 672 691 710 691 672 652 608 564 520 47622.5 502 549 596 643 690 711 732 753 732 710 690 643 596 549 501Page 4-54 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 4MtI01/2012C. Temperature Profiles and Flow RatesTRACE calculation provides thermal response for a unit cell. Temperature calculations arebased on heat flux (heat per area) and are therefore valid for both the unit cell and the fuelelement. However, Fig. 4.25 shows a unit cell to be Y2 of a fuel rod so that the total powergenerated in a fuel element is twice the unit cell value.Fuel, cladding, and coolant temperatures based on TRACE data in Tables 4.24, 4.25, and 4.27are provided in Fig. 4.26 for two unit cell power levels, 10.5 and 22.5 kW. Flow rate versuspower for a single fuel element is provided in Fig. 4.25.Total core flow is the sum of the flow rates of individual fuel rods operating at specific powerlevels. The power level an element generates is determined by the peaking factors associatedthe position in the core. Instrumented fuel elements (modified to accept thermocouples) haveslightly less fuel mass than other standard fuel elements. Fuel followers have a smaller radiusthan standard fuel elements. Therefore the power production in thermocouple elements andfuel followers is less, approximately by the ratio of the mass of the element to the mass of astandard fuel element. Table 4.28 provides the data and calculation of total core flow based ona 116 element core operating at 1100 kW. Similar calculations were performed for a 120, 100and 85 element core over a range of power level with results in Fig. 4.27.10.5 kW Unit Cell Axial Temperature Profile675625I575 --%M4-525475 -7425 -- ------45...... ...............................---------------------------375325 .-.. .72751 3 511 13 15Node--Fuel Temp --- Inner Clad Temp -Outer Clad Temp --Coolant TempFigure 4.26a, Unit Cell Temperature Distribution (10.5 kW)Page 4-55 CHAPTER 4: REACTORI01/2012CHAPTER 4: REACTOR 01/201222.5 kW Unit Cell Axial Temperature ProfileE99758757756755754753752751 3 5 7 9 11 13 15-Fuel Temp .Inner Cad Temp -CouterClad Temp --CoacantTemnpNodeFigure 4.26, Unit Cell Temperature DistributionFlow rate versus Unit Cell Power0.130-120.110.100090.080.070.360.050.0]4 -i ; .-- --. ... .... .0~3 yl 1.02541EO5x5 -4.390E-O4xz + I.412BE-O3x+3.0123E4020.02 .992S-01"-0.000 2 4 6 8 10 12 14 16 18 20 22 24Unit Cd power (W,Figure 4.27, Single Rod Flow Cooling Flow Rate versus Power Level 49°C 6.5 PoolPage 4-56 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 40 0I01/2012Table 4.28, Coolant Flow for 1100 kW Operationno. peaking Power (kW) Flow Rate (kg/sring elements factor Unit Cell Element RING Unit Cell Element RINGB (6) 4 1.57 7.44 14.89 59.55 0.07 0.15 0.58B IFE 2 1.57 7.31 14.61 29.23 0.07 0.14 0.29C (12) 10 1.46 6.92 13.84 138.45 0.07 0.14 1.41C FFCR 2 1.46 5.78 11.56 23.11 0.07 0.13 0.26D (18) 16 1.29 6.12 12.23 195.72 0.07 0.14 2.16D FFCR 2 1.29 5.11 10.21 20.42 0.06 0.13 0.25E (24) 24 1.07 5.07 10.15 243.52 0.06 0.13 3.02F (30) 28 0.81 3.84 7.68 215.07 0.06 0.11 3.17G (30) 28 0.66 3.13 6.26 175.24 0.05 0.10 2.94TOTAL: 116 TOTAL: 1100 TOTAL: 14.08D. Comparison to Operational dataDuring calendar year 2011 the UT TRIGA core consisted of 109 standard fuel elements, 2instrumented fuel elements and 3 fuel followers. Operational data was collected (Table 4.29)to compare calculated fuel temperature values for specific operations with observedindications. Each of the selected values follows an operating interval that approaches steadystate fuel temperatures, except for *the 10 kW values (there was a series of 10 kW operationsfor less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> during 2011). The power produced by the individual fuel element iscalculated as total core power divided by the number of fuel elements (114) and multiplied bythe nominal B ring peaking factor (rt/2).Instrumented fuel elements contain three thermocouples, one at the axial ',midplane with theremaining thermocouples offset 1 inch above and below. Only one thermocouple in each IFE isin use. Fig. 4.24 shows approximately'20 'C difference between the center of the element andpositions approximately 1 in. from the center at 10.5 kW unit 6ell power. The, core power is notexpected to be homogenous in the B ring; consequently the peaking factors not expected tobe uniform for all B ring elements. The position of an IFE and the position of an individualthermocouple within IFE may affect temperature indication; the agreement between FT1 andFT2 measuring channels is therefore considered remarkably good.Table 4.29, Observed Fuei TemperaturesDate Power (kW) Observed Temperatures (°C)Core B Ring IFE FT1 FT2 Pool10/6/2011 10 0.13 26 28 21.112/21/2011 100 1.31 86 97 22.812/20/2011 500 6.54 240 261 2312/16/2011 950 12.44 340 359 21.9Page 4-57 CHAPTER 4: REACTORI01/2012TRACE calculations were performed at power levels corresponding to 100, 500 and 950 kW withpool water temperatures corresponding to the values recorded in Table 4.29. The maximumfuel temperatures from the TRACE calculation are provided in Table 4.30 along with theobserved values for comparison.Table 4.30, Fuel Temperature ComparisonPower (kW) Fuel Temperatures (°C)Core B Ring IFE TRACE FT1 FT2.100 1.3 95 86 97500 6.5 289 240 261950 12.4 446 340 359Fuel Element Temperature Versus Power-'-CALCJLATED -o-FT I VALUES -FT 2 VALUES500EaEtU.-4504003503O0250200150100...... 4-41508 9 10 11 1200 1 2 3 4 5 6 7Fuel Element Power (kV13NI)Figure 4.28, Comparison of Calculated and Observed Fuel TemperaturesThe information (provided graphically in Fig. 4.28) shows the temperature from TRACEcalculations to be reasonably close and consistently higher than observed values, with thedeviation increasing as power level increases. Therefore, modeling bounds actual conditions.Page 4-58 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 00o0 I 01/2012SAFETY ANALYSIS REPORT, CHAPTER 9 ..09.0 AUXILIARY SYSTEMS9.1 Confinement SystemThe design of a structure to contain the TRIGA reactor depends on the protection requirementsfor the fuel elements and the control of exposures to radioactive materials. Fuel elements andother special nuclear materials are protected by physical confinement and surveillance.The floor of the reactor bay is approximately The lower walls of thereactor bay are cast in place concrete. Above grade, the walls are reinforced precast concretetilt panels, approximately with integral columns and embeddedreinforcing steel. The wall panels were then set in place vertically using a crane with space leftin between each panel for a structural column and temporarily braced. Next the column formswere placed around reinforcing steel extending from the edges of the panels which wasinterlaced with additional steel reinforcing internal 'to the columns. Concrete was then pouredinto these forms resulting in a finished wall system with columns that resemble a poured inplace design rather than the typical tilt panel welded design. The roof is sealed usingstandard tar and gravel techniques. All penetrations in the reactor bay confinement envelopeare on the south side, interfacint with the reactor wing offices, machine room spaces,equipment staging area, and confinement (and auxiliary purge) ventilation system..'.9.2 HVAC (Normal Operations)Building environment controls use air handling units for ventilation and comfort with cold andhot water coils for temperature and humidity control. There are two separate HVAC systemswith three air handling units, located on the fourth level of the reactor bay wing adjacent to thereactor bay. One unit contains both cold and hot water coils in a single duct.system, dedicatedto the reactor bay. This system supports confinement functions. The other two units are thecold- and hot-deck components of a double duct system that conditions air in all building zonesother than: the reactor bay. A fume/sorting hood is installed in the reactor bay, using aseparate exhaust fan and isolation damper that disc"h'arges into a separate roof stack.Water temperatures of the heating and cooling coils in the air handling units are controlled byset of on-site and off-site systems. The heating system is an ori-site boiler unit with a designcapacity set by local building (HVAC) requirements. The cooling system is a PRC chilled watertreatment plant with design ýcapacity Set by overall research campus requirements; withthermostats controlling zone or room temperatures. A local instrument air system providescontrol air for HVAC systems. Controls and air balancing of the two air handling systemsprovide user comfort and pressure differentials between the reactor bay (confinement) andadjacent zones, and between the adjacent zones and the academic wing of the building.Page 9-1 CHAPTER 9, AUXILIARY SYSTEMS 1 01/2012The ventilation system is designed to maintain a series of negative pressure gradients withrespect to the building exterior and other building areas, with the reactor bay (confinement) atthe lowest pressure. Confinement functions of ventilation control the buildup of radioactivematerials generated as a byproduct of reactor operations, and isolate the reactor bay in theevent that an abnormal 'release is'detected in the reactor areas. Confinement and isolation isachieved by air control;dampers and leakage prevention material at doors and other roompenetration points.A conceptual diagram of the system is provided in Fig. 9.1. Manual operation controls for bothmain and purge air systems are in the reactor control room.wowFigure 9.1, Conceptual Diagram of the Reactor Bay HVAC SystemAn exhaust stack on the roof combines the ventilation exhausts from both the main and thepurge systems. As illustratedJin Fig. 9.1, the auxiliary purge system discharge is within the HVACexhaust stack.. The auxiliarypurge exhaust is a 6 in. (15.24 cm) internal ID and 8.63 in. (21.92cm) OD. The HVAC exhaust has an 18 in. (45.72 cm).9.2.1 Design basisConfinement system ventilation has three modes of operation, reactor run mode, quiescentmode, and confinement isolation. The design goal for HVAC system in the reactor run mode isto control the reactor bay, adjacent zones and academic wing of the building at a negativepressure difference relative to ambient atmospheric pressure during routine operations. Thedifferential pressures are 0.06: 0.04: 0.03 in. water (0.15: 0.10: 0.80 cm of water). This pressuregradient assures that any radioactive material released during routine operations is dischargedthrough the stack and does not build up in the reactor bay. Release of airborne radioactivityconsists mostly of activated 41Ar from routine operation. The design goal of the confinementPage 9-2 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 900 B0if01/2012system ventilation during quiescent mode is to minim; e energy utilization during periods whenthe reactor is not operated while maintaining pressure in the reactor bay 0.06 in. belowatmospheric pressure. The reactor-room confinement is 4esigned to control the exposure ofoperation personnel and the public from radioact.ive material or its release caused by reactoroperation. During potential accident conditions, sensors initiate confinement system isolationwhen high levels of radioactivity are detected in reactor bay air, e.g. If a fuel element. failurereleases fission products or if an experiment with sufficient inventory of radioactive materialfails. The confinement isolation secures fans and dampers in the. confinement HVAC,fume/sorting hood, and auxiliary purge 1system. Provisions are made to allow subsequentoperation of the auxiliary purge system with the remaining HVAC confinement in isolation.Release criterion is based on Title 10 Chapter 20 of the U.S. Code of Federal Regulations.9.2.2 System descriptionDuring operating modes supply fans draw air from either the return fan or the environmentinto a conditioning unit that subcools the air:,to control humidity then heats the air forhabitability/comfort. Air filtration is the typical design for normal. HVAC operation withfiberglass roughing filters only. The ccnfinement systemn uses heating and cooling in a singleunit, the remainder of the building HVAC system has air conditioning split into separate hot andcold decks.COMM-~SLPPLYAIR-~ ~~ FILTER a eJ~EA1 ~ 1.AiISCMfICFigire 9.2A, Ma'n Reactor Bay HVAC SystemPage 9-3 CHAPTER 9, AUXILIARY SYSTEMS 01/2012Table 9.1, Typical Confinement Vent & Purge ParametersDuct Velocity Exit VelocityAux Purge 3900 fpm 20 m/s 35.23 m/sConfinement Vent 1800 fpm 9 m/s 26.87 m/sFlow RateAux Purge 1100 cfm 0.52 m3/sConfinement Vent 7200 cfm 3.40 m3/sPAMAIRWOAMIUET uq.Y S3m IA"M ISMATM(P. LWIII~f. HVAC OPERAT I ON all R.P. MATISO- -W 6HI. FNG CMD'A= AM FZCE=*am AM plýcadvsFiue92,Mi- RatrByHA Contro SyteCntoFMOFfify fro" W=M .U~IFigure 9.21B, Main Reactor Bay HVAC Control System Control9.2.3 Operational analysis and safety functionSpeed of the confinement system supply fan is regulated to produce 0.06 in. water vacuum inthe reactor bay by differential pressure control between the reactor bay and a representativeambient external building measurement point. Additional measurement points in ventilationzones adjacent to the reactor bay are used to maintain differential pressure between thereactor bay and adjacent access areas. Supply air is distributed through a rectangular duct nearthe ceiling and then to distributed ducts and vents running down the wall and ending near thefloor), enhancing mixing and preventing stratification. Air is discharged from the bay through 4return grills, two parallel ducts to grills near the floor, and two grills near the ceiling. In thereactor run mode the confinement system exhaust fan is controlled to maintain stack velocitydesigned to exceed the minimum air change specification. Control dampers are located at thesupply fan inlet (fresh air intake) and the exhaust fan outlet (discharge to stack), and in a linebetween the supply and return fans. Confinement system ventilation discharge is through aPage 9-4 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 9iSso e REI'tI01/2012stack on the reactor building roof. Schematics of the ventilation system for the reactor bayarea and a logic diagram of the ventilation control system sensors and controls are provided inFig. 9.2A and B.Table 9.2, Reactor Venti'Eation System ModesMODESYTMCMOETREACTOR RUN QUIESCENT ISOLATIONcontrol damper CLOSED OPEN CLOSEDsupply & exhaust OPEN MINIMUM MINIMUMcontrol dampersConfinement HVAC Controlled for StackSupply Fan velocity Constant Speed OFFExhaust fan Controlled for bay Controlled for bay OFFdp dpsupply & exhaust OPEN CLOSED(i CLOSEDcontrol dampers OPEN[21Auxiliary PurgeSystem OFFExhaust fan ON OFF[1]OFF or ON121supply & exhaust OPEN or CLOSED'31 -CLOSED"' CLOSEDFume/Sorting Hood control dampersExhaust fan ON or OFF[31' OFF111 OFFNOTE [1]: Mode is set manuallyNOTE [2]: Provisions have been made to permit operation of auxiliary purge system in conferment isolationNOTE [3]: Fume hood is operated manually, as required, and not correlated to reactor operationWhen the reactor is operating (reactor run mode) the system is operated to generate a rate ofair exchange exceeding 2 air volumes (4120 M3) per hour. maintain a controlled stack velocity,and regulate negative pressure in the reactor bay. In the reactor run mode, the confinementHVAC supply fan is controlled to maintain the reactor bay at nominal minimum 0.06 in. water.In the quiescent mode, the confinement ventilation system is balanced for recirculation flowwith a small amount of effluent. When the reactor is not operating (quiescent mode), theventilation system is operated to minimize requirements for conditioning incoming air, in arecirculation mode with a minimal exhaust flow rate and fresh air intake as required tomaintain a negative pressure in the reactor bay with respect to adjoining spaces.In the confinement isolation mode the confinement HVAC and the reactor bay fume/sortinghood are secured; the auxiliary purge system is secured when isolation occurs, but may bemanually configured to operate. In the event that airborne radioactive material exceeding atrip set point is detected, the system is designed to establish a shutdown and isolatedPage 9-5 CHAPTER 9, AUXIUARY SYSTEMS 01/2012condition. Separate controls allow the confinement HVAC and the reactor bay fume/sortinghood to be isolated while the auxiliary purge system can be operated.Atmospheric dispersion' using a'st4ck, model reqtires:stack discharge 60 (18.23 m) feet abovethe ground, and at least :2 and 1A times the height of adjacent structures. The nearest structureismapproximately 80 meter's from the" reactor bay'. Ground elevation in the area is 794 feet, withroof elevation at the stack 843 feet, a distance of 49 feet (14.94 m) above grade. The exhauststack extends:14 feet (4.24 ýmeters) above the roof level so that the stack discharge is 63 feet(19.202inm). The ýeffective release point above the exhaust'stack can be calculated from theBryan -Davidson *equation: -1.4(VS.Where:Ah is the height of plume rise above release point (m)iD is the diameter of stack (m), confinement vent 0.40122m2, auxiliary purge 0.152 m2/7 is the mean wind speed at stack heght (m/s)V, is the effluent vertical eff!ux velocity (m/s), confinement vent 26.87 m/s, purge 35.23 m/sThe effective stack height for the reactor HVAC confinement vent system (in units of meters) istherefore 40.19/{wind velocityl. rn above the stack, and the effective stack height for theauxiliary purge system is 22.25/{wind velocityl above the top of the:stack at 63 feet (19.202 m).Mixing of the two effluent streams occurs at the exit of the stack.Pneumatically operated isolation dampers in the confinement system ventilation are located atthe supply fan outlet (supply-to 'he reactor bay) and the exhaust fan inlets (return from thereactor bay) near the reactor bay wall penetrations as indicated in Fig. 9.1, as* weli as thefume/soring hood in the reactor bay auxiliary purge system. Controls close the dampers andsecurethe fans in response to manual or automatic signal initiated by high airborne particulateradioactivity. Loss of instrument air or loss of control power will cause the'dampers to isolatethe reactor bay.9.2.4 Instruments and controlsAs indicated, the HAVC control, system is controlled' by a, set of temperature, flow, anddifferential pressure sensors that develop control signals. The signals are used in variablefrequency controllers that regulate fan speed to maintain pressure and temperature.Control room switches establish the operating mode of the confinement ventilation system.The auxiliary purge system is controlled from the same panel.Page 9-6 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR OF, 01/2012SAFETY ANALYSIS REPORT, CHAPTER 9 e MtFigure 9.3, Confinement System Ventilation ControlsConfinement HVAC mode is controlled by a toggle switch labeled "Reactor Off/ Reactor On."Reactor Off establishes quiescent mode described above. Reactor On establishes the reactorrun mode described above. Alarm indicators on tie control panel provide indication that the diffarential pressures arenormal or abnormal. Flow and differential pressure indicators inside the panel provideindication of the zone static pressure, and confinement system and auxiliary purge systemvelocities.A continuous air particulate detector located in the reactor bay provides a control signal toinitiate confinement isolation when the count rate exceeds a preset level. Indicators at thereactor control console provide alarm level information. A count rate associated with 2,000pCi/ml detects particulate activity at occupational levels of 1OCFR20. The alarm setpointexceeds the occupational values for any single fission product radionuclide in the ranges of 84-105 and 129-149. Seventy per cent of the particulate radionuclides are also detectable at thereference concentrations within two hours.Page 9-7 CHAPTER 9, AUXILIARY SYSTEMS, 01/20129.2.5 Technical Specifications, bases, testing and surveillancesEither the confinement ventilation system or the auxiliary purge system is required to beoperating when the reactor is operating to control the buildup of gaseous radioactive materialin the reactor bay. Ifthe, confinement, ventilation system is operating, instrumentation toinitiate confinement .isolation on high airborne contamination levels will be operable. Theconfinement system.will be: checked periodically to assure proper function. The particulatemonitor will be calibrated periodically.9.3. Auxiliary Purge SystemA separate, low volume air purge system is designed to exhaust' air that may containradionuclide products from strategic locations in the reactor bay.'9.3.1 Design basisThe purge system collects and exhausts air from potential sources of neutron activation suc h asbeam tubes, sample transfer systems, rotary specimen rack, and material evolving from thesurface of the pool. The purge system filters air in the system. through a rough prefiltersfollowed by a high efficiency particulate: filter. .;Design provisions allow for the addition ofcharcoal filters if experiment conditions or.other' situations should require the additionalprotection.. -- .I9.3.2 System description:Mai AIRT " ' ...' ..LP;i ' " ' ". avr z ,ARII PLO WI- II 93---1 FILTER -Li- q ....ARW RPUE 2 02,.97;1 KEA FILTERE)O4AL5T F74"4 Eg 7.3J67ACX 3 FUT1JM OWJAL FILTER CA I-fO, .MIN. *P. ILzmi........ '--"-' -" -- -0 I~.ATrI Tap~iISO" ' N I , .. W ... I. SIM e "Z LVALVE 2 4D--AIXIM 'AIR ONREACTORBSAYH_ IFigure 9.4A, Purge Air System Figure 9.4B, Purge Air ControlsPage 9-8 THE UNIVERSITY OF TEXAS TRIGA 11 RESEARCH REACTOR 0o0o 00 01/201200SAFETY ANALYSIS REPORT, CHAPTER 9 09.3.3 Operational Analysis and Safety FunctionThe primary nuclide of interest is argon-41. Fig. 9.4A and .9.4B are schematics of the auxiliarypurge system and its control logic.. Sample ports in the turbulent flow stream of the purgesystem exhaust .provide for measurement of exhaust activities. The isolationi damper in thepurge system is actuated manually, using the.fan control switch. Autornatic iso!atior, o1 thesystem is generated by the same particulate radiation mcnitcr as is used by.the HVACconfinement ventilation system.Purge flow is nominally adjusted for continuous operations with approximately 525 cfm fromthe pool and a similar dilution flow rate from the reactor bay environment. The dilution flowcontrols effluent humidity from the reactor pool area to limit possible degradation of the purgesystem HEPA filters. A purge flow of approximately 4 cfm is drawn from the beam port interiorwhen a beam port is used. The beam guide prevents closure of the outer shutter door, andbeam port three is normally purged. The rotary specimen rack is purged prior to loading orunloading for about 10 minutes to control personnel exposure and also to remove hydrogenthat may evolve from the polyethylene capsules during irradiation. -The auxiliary purge system may be operated with the confirement HVAC system secured. Sincethe confinement HVAC operates continucusly except during isolation, confinement HVAC canbe secured using thei HVAC Contro!, toggle switch (inside the HVAC control panel, describedpreviously). Since the auxiliary purge system is equipped with HEPA filters and :has thecapability for using charcoal filters, operation of the auxiliary purge system could reduceelevated airborne radionuclide contamination in the reactor bay and contain a large fraction ofthe radionuclides in filtration.. Qperation in this mode requires that the confinement HVAC besecured to prevent unfiltered releases, and then bypassing the confinement isolation tripsignal.9.3.4 Instruments and controlsThe auxiliary purge system is :controlled from the same panel as the confinement ventilationsystem. Toggle switches on the controlroormiconfinement HVAC cdhtr6lp3nbl. open dampersto allow the pool surface purge flow, 'and-independently flow from a manual valve manifoldaccessible on the ground level of the rector bay. The manual valve manifold controls purgeflow from the experiment facilities. A separate manually operated: valve' in the same areacontrols the amount of dilution flow to the purge system.A flow gage indicates purge stack velocity at the panel. The exhaust point is concentric to thecenter of the HVAC confinement ventilation exhaust stack.The auxiliary purge system is monitored by a gaseous effluent radiation detector. The effluentmonitor has an alarm setpoint based on ten times the occupational limit or a referenceconcentration at the ground.Page 9-9 CHAPTER 9, AUXILIARY SYSTEMS9.3.5 Technical Specifications, bases, testing and surveillancesIf the auxiliary purge system is operating, a gaseous effluent monitor will be operating. Theauxiliary purge system will have a high efficiency particulate filter. Auxiliary air purge systemvalve alignment will be checked periodically. The gaseous effluent monitor will be calibratedperiodically.9.4 Fuel storage and handlingSpecial provisions are necessary for the storage of fuel elements that are not in the coreassembly. The design of fuel storage systems requires consideration of the geometry, cooling,shielding, and the ability to account for each of the fuel elements. These storage systems arespecially designed racks inside the reactor pool and outside the reactor shield.Irradiated fuel is manipulated remotely, using a standard TRIGA fuel tool. Irradiated fuel istransferred out of the pool using a transfer cask modeled on the BMI cask TRIGA basket. Thereare two different loading templates for use with the transfer cask, permitting loading operationeither for a single TRIGA fuel element, or to up to three elements. A 5-ton overhead crane isused to move the fuel transfer cask.9.4.1 Design basisStored fuel elements are required to have an effective multiplication factor of less than 0.8 forall conditions of moderation. Fuel handling systems and equipment are designed to allowremote operation of irradiated fuel, thus minimizing personnel exposure.9.4.2 System description..Space inside the reactor pool is adequate for a large number of fuel racks. The racks arealuminum, suspended from the pool edge by connecting rods. Tofacilitate extra storage, 2 racks may be attached to the same connecting rods by'locating onerack at a different vertical level and offsetting the horizo0tal position.. slightly. ,Outside the react-or pool, rack design is intended to fit in, special storage wells (Fig. 9.4). Water may be added for shielding or cooling. Page 9-10 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 9'.ETII01/20121Most routine fuel storage is intended to be inside the reactor pool. Outside the reactor pool,supplemental fuel storage is planned for temporary storage of elements transferred to or fromthe facility, for isolation of fuel elements with clad damage, emergency storage of elementsfrom the reactor pool and core assembly and routine storage o" other radioactive materials.Temporary storage for some reactor components or experiments may also use the fuel storagePage 9-11 CHAPTER 9, AUXILIARY SYSTEMS 01/2012racks in the reactor pool. Other locations not in the pool will also provide storage forradioactive non-fuel materials.A fuel transfer cask modeled after the BMI cask TRIGA basket is used to transfer irradiated fuel.A standard TRIGA fuel handling tool is used to remotely grapple irradiated fuel elements.A 5-ton crane is used in conjunction with the fuel handling tool and the transfer cask to allowremote handling of irradiated fuel.9.4.3 Operational analysis and safety functionBench mark experiments conducted by TRIGA International indicate minimum mass forcriticality requires 64 fuel elements in a favorable geometry.Pool storage racks do not have the capacity or the geometry to support criticality. Spent fuelstorage has a higher fuel density in storage, but does not have the capacity to hold 64 fuelelements, and does not have favorable geometry.The fuel handling tool has been used successfully at the UT TRIGA reactor, including the originalreactor on the main campus as well as the current installation. This design is widely used byTRIGA reactors, with good performance history although the first generation tool occasionallyreleased an element if pressure was not maintained on the tool operator.The fuel transfer cask is a top loading cask, with no potential for failure or mishandling as existsin a bottom loading cask. The cask does not provide adequate shielding for close-in work, andall handling is conducted remotely.The crane exceeds load requirements for spent fuel handling by a large margin. There is littlepotential for failure.9.4.4 Instruments and controlsNew fuel storage is in a locked room on the middle level of the reactor bay. A criticalitymonitor is installed, with neutron and gamma channels. The system has a local indicatordirectly outside the storage room, and a remote readout in the control room.9.4.5 Technical Specifications, bases, testing and surveillancesFuel elements are required to be stored in a configuration with keff less than 0.8. Irradiated fuelis required to be stored in a configuration where convective cooling by water or air is adequateto maintain temperature below the safety limit.Page 9-12 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 0 l r, 01/2012SAFETY ANALYSIS REPORT, CHAPTER 9 E Tt9.5 Fire protection systemsActive fire protection elements generally have automatic operation, manual response, orpersonnel action for the intended function. Active elements to be considered include automaticfire detection, automatic fire suppression in labs and office spaces, fire informationtransmission, manual fire suppression and other manual fire control.Passive fire protection provides fire safety that does not require physical operation or personalresponse to achieve the intended function. Passive elements include inherent design features,building physical layout, safety-related systems layout, fire barriers, and construction orcomponent materials, and drainage for control of fire protection runoff water. Penetrations infire barriers have fire resistant ratings compatible with the purpose of the fire barrier.9.5.1 Design basisThe goal of fire protection is to provide reasonable assurance that safety-related systemsperform as intended and that other defined loss criteria are met1' 2. For the purpose of fireprotection, loss criteria should include protection of safety-related systems, prevention ofradioactive releases, personnel protection, minimization of property damage, and maintenanceof operation continuity. Three components shall be applied to the fire protection objective. Thethree components are passive and active fire protection, and fire prevention.A fire detection, suppression, and information management system is designed to ensure thatfires can be detected, suppressed (where possible), and alert response organizations.Basic design features of the reactor assembly, pool and shield system, and the instrumentation,control, and safety system represent passive fire protection elements. These basic features aresufficient passive protection to protect safety-related systems.9.5.2 System descriptionManual protection consists of manual firefighting actions and the systems necessary to supportthose actions such as extinguishers, pumps, valves, hoses, and the inspection, maintenance andtesting of equipment to assure reliability and proper operation. Other manual actions that areelements of active fire protection include utility control, personnel control, and evacuation.Preplanning and training by facility and emergency personnel ensures awareness of appropriateactions in fire response and possible hazards involved.'Code of Federal Regulations, Chapter 10 part 20, U.S. Government Printing Office, 1982.2 Dorsey, N.E., Properties of Ordinary Water-Substance, Reinhold Publishing Corp., New York p. 537.Page 9-13 CHAPTER 9, AUXILIARY SYSTEMS -1/2012Automatic and manual protection systems in the building include several different typesystems. In the academic wing of the building automatic protective actions are provided by asprinkler system with heat sensitive discharge nozzles, detectors for heat and smoke, andventilation systems, dampers. The reactor bay wing uses smoke detectors for areas outside thereactor bay that are radiation areas.' The reactor bay ventilation system has smoke detectorsthat provide a warning of problems within the reactor bay. Although not a strict safetyrequirement, a gaseous extinguisher system (halon) is installed to protect the reactorinstrumentation, control and safety system.Manual protection equipment includes a dry stand pipe system in each building stairwell.Portable extinguishers such as C02, halon and- dry chemical are placed in specific locationsthroughout the building.Elements of the passive fire protection include the structural construction system and thearchitectural separation into two separate buildings. Building structural materials are concretecast in place for foundation, concrete walls, support columns and roof. Steel beam, metal andconcrete deck comprises the reactor bay roof. A built-up composition roof with fire barriermaterials completes the roof system. The building has pre-cast panels that are cast at theconstruction site cover 75% of the external perimeter. Metal paneling covers the other 25% ofthe perimeter. Design and installation of systems'and components are subject to the applicablebuilding codes. .-"The common wall between the academic win g.and the reactor bay wing of the building is a firebarrier. Doors between these two building sections and other penetrations such as HVACchases will conform to applicable codes. Although a few metal stud and plaster board wallshave been used in the reactor bay wing, the typical wall system is of concrete blockconstruction. .Design specifications are to meet life-safety requirements appropriate for the conditions. Thesespecifications have requirements for emergency lighting, stairwells and railings, exit doors, andother building safety features.: An emergency shower -aid /ye wash are available in the hallwayadjacent to laboratory areas.,:Eachof the three components of the fire protection program is applied to the design, operationand modification of the reactor facility and cornporients. Fir'eprevention is primarily a functionof operation rather than design.9.5.3 Operational analysis and safety functionThe University of Texas maintains an active fire protection system, with periodic testing andinspections to assure systems are prepared to respond.The halon system automatically actuates if detectors in two control room ceiling units sense aninitiating condition in coincidence. The haion system is equipped with a local alarm to promptPage 9-14 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 0 01/2012SAFETY ANALYSIS REPORT, CHAPTER 9 R'..flETIevacuation of the control room prior to system actuation; a manual override can defeat thesystem if the nature of the event does not require actuation 3f the system.Fire suppression is used only in areas where there are no, significan't quan iites of radioactivematerials or criticality concerns.9.5.4 Instruments and controlsA fire alarm panel transmits status and alarm information to the University of Texas PoliceDepartment dispatch station and a campus information network monitor.9.5.5 Technical Specifications, bases, testing and surveillancesThere are no Technical Specifications associated with fire protection.9.5 Communications systemsA communication system of typical te!ephone equipment provides basic services between thebuilding and other off-site points. Supplemental features to this system, such as intercom linesbetween terminals or points within the building and zone speakers for general announcementsare to provide additional communication within the building.9.5.1 Design basisCommunications is required to support routine and emergency operations.9.5.2 System descriptionThe telephone system is insta!led and maintained by the university. Connection of the mainuniversity telephone system is to standard commercial telephone network.. Telephones withintercom features are to be located at several locations in the building. Locations include thereactor control room, the reactor bay, and several offices. By use of the intercom feature, eachof these locpations will be able to access public address speakers in one of several buildingzones.A video camera system and a separate intercom system supplement the buildingtelecommunication network. These two systems contribute to safe operation by enhancementof visual and audio communication between the operator and an experimenter. Each systemhas a central station in the control room with other remote stations in experiment areas.A public address system allows personnel to direct emergency actions or summon help, asrequired. A building evacuation alarm system prcmpts personnel to initiate protective actions.Page 9-15 CHAPTER 9, AUXILIARY SYSTEMS 01/2012An emergency cell phone is maintained in the control room to compensate for loss of normalcommunications. A digital radio is kept in the control room to provide emergencycommunications on first responder and campus frequencies, and to compensate for loss ofnormal communications.9.5.3 Operational analysis and safety functionThe control room has'adequate capabilities to initiate and coordinate emergency response.There are multiple provisions specificallyto address failures on normal communicationschannels.9.5.4 Instruments and controlsAs specified above9.5.5 Technical Specifications, bases, testing and surveillancesThere are no specific Technical Specifications related to communications, but the reactorEmergency Plan specifies communications as indicated above.9.6 Control, storage, use of byproduct material (including labs)Experimental facilities in the reactor building include a room with 4' thick walls supportingirradiation programs and a series of laboratories in the lab andoffice. wing.9.6.1 Design basisThe design;basisof the NETL laboratories is to allowthe safe arid controlled use of radioactivematerials-. .9.6.2 System DescriptionStrategic lab and office wing rooms are equipped with fume hoods and ventilation to controlthe po.tential.for release of radioactive materials.: One rocm is equipped with two pneumatictransfer systems and a manual port. One system terminates in-a fume hood, monitored by aradiation detector. The other system delivers samples within the tube to a detector. Themanual port allows samples to be transferred from the reactor bay to the lab without exitingthe reactor bay through normal passageways. A more complete description of the associatedlaboratories is provided: in Chapter 10.Page 9-16 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 90.9 MNtII01/20129.6.3 Operational analysis and safety functionEngineered controls permit safe handling of radioactive materials.9.6.4 Instruments and controlsAn installed radiation monitor ensures persornel handlir.g samples from the manual pneumaticsample transfer system are aware of the potential exposure.9.6.5 Technical Specifications, bases, testing and surveillancesThere are no specific Technical Specifications related to the laboratories; all operations involvedwith potential radiation exposure at NETL are managed under the approved reactor RadiationProtection Program.9.7 Control and storage of reusable componentsSeveral experiment facilities that are used in the core are designed to be removed and insertedas required to support various programs.9.7.1 Design basisManagement of experiment facilities is designed to minimize potential exposure to personnel.9.7.2 System descriptionThe 3 element facility, 6 element faiity, pneumatic in-core terminals, and centra! thimble aredescribed in chapter 10. Once irradiated, these facilities are maintained with activated portionsin the pool, using pool water as shielding or in other locations typically within the reactor bay9.7.3 Operational analysis and safety functionMaintaining irradiated facilities under water minimizes potential exposure. Corncrete blocksprovide temporar/ shielding as needed.9.7.4 Instruments and controlsInstruments and controls associated with specific facilities are addressed in Chapter 10.Page 9-17 CHAPTER 9, AUXILIARY SYSTEMS 1 01/20129.7.5 Technical Specifications, bases, testing and surveillancesThe basis for Technical Specifications specific to the pool is in Chapter 5, the basis forexperiment in Chapter 10.9.8 Compressed gas systemsThere are two separate compressed air systems use at the UT facility. One system provides airfor laboratories and service connections. One system provides control air.9.8.1 Design basisService air is provided to support laboratory and service operations with high capacityapplications. Instrument air is intended to support HVAC and reactor operations.9.8.2 System descriptionOne dual compressor system provides oil free compressed air for laboratories and services. Thelab air compressor motor is rated at 30 hp. The other system also uses a dual compressor andmotor, with 2-stage compressors. The instrument air compressor provides air to HAVCpneumatic controls, pool cooling flow controls. The laboratory air compressor provides aiur toshops and to the transient rod drive system.9.8.3 Operational analysis and safety functionThe two systems have dual motors and compressors to provide maximum reliability. The twosystems are connected through a manual shut off valve, providing maximum flexibility in theevent of a system (or associated air dryer) failure.Failure of the instrument air system will prevent air from supporting control systems. The pulserod drive system requires air to couple the drive to the rod; a failure will cause the rod to fallinto the core. This is a fail-safe condition, causing negative reactivity to be inserted in the core.Instrument air failure will cause chill water flow control valves to shut, stopping pool cooling.This is a fail-safe condition that prevents potential leakage from the pool to the chill watersystem. Other operational aspects of this type of event are addressed in Chapter 13.Instrument air failure will cause isolation dampers in the confinement ventilation system to failclosed, initiating confinement isolation. This is a fail-safe condition, assuring that there is nopotential for inadvertent release of radioactive material into the environment in the absence ofinstrument air.Page 9-18 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 90000 ETI01/20129.8.4 Instruments and controlsThe air compressors and their associated moisture reduction systems are locally controlled.The compressors and air dryers have operating indicators.9.8.5 Technical Specifications, bases, testing and surveillancesThere are no Technical Specificatorns specifically associated with the compressed air systems.Page 9-19