ML12156A097

University of Texas at Austin - *Redacted* Request for Renewal of Facility Operating License R-129

ML12156A097
Person / Time
Site:	University of Texas at Austin
Issue date:	12/12/2011
From:	Biegalski S University of Texas at Austin
To:	Lising A J Document Control Desk, Office of Nuclear Reactor Regulation
Lising A J
References
TAC ME7694
Download: ML12156A097 (437)

Text

UNIVERSITY OF TEXAS AT AUSTIN RESEARCH REACTOR LICENSE NO. R-129 DOCKET NO. 50-602 UNIVERSITY OF TEXAS AT AUSTIN LICENSE RENEWAL APPLICATION DECEMBER 12, 2011 REDACTED VERSION* SECURITY-RELATED INFORMATION REMOVED *REDACTED TEXT AND FIGURES BLACKED OUT OR DENOTED BY BRACKETS TDepartment of Mechanical Engineering) THE UNIVERSITY OF TEXAS AT AUSTINNuclear Engineering Teaching Laboratory Austin, Texas 78758512-232-5370" FAX 512-471-4589" http://www. me. utexas.edul- netl/December 12, 2011ATTN: Document Control Desk,U.S. Nuclear Regulatory Commission,Washington, DC 20555-0001Allan Jason LisingProject ManagerDivision of Policy and RulemakingResearch and Test Reactors Licensing BranchSUBJECT: Docket No. 50-602, Request for Renewal of Facility Operating License R-129Sir:In accordance with direction provided by L. N. Tranh (ADAMS ML110040316), we respectfully requestrenewal of the University of Texas at Austin TRIGA II nuclear research reactor located at the NuclearEngineering Teaching Laboratory at the University of Texas at Austin. Enclosed you will find:1) A completed, updated Safety Analysis Report (SAR)2) Financial qualifications specified in 10CFR50.33 is incorporated in the SAR Chapter 15;a. None of the provision of 10CFR50.33(d) apply.b. Based on current budget and expenditures, estimated annual operating costs with thesource of funding indicated for the first 5-year period after license renewal isincorporate in Chapter 15.3) Financial qualifications regarding decommissioning is provided in Chapter 15,a. An estimate of decommissioning based on NR guidance,b. A statement of intent regarding intent to seek support for decommissioning at theappropriate time,c. A description of cost adjustment of decommissioning costs is provided with the currentestimate based on the methodology,d. Documentation that the University of Texas is a State agency and a State of Texasgovernment licensee under 10CFR50.75*(2)(iv), and that University funding obligationsare backed by the State.4) An Environmental Report5) The Technical Specifications6) The University of Texas facility has currently approved programs on file with the NRC forOperator Requalification Program, Emergency Plan, and Physical Security Plan, and does notpropose changes in these plans and programs at this time.d(UL(

Your attention in this matter is greatly appreciated.I declare under penalty of perjury that the foregoing is true and correct.If you have any questions relating to this submission, please feel free to contact me by phone at 512-232-5380 or by email at biegalski@mail.utexas.edu.Regards,Steven Biegalski, Ph.D., P.E.Director, Nuclear Engineering Teaching LaboratoryCc: Paul Whaley, Associate Director, NETL

.AETLSafetyAnalysisReportThe University of Texas at AustinNuclear Engineering Teaching LaboratoryTRIGA Mark II Nuclear Research Reactor.. .. ...... ...... ........ .... .... ... ..... ............ ..... .I.. .: ..: -.. .... ........ ........ ... : , .. ' ...... ... .... ... ........... ...2 -... .... ..... ..License R-129Docket 50-60212 December 2011The University of Texas at AustinNuclear Engineering Teaching Laboratory10100 Burnet Rd, Bldg 159Austin, TX 78758 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORTTable of ContentsSection Page1. THE FACILITY 1-11.1 Introduction 1-11.2 Summary and conclusions on principle safety considerations 1-11.2 General description of the facility 1-2A. Site 1-3B. Building 1-3C. Reactor 1-3C.1 Reactor Core. 1-4C.2 Reactor Reflector. 1-5D. Reactor Control. 1-5E. 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 (BP5) 1-10F Other Experiment and Research Facilities 1-101.3 Overview of shared facilities and equipment 1-101.3.3 Reference the other facilities operating history, safety and reliability 1-101.4 Summary of operations 1-121.5 Compliance with NWPA of 1982 1-121.6 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 3-33.1.2 Control Rods 3-43.1.3 Core and structural Support 3-5 SAFETY ANALYSIS REPORT, TABLE OF CONTENTS 11/2011Table of ContentsSection Page3.1.4 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 3-63.1.7 Ventilation Systems 3-73.1.8 Instruments and Controls 3-83.1.9 Sumps and Drains 3-83.3 Water Damage 3-93.4 Seismic Damage 3-9A. Core and structural Support 3-10B. Pool and pool cooling 3-10C. Building 3-104.0 Reactor 3-104.1 Summary description 4-14.2 Reactor Core 4-14.2.1 Reactor Fuel 4-2A. Fuel matrix, 4-2A (1) Fabrication 4-3A (2) Physical Properties 4-4A (3) Operational Properties 4-6A (4) Neutronic Properties 4-7A (5) Fuel Morphology & Outgassing 4-8A (6) Zr water reaction 4-9A (7) Mechanical Effects 4-10A (8) Fission Product Release 4-10B. Cladding 4-104.2.2 Control Rods and Drive Mechanisms 4-13A. Control Rods 4-13B. Standard Control Rod Drives 4-16C. Transient Control Rod Drive 4-16D. Control Functions 4-18E. Evaluation of the Control Rod System 4-194.2.3 Neutron Moderator and Reflector (Core Structure) 4-19A. Upper grid plate 4-19B. Reflector 4-21B (1) Radial Reflector 4-21B (2) Graphite Rods. 4-23B (3) Axial Reflector 4-23C. Lower grid plate 4-23ii THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORTTable of ContentsSection Page4.2.4 Neutron Startup Source 4-254.2.5 Core support structure 4-25A. Core Support Platform 4-25B. Safety plate 4-264.3 Reactor Pool 4-264.4 Biological Shield 4-294.5 Nuclear Design 4-304.5.1 Normal Operating Conditions 4-314.5.2 Nominal Reactivity Worth Values 4-314.5.3 Reactor Core Physics 4-32A. Reference Calculations 4-33B. Prompt Negative Temperature Coefficient 4-344.5.4 Operating Limits 4-37A. Core Peaking Factors 4-37B. Power distribution within a Fuel Element. 4-39C. Power per rod 4-394.6 Core Reactivity 4-434.7 Thermal Hydraulic Design 4-454.7.1 Heat Transfer Model 4-464.7.2 Results 4-47Appendix 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 5-136.0 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 7-3iii SAFETY ANALYSIS REPORT, TABLE OF CONTENTS 11/2011Table of ContentsSection Page7.1.2. 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) 9-29.2.1 Design basis 9-39.2.2 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 9-69.3 Auxiliary Purge System 9-79.3.1 Design basis 9-79.3.2 System description 9-79.3.3 Operational Analysis and Safety Function 9-79.3.4 Instruments and controls 9-89.3.5 Technical Specifications, bases, testing and surveillances 9-89.4 Fuel storage and handling 9-89.4.1 Design basis 9-89.4.2 System description 9-99.4.3 Operational analysis and safety function 9-109.4.4 Instruments and controls 9-119.4.5 Technical Specifications, bases, testing and surveillances 9-119.5 Fire protection systems 9-119.5.1 Design basis 9-119.5.2 System description 9-129.5.3 Operational analysis and safety function 9-139.5.4 Instruments and controls 9-139.5.5 Technical Specifications, bases, testing and surveillances 9-139.5 Communications systems 9-139.5.1 Design basis 9-139.5.2 System description 9-149.5.4 Instruments and controls 9-149.5.5 Technical Specifications, bases, testing and surveillances 9-149.6 Control, storage, use of byproduct material (including labs) 9-14iv THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORTTable of ContentsSection Page9.6.1 Design basis 9-159.6.2 System description (drawings, tables) 9-159.6.3 Operational analysis and safety function 9-159.6.4 Instruments and controls 9-159.6.5 Technical Specifications, bases, testing and surveillances 9-19.7 Control and storage of reusable components 9-159.7.1 Design basis 9-159.7.2 System description 9-159.7.3 Operational analysis and safety function 9-169.7.4 Instruments and controls 9-169.7.5 Technical Specifications, bases, testing and surveillances 9-169.8 Compressed gas systems 9-169.8.1 Design basis 9-169.8.2 System description 9-169.8.3 Operational analysis and safety function 9-169.8.4 Instruments and controls 9-179.8.5 Technical Specifications, bases, testing and surveillances 9-1710.0 EXPERIMENTAL FACILTIES AND UTILIZATION 10-110.1 Summary Description 10-110.2 In-Core Facilities10-310.2.1 Central Thimble (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)10-810.2.2.1 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 10-13V SAFETY ANALYSIS REPORT, TABLE OF CONTENTS 11/2011Table of ContentsSection PageA. 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 SPESIFICATIONS10-22C. 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)10-25H (5) Mass10-25H (6) Structural 10-2510.2.3 Rotary Specimen Rack 10-26A. DESCRIPTION 10-26B. DESIGN SPECIFICICATIONS10-26C. REACTIVITY 10-28D. RADIOLOGICAL ASSESSMENT 10-28E. INSTRUMENTATION 10-29F. PHYSICAL RESTRAINTS, SHIELDS OR BEAM CATCHERS10-29G. OPERATING CHARACTERISTICS 10-29vi THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORTTable of ContentsSection PageH. SAFETY ASSESMENT 10-2910.3 Beam Ports10-29A. DESCRIPTION 10-29B. DESIGN AND SPECIFICATIONS10-30C. REACTIVITY 10-30D. 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 SPECIFICATIONS10-37C. 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 10-4311 Radiation Protection and Waste Management 11-111.1 Radiation Protection 11-111.1.1 Radiation Sources11-111.1.1.1 Airborn Radiation Sources11-111.1.1.2 Liquid Radioactive Sources11-311.1.1.3 Solid Radioactive Sources11-411.1.2 Radiation Protection Program 11-611.1.2.1 Management and Administration 11-611.1.2.2 Health Physic Procedures 11-1111.1.2.3 Radiation Protection Training 11-1111.1.2.4 Audits of the Radiation Protection Program 11-12vii SAFETY ANALYSIS REPORT, TABLE OF CONTENTS 11/2011Table of ContentsSection Page11.1.2.5 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 11-1611.1.5.4 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 11-1811.2 Radioactive Waste Management 11-1911.2.1 Radioactive Waste Management Program 11-1911.2.2 Radioactive Waste Controls 11-2011.2.2.1 Gaseous Waste 11-2011.2.2.2 Liquid Waste 11-2011.2.2.3 Solid Waste 11-2011.2.2.4 Mixed Waste 11-2111.2.2.5 Decommissioning Waste 11-2111.2.3 Release of Radioactive Waste 11-2112 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 Nuclear Engineering Teaching 12-3Laboratory12.1.2.5 Associate Director of Nuclear Engineering 12-3Teaching Laboratory12.1.2.6 Reactor Oversight Committee 12-412.1.2.7 Radiation Safety Officer 12-4viii THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORTTable of ContentsSection Page12.1.2.8 Radiation Safety Committee 12-412.1.2.9 Reactor Supervisor 12-512.1.2.10 Health Physicist 12-512.1.2.11 Laboratory Manager 12-512.1.2.12 Reactor Operators12-512.1.2.13 Technical Support 12-512.1.2.14 Radiological Controls Technicians12-512.1.2.15 Laboratory Assistants12-612.1.3 Staffing 12-612.1.4 Selection and Training of Personnel 12-712.1.4.1 Qualifications12-712.1.4.2 Job Descriptions12-712.1.5 Radiation Safety 12-912.2 Review and Audit Activities12-912.2.1 Composition and Qualifications 12-1012.2.2 Charter and Rules 12-1012.2.3 Review Function 12-1012.2.4 Audit Function 12-1112.3 Procedures 12-1112.4 Required Actions 12-1212.4.1 Safety Limit Violation 12-1212.4.2 Release of Radioactivity 12-1312.4.3 Other Reportable Occurrences 12-1312.5 Reports 12-1312.5.1 Operating Reports 12-1412.5.2 Other or Special Reports 12-1412.6 Records 12-1512.6.1 Lifetime Records 12-1512.6.2 Five Year Period 12-1512.6.3 One Training Cycle 12-1612.7 Emergency Planning 12-1612.8 Security Planning 12-1612.9 Quality Assurance 12-1612.10 Operator Requalification 12-1712.11 Startup Program 12-1812.12 Environmental Report 12-1813.0 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 13-4ix SAFETY ANALYSIS REPORT, TABLE OF CONTENTS 11/2011Table of ContentsSection Page13.3.1 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 Analysis13-11E. DAC Consequence Analysis13-14F. Effluent release Consequence 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 13-3113.5.1 Initial Conditions, Assumptions, and Approximations 13-3313.5.2 Heat Transfer to Air 13-33A. Buoyancy Forces13-34B. Friction Losses13-34C. Losses from Flow Restrictions 13-3413.5.7 Radiation Levels from the Uncovered Core 13-3813.5.8 Results and Conclusions 13-4113.6 Loss of Coolant Flow 13-4213.6.1 Initialing Events and Scenarios 13-4213.6.2 Accident Analysis and Determination of Consequences 13-4213.7 Mishandling or Malfunction of Fuel 13-4313.7.1 Initiating Events and Scenarios 13-4313.7.2 Analysis 13-4313.8 Experiment Malfunction 13-4313.8.1 Accident Initiating Events and Scenarios 13-4313.8.2. Analysis and Determination of Consequences13-44A. Administrative Controls13-44B. Reactivity Considerations13-44C. Fueled Experiment Fission Product Inventory 13-45D. Explosives 13-4613.9 Loss of Normal Electric Power 13-4813.9.1 Initiating Events and Scenarios13-48x THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORTTable of ContentsSection Page13.9.2 Accident Analysis and Determination of Consequences 13-4813.10 External Events 13-4813.10.1 Accident Initiating Events and Scenarios 13-4813.10.2 Accident Analysis and Determination of Consequences 13-4913.11 Experiment Mishandling or Malfunction 13-4913.11.1 Initiating Events and Scenarios 13-4913.11.2 Accident Analysis and Determination of Consequences 13-49Appendix 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 15.5-1xi SAFETY ANALYSIS REPORT, TABLE OF CONTENTS 11/2011LIST 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, Ai 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 4-12Figure 4.4B, Temperature, Cladding Strength, and Stress 4-13Figure 4.5, Lower Gird Plate Control Rod Positions 4-14Figure 4.6, Standard Control Rod Configuration 4-15Figure 4.7a, UT TRIGA Core 4-19Figure 4.7b, Core Top View 4-19Figure 4.8a, 6/7-Element Facility Grid 4-21Figure 4.8b, Upper Grid Plate Cut-out for 6/7-Element Grid 4-21Figure 4.9a, Reflector Top Assembly 4-22Figure 4.9b, Reflector Bottom Assembly 4-22Figure 4.10b, Graphite Reflector Through port Detail 4-22Figure 4.10c, Graphite Reflector, Radial & Piercing-Beam Ports 4-22Figure 4.11a, Tangential Beam Port Insert 4-23xii THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORTLIST OF FIGURES PageFigure 4.11b, Radial Beam Port insert 4-23Figure 4.11c, Inner Shroud Surface 4-23Figure 4.12, Reflector Component and Assembly Views 4-24Figure 4.13, Fuel Element Adapter 4-24Figure 4.14, Core Support Views 4-25Figure 4.15, Core and Support Structure Views 4-26Figure 4.16, Safety Plate 4-26Figure 4.17a, Pool 4-28Figure 4.17b, Side View 4-28Figure 4.17c, Top View 4-28Figure 4.18, Biological Shielding, Base Dimensions 4-29Figure 4.19, Reactivity Loss with Power 4-33Figure 4.20, Radial Variation of Power Within a TRIGA Fuel Rod. (Data Points from MonteCarlo Calculations [Ahrens 1999a])Figure 4.21, Critical Heat Flux Ratio (Bernath and Biasi Correlations) 4-42Figure 4.22, Core Power, 45 kW Hot Element 4-43Figure 4.23, Power Coefficient of Reactivity 4-44Figure 4.24, Unit Cell Temperature Distribution 4-50Figure 4.25, Single Rod Flow Cooling Flow Rate versus Power Level 49°C 6.5 Pool, 4-50Figure 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 9-4Figure 9.3, Confinement System Ventilation Controls 9-6Figure 9.4A, Purge Air System 9-7Figure 9.4B, Purge Air Controls 9-7Figure 9.5A, Storage Well 9-9Figure 9.5b, Fuel Storage Closure 9-10xiii SAFETY ANALYSIS REPORT, TABLE OF CONTENTS11/2011LIST OF FIGURES PageFigure 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 Temperaurer and Pulsed Reactivity 13-35Figure 13.4A, Cooling Time 13-36Figure 13.4B, Cooling Time and Power Density 13-37Figure 13.5, Core Model 13-40Figure 13.6A, Bay Model Top View 13-40Figure 13.6B, Bay Model Cross Section 13-40Figure 13.7A, Building Model 13-40Figure 13.7B, MCNP Side View 13-40Figure 13.7C, Top View 13-41xiv THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORTTable 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 4-14Table 4.5, Control Rod Information 4-15Table 4.6, Summary of Reactor SCRAMs 4-18Table 4.7, Summary of Control Rod Interlocks 4-18Table 4.8, Upper Grid Plate Penetrations 4-20Table 4.9, Displaced Fuel Spaces 4-21Table 4.10, Lower Grid Plate Penetrations 4-24Table 4.11, Reactor Coolant System Design Summary 4-27Table 4.12, Significant Shielding and Pool Levels 4-30Table 4.13, Control Rod Worth 4-31Table 4.14, Reactivity Values 4-31Table 4.15, GA-4361 Calculation Model 4-33Table 4.16, Selected TRIGA II Nuclear Properties 4-34Table 4.17, UTTRIGA Data 4-34Table 4.18, Critical Heat Flux ratio, Bernath Correlation 4-41Table 4.19, Core Power, 45 kW Hot Element 4-42Table 4.20, Reactivity Limits 4-45Table 4.21, Limiting Core reactivity 4-45Table 4.22, Thermodynamic Values 4-46Table 4.24, Coolant Temperature for 49°C 6.5 m Pool 4-47Table 4.25a, Heat Flux (Nodes 1-9) 49°C 6.5 Pool, 4-48Table 4.25b, Heat Flux (Nodes 10-15) 49°C 6.5 Pool 4-48Table 4.26, Peak Fuel Centerline Line Temperature (K) 49°C 6.5 Pool, 4-49Table 5.1, Reactor Coolant System design Summary 5-2Table 5.2, Heat Exchanger, Heat Transfer and Hydraulic Parameters 5-9Table 10.1: Composition of Al 6061 10-6Table 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 at 100 kW 10-12xv SAFETY ANALYSIS REPORT, TABLE OF CONTENTS 11/2011Table 10.6: Activity of Three Element Irradiator Cd Liner 10-18Table 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 Fuel 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 1OCFR20 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 13.71B, 1 MTU Particulate Fission Product Inventory (Ci) 13-9Table 13.8A. Gaseous Fission product Release from Single Element (pCi) 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 10CFR20 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 IOCFR20 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 13-2.1Table 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 13-38Table 13.23, Height/Thickness Dimensions of Unit Cell 13-39Table 13.24, Unit Cell Areas 13-39Table 13.25, Material Characterization 13-39Table 13.26, Post LOCA Doses 13-41Table 13.27, Calculations Supporting Limits on Fueled Experiments 13-46Table 13.28, Material Strengths 13-47Table 13.29, Container Diameter to Thickness Ratio 13-48xvi UT NETL TRIGA II Nuclear Research Reactor Safety Analysis Report THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 11. THE FACILITYThis report describes the research reactor operated by the University of Texas at Austin. Thisreport provides the basis for a safety evaluation demonstrating the facility and the reactor doesnot cause undue risk to the health and safety of the public. This chapter of the Safety AnalysisReport reflects and summarizes descriptions and analyses in the, individual chapters, and willprovide:" Introduction/overview" Summary and conclusions on principle safety considerations" General facility description* Overview of shared facilities and equipment" Comparison with similar facilities" Summary of operations" Compliance with NWPA of 1982" Facility modifications & history1.1 IntroductionThe University of Texas operates a 1.1 MW TRIGA II research reactor (with pulsing to amaximum permitted reactivity addition of 2.2% Ak/k) at the Pickle Research Campus (PRC),approximately 10 miles north of the main campus in Austin, Texas. A more completedescription of the general facility location and location within the PRC is provided in Chapter 2.This Safety Analysis Report provides information and analysis to demonstrate that there isreasonable assurance operations for an additional 20 year term do not significantly challengesafety. Analysis shows a large margin to thermal hydraulic conditions that might lead to achallenge of fuel cladding using passive, natural convection cooling.The reactor is located in the Nuclear Engineering Teaching Laboratory (NETL), a building thathouses an operating unit of the UT Department of Mechanical Engineering in the CockerelSchool of Engineering. The NETL serves a multipurpose role, with the primary function as a"user facility" for faculty, staff, and students of the College of engineering. The facility supportsthe Nuclear and Radiation Engineering program of the Department of Mechanical Engineeringfor laboratory exercises in UT courses, undergraduate research, and graduate research. TheNETL supports educational programs for other organizations and institutions notably (but notlimited to) the Big-12 Consortium and Historically Black College and Universities. The facilitysupports development and application of nuclear methods for researchers from otheruniversities, industry, and government organizations. The NETL provides nuclear analyticPage 1-1 CHAPTER 1, THE FACILITY 12/2011services to researchers, industry, and other research and industrial laboratories for testing andevaluation of materials. The NETL provides public education through tours anddemonstrations.1.2 Summary and conclusions on principle safety considerationsThe decision to build a new TRIGA was based on historical experience with a TRIGA II onmain campus. Space considerations on main campus and a well-established infrastructureat the PRC campus led to facility siting.TRIGA II reactors routinely operate at power levels up to approximately 2 MW with naturalconvection. At power levels less than 2 MW, fission product inventory is limited enoughthat emergency planning requirements are somewhat simplified. Therefore, 1.1 MW wasinitially selected as the maximum steady state license limit providing a large margin tothermal limits and complex emergency planning.Heat generation in TRIGA fuel produces less than Y2 of critical heat flux with naturalconvection at power levels up to about 2 MW (see Chapter 5). The initial license powerlevel of 1.1 MW provides an extremely large margin to thermal hydraulic limits in passive,natural circulation. The TRIGA ZrH fuel inherently reduces the potential for thermal fissionas fuel temperature increases so that temperature increases with operation at powerintrinsically limit maximum steady state power level. The TRIGA fuel design retains a largefraction of fission products generated during operation, with stainless steel cladding actingas a passive barrier to release for the fission products that escape the fuel matrix.The NETL TRIGA shielding was designed to limit personnel exposure rates from radiationgenerated during reactor operation in accessible areas of the pool and shield structure at1.5 MW to less than 1 mrem/hr. The maximum dose rate is shown to be at floor level.Current experimental programs at the beam ports limit routine access to the biologicalshielding surface near the core. Additional shielding information is provided in Chapters 3,4 and 10.The principle off-site exposure source term during normal operations is 4'Ar, a noble gaswith a 110 minute half life. Stack effluent is limited to maintain receptor doses to109CFR20 limits, as discussed in Chapters 9 and 11. There are no routine liquid releases,and the production of radioactive waste during normal operations is extremely limited (withmost radioactive waste held for decay). Accident analysis (Chapter 13) demonstrates potentialconsequences from postulated scenarios do not result in unacceptable consequences.The reactor design has many safety features, including a large margin to thermal hydrauliclimits, passive cooling, robust shielding, fuel matrix characteristics, and stainless steelPage 1-2 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 1cladding. Buildup of radioactive materials in the facility is controlled by a dynamicconfinement and an argon purge system.1.2 General description of the facilityA. SiteLand development in the area of the current NETL installation began as an industrial site duringthe 1940's. Following the 1950's, lease agreements between the University and the Federalgovernment led to the creation of the Balcones Research Center. The University became ownerof the site and in 1994 the site name was changed to the J.J. Pickle Research Campus (PRC) inhonor of retired U.S. Congressman James "Jake" Pickle.The PRC is a multidiscipline research campus on 1.87 square kilometers. The site consists oftwo approximately equal areas, east and west. An area of about 9000 square meters on theeast tract is the location of the NETL building. Sixteen separate research units and at least fiveother academic research programs conduct research on the PRC. Adjacent to the NETL site arethe Center for Research in Water Resources, the Bureau of Economic Geology, and theResearch Office Complex, illustrating the diverse research activities on the campus. ACommons Building provides cafeteria service, recreation areas, meeting rooms, and conferencefacilities. A more complete description of the environment surrounding the NETL is provided inChapter 2.B. BuildingOne of the primary laboratories contains the TRIGA reactor pool, biological shield structure, andneutron beam experiment area. A second primary laboratory has walls 1.3 meter (4.25 ft) thickfor use as a general purpose radiation experiment facility. Other areas of the building includeshops, instrument & measurement laboratories, and material handling facilities. An Annex wasinstalled adjacent to the NETL building in 2005, a 24 by 60 foot modular building. The annexprovides classroom space and offices for graduate students working at the NETL.C. ReactorThe largest room in the NETL building is a vault type enclose that serves as a confinementvolume for the UT TRIGA nuclear research reactor. The TRIGA Mark II reactor is a versatile andinherently safe research reactor conceived and developed by General Atomics to meeteducation and research requirements. The UT-TRIGA reactor provides sufficient power andPage 1-3 CHAPTER 1, THE FACILITY 12/2011neutron flux for comprehensive and productive work in many fields including physics,chemistry, engineering, medicine, and metallurgyC31ETHOL RO1D DRIVrEREACTOR RIOCDEiCENRA /AL.UlqIRINU TANK[E" XPEiIR IEN T* .., .~EEHLABL E* ., *u .o--:*S ~TRANIIF[R"4* -t " iCOB[C I ..ROTAPIl.-.5SPaCs aon, , a r o leRAC;K * '1 ,.,REFLECTOR.CON O LT0 RO0 [D PRrFLOOR L]INEcore is surrounded by a reflector, a 1 foot thick graphite cylinder.C.l Reactor Core.The reactor core is an assembly of cylindrical fuel elements surrounded by an annular graphiteneutron reflector. Fuel elements are positioned by an upper and lower grid plate, withpenetrations of various sizes in the upper grid plate to allow insertion of experiments. Each fuelelement consists of a fueled region with graphite sections at top and bottom, contained in athin-walled stainless steel tube. The fuel region is a metallic alloy of low-enriched uranium in azirconium hydride (UZrH) matrix. Physical properties of the TRIGA fuel provide an inherentlysafe operation. Rapid power transients to high powers are automatically suppressed withoutusing mechanical control; the reactor quickly and automatically returns to normal power levels.Pulse operation, a normal mode, is a practical demonstration of this inherent safety feature.Page 1-4 THE UNIVERSITY OF TEXAS TRIGA 11 RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 1-' .Figure 1.2, C:ore and Support Structure DetailsC:.2 Reactor Reflector.The reflector is a graphite cylinder in an aluminum-canister. A 10" well in the upper surface ofthe reflector accommodates an irradiation facility, the rotary specimen rack (RSR), andhorizontal penetrations through the side of the reflector allow extraction of neutron beams. In2000 the canister was flooded to limit deformation stemming from material failure in weldingjoints. In 2004, the reflector was replaced with some modification, including a modification tothe upper grid plate for more flexible experiment facilities.D. Reactor Control.The UT-TRIGA research reactor can operate continuously at nominal powers up to 1.1 MW, orin the pulsing mode with maximum power levels up to about 1500 MW (with a trip setpoint of1750 MW) for durations of about 10 msec. The pulsing mode is particularly useful in the studyof reactor kinetics and control. The power level of the UT-TRIGA is controlled by a regulatingrod, two shim rods, and a transient rod. The control rods are fabricated with integralextensions containing fuel (regulating and shim rods) or air (transient rod) that extend throughthe lower grid plate for full span of rod motion. The regulating and shim rods are fabricatedfrom 134C contained in stainless steel tubes; the transient rod is a solid cylinder of boratedgraphite clad in aluminum. Removal of the rods from the core allows the rate of neutroninduced fission (power) in the UZrH fuel to increase. The regulating rod can be operated by anautomatic control rod that adjusts the rod position to maintain an operator-selected reactorPage 1-5 CHAPTER 1, THE FACILITY120112/2011power level. The shim rods provide a coarse control of reactor power. The transient rod can beoperated by pneumatic pressure to permit rapid changes in control rod position. The transientrod moves within a perforated aluminum guide tube. Details of the control rods are provide inChapter 4.The UT-TRIGA research reactor rod control system uses a compact microprocessor-drivencontrol system. The digital control system provides a unique facility for performing reactorphysics experiments as well as reactor operator training. This advanced system provides forflexible and efficient operation with precise power level and flux control, and permanentretention of operating data. A more complete description of the rod control system is providedin Chapter 7.E. Experiment Facilities.Facilities for positioning samples or apparatus in the core region include cut-outs fabricated inthe upper grid plate, a central thimble in the peak flux region of the core, a rotary specimenrack in the reactor graphite reflector, and a pneumatically operated transfer system accessingthe core in an in-core section. Beam ports, horizontal cylindrical voids in the concrete shieldstructure, allow neutrons to stream out away from the core. Experiments may be performedinside the beam ports or outside the concrete shield in the neutron beams. Areas outside thecore and reflector are available for large equipment or experiment facilities. A brief descriptionof the facilities follows; a more complete description is provided in Chapter 10.E.1 Upper Grid Plate 7L and 31 FacilitiesThe upper grid plate of the reactor contains four removable sections configured to providespace for experiments otherwise occupied by fuel elements (two three-element and two seven-element spaces). Containers can be fabricated with appropriate shielding or neutron absorbersto tailor the gamma and neutron spectrum to meet specific needs. Special cadmium-linedfacilities have been constructed that utilize three element spaces.E.2 Central ThimbleThe reactor is equipped with a central thimble for access to the point of maximum neutron flux.The central thimble is an aluminum tube extending through the central penetration of the topand bottom grid plates. Typical experiments using the central thimble include irradiation ofsmall samples and the exposure of materials to a collimated beam of neutrons or gamma rays.E.3 Rotary Specimen Rack (RSR)A rotating (motor-driven) multiple-position specimen rack located in a well in the top of thegraphite reflector provides for irradiation and activation of multiple samples and/or batchPage 1-6 THE UNIVERSITY OF TEXAS TRIGA 11 RESEARCH REACTOR a12/2011SAF ETY ANALYSIS REPORT, CHAPTER 1production of radioisotopes. Rotation of the RSR minimizes variations in exposure related tosample position in the rack. Samples are loaded from the top of the reactor through a tube intothe RSR using a specimen lifting device. A design feature provides the option of usingpneumatic pressure for inserting and removing samples.E.4 Pneumatic TubesA pneumatic transfer system supports applications using short-lived radioisotopes. The in-coreterminus of the system is normally located in the outer ring of fuel element positions, withspecific in-core sections designed to support thermal and epithermal irradiations. The samplecapsule is conveyed to a sender-receiver station via pressure differences in the tubing system.An optional transfer box permits the sample to be sent and received to three different sender-receiver stations. One station is in the reactor confinement, one is in a fume hood in alaboratory room, and the third operates in conjunction with an automatic sample changer andcounting system.E.5 Beam Port FacilitiesFive neutron beam ports penetrate the concrete biological shield and reactor water tank atcore level, as shown in Fig.1.3. The beam ports were designed with different characteristics toaccommodate a wide variety of experiments. Specimens and/or equipment supportingexperiment programs may be placed inside a beam port or outside the beam port in a neutronbeam from the beam port.Shielding reduces radiation levels outside the concrete biological shield to safe values whenbeam ports are not in use. Beam port shielding is configured with an inner shield plug, outershield plug, lead-filled shutter, and circular steel cover plate. A neutron beam coming from abeam port may be modified by using collimators, moderators and/or neutron filters.Collimators are used to limit beam size and beam divergence. Moderators and filters are usedto change the energy distribution of neutrons in beams (e.g., cold moderator).E.5 (1) Beam Port 1 (BP1)BP1 is connected to BP5, forming a through port. The through port penetrates the graphitereflector tangential to the reactor core, as seen in Figure 5-2. This configuration allowsintroduction of specimens adjacent to the reactor core to gain access from either side of thebiological shield, and can provide beams of thermal neutrons with relatively low fast-neutronand gamma-ray contamination.Page 1-7 CHAPTER 1, THE FACILITYI12/2011A reactor-based slow positron beam facility is being fabricated at BP1. The facility (TexasIntense Positron Source) will be one of a few reactor-based slow positron beams in the world.The Texas Intense Positron Source concept includes a copper source, a source transport system,a combined positron moderator/remoderator assembly, a positron beam line and a samplechamber.E.5 (2) Beam Port 2 (BP2)BP2 is a tangential beam port, terminating at the outer edge of the reflector. A void in thegraphite reflector extends the effective source of neutrons into the reflector for a thermalneutron beam with minimum fast-neutron and gamma-ray backgrounds. Tangential beamsresult in a "softer" (or lower average-) energy neutron beam because the beam consists ofscattered reactor neutrons. BP2 is configured to support neutron depth profiling applications,with a prompt-gamma neutron activation analysis sharing the beam port.Neutron Depth Profiling (NDP) Some elements produce charged particles with characteristicenergy in neutron interactions. When these elements are distributed near a surface, theparticle energy spectrum is modulated by the distance the particle traveled through thesurface. NDP uses this information to determine the distribution of the elements as a functionof distance to the surface.Page 1-8 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 1Prompt-Gamma Neutron Activation Analysis (PGNAA) Characteristic gamma radiation isproduced when a neutron is absorbed in a material. PGNAA analyzes gamma radiation toidentify the material and concentration in a sample. PGNAA applications include: i)determination of B and Gd concentration in biological samples which are used for NeutronCapture Therapy studies, ii) determination of H and B impurity levels in metals, alloys, andsemiconductor, iii) multi-element analysis of geological, archeological, and environmentalsamples for determination of major components such as Al, S, K, Ca, Ti, and Fe, and minor ortrace elements such as H, B, V, Mn, Co, Cd, Nd, Sm, and Gd, and iv) multi-element analysis ofbiological samples for the major and minor elements H, C, N, Na, P, S, Cl, and K, and traceelements like B and Cd.E.5 (3) Beam Port 3 (BP3)BP3 is a radial beam port. BP3 pierces the graphite reflector and terminates at the inner edgeof the reflector. This beam port permits access to a position adjacent to the reactor core, andcan provide a neutron beam with relatively high fast-neutron and gamma-ray fluxes. BP3contains the Texas Cold Neutron Source Facility, a cold source and neutron guide system.Texas Cold Neutron Source. The TCNS provides a low background subthermal neutron beamfor neutron reaction and scattering research. The TCNS consists of a cooled moderator, a heatpipe, a cryogenic refrigerator, a vacuum jacket, and connecting lines. The TCNS uses eightymilliliters of mesitylene moderator, maintained by the cold source system at "36 K in a chamberwithin the reactor graphite reflector. A three-meter aluminum neon heat pipe, orthermosyphon, is used to cool the moderator chamber. The heat pipe working fluid evaporatesat the moderator chamber and condenses at the cold head.Cold neutrons from the moderator chamber are transported by a 2-m-long neutron guide insidethe beam port to a 4-m-long neutron guide (two 2-m sections) outside the beam port. Bothneutron guides have a radius of curvature equal to 300 m. All reflecting surfaces are coatedwith Ni-58. The guide cross-sectional areas are separated into three channels by 1-mm-thickvertical walls that block line-of-sight radiation streaming.Prompt Gamma Focused-Neutron Activation Analysis Facility The UT-PGAA facility utilizes thefocused cold-neutron beam from the Texas Cold Neutron Source. The PGAA sample is locatedat the focal point of the converging guide focusing system to provide an enhanced reaction ratewith lower background at the sample-detector area as compared to other facilities usingfiltered thermal neutron beams. The sample handling system design permits the study of awide range of samples and quick, reproducible sample-positioning.The neutron guide and capillary focusing .assembly may be used independent of the TCNSutilization.Page 1-9 CHAPTER 1, THE FACILITY 12/2011E.5 (4) Beam Port 4 (BP4)BP4 is a radial beam port that terminates at the outer edge of the reflector. A void in thegraphite reflector extends the effective source of neutrons to the reactor core. Thisconfiguration is useful for neutron-beam experiments which require neutron energies higherthan thermal energies. BP4 was configured in 2005 to support student laboratories.E.5 (5) Beam Port 5 (BP5)A Neutron Radiography Facility is installed at BP5. Neutrons from BP5 illuminate a sample. Theintensity of the exiting neutron field varies according to absorption and scatteringcharacteristics of the sample. A conversion material generates light proportional to theintensity of the neutron field as modified by the sample.F Other Experiment and Research FacilitiesThe NETL facility makes available several types of radiation facilities and an array of radiationdetection equipment. In addition to the reactor, facilities include a subcritical assembly,various radioisotope sources, machine produced radiation fields, and a series of laboratories forspectroscopy and radiochemistry.1.3 Overview of Shared Facilities and EquipmentUtilities are provided (underground) by the Pickle research Campus infrastructure. Chill waterfor HVAC and pool cooling is provided by a central chill water plant. Electrical power isprovided by a transformer near the NETL.1.3.3 Other TRIGA FacilitiesThe inherent safety of this TRIGA reactor has been demonstrated by the extensive experienceof similar TRIGA systems throughout the world. Forty-eight TRIGA reactors are now inoperation world-wide, and .31 of these are pulsing reactors. TRIGA reactor installations inthe U.S. are reflected in Table 1.1 (shutdown or decommissioned) and 1.2 (currentlyoperating). TRIGA reactors have more than 450 reactor years of operating experience, over30,000 pulses, and more than 15,000 fuel element years of operation. The safety arises from alarge, prompt negative temperature coefficient that is characteristic of uranium zirconium hydridefuel-moderator elements used in TRIGA systems. As the fuel temperature increases, thiscoefficient immediately compensates for reactivity insertions. The result is that reactor powerexcursions are terminated quickly and safely.The prompt shutdown mechanism has been demonstrated extensively in many thousands oftransient tests performed on two prototype TRIGA reactors at the GA TechnologiesPage 1-10 THE UNIVERSITY OF TEXAS TRIGA 1I RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 1I12/2011laboratory in San Diego, California, as well as other pulsing TRIGA reactors in operation. Thesetests included step reactivity insertions as large as 3.5% Ok/k with resulting peak reactorpowers up to 8400 MW(t) on TRIGA cores containing similar fuel elements as are used in thisTRIGA reactor.Because the reactor fuel is similar, the experience and tests form other TRIGAinstallations apply to this TRIGA system. As a result it has been possible to use acceptedsafety analysis techniques applied to other TRIGA facilities to update evaluations with regardto the characteristics of this facility.Table 1.1, SHUTDOWN OR DECOMMISSIONED U.S. TRIGA REACTORSGA-TRIGA IIITRIGA MK F, NORTHRUPUT TRIGA UNIV TEXASBRR UC BERKELEYTRIGA MK I MICH ST UNIVTRIGA COLUMBIA UNIVTRIGA PUERTO RICO NUC CTRUI-TRIGA UNIV. ILLINOISNRF NEUTRON RAD FACILITYTRIGA CORNELLDORF TRIGA MARK FATUTRGA-TRIGA FGA-TRIGA IUI-TRIGA MK ITRIGA, VET. ADMIN.thermalpower1,500.001,000.001,0001,0002502502,0001,5001,00050025025025025010020typeTRIGA MARK IIITRIGA MARK FTRIGA MARK ITRIGA MARK IIITRIGA MARK ITRIGA MARK IITRIGA CONVTRIGA MARK 11TRIGA MARK ITRIGA MARK 11TRIGA MARK FTRIGA MARK ITRIGA MARK ITRIGA MARK ITRIGA MARK ITRIGA MARK Iinitial crit1/1/19661/1/19631/1/19638/10/19663/21/19691/1/19778/1/19607/23/19693/1/19771/1/19621/1/19611/1/19897/1/19605/3/19588/1/19606/26/1959Table 1.2, U.S. OPERATING RESEARCH REACTORS USING TRIGA FUELthermal initial critpowerANN. CORE RES. REACTOR (ACRR)UC DAVIS/MCCLELLAN N. RAD. CENTEROSTR, OREGON STATE UNIV.TRIGA II UNIV. TEXASNSCR TEXAS A&M UNIV.UWNR UNIV. WISCONSINWSUR WASHINGTON ST. UNIV.PSBR PENN ST. UNIV.AFRRI TRIGA4,0002,0001,1001,1001,0001,0001,0001,0001,000TRIGA ACPRTRIGA MARK IITRIGA MARK IITRIGA MARK IITRIGA CONVTRIGA CONVTRIGA CONVTRIGA MARK CONVTRIGA MARK F6/1/19671/20/19903/8/19673/12/19921/1/19623/26/19613/13/19618/15/19551/1/1962Page 1-11 CHAPTER 1, THE FACILITYI12/2011Table 1.2, U.S. OPERATING RESEARCH REACTORS USING TRIGA FUELthermal initial critpowerANN. CORE RES. REACTOR (ACRR) 4,000 TRIGA ACPR 6/1/1967GSTR GEOLOGICAL SURVEY 1,000 TRIGA MARK I 2/26/1969DOW TRIGA 300 TRIGA MARK I 7/6/1967ARRR 250 TRIGA CONV 7/9/1964RRF REED COLLEGE 250 TRIGA MARK I 7/2/1968UCI, IRVINE 250 TRIGA MARK I 11/25/1969KSU TRIGA MK II 1,250 TRIGA MARK II 10/16/1962NRAD 250 TRIGA MARK II 10/12/1977MUTR UNIV. MARYLAND 250 TRIGA MODIFIED 12/1/1960TRIGA UNIV. UTAH 100 TRIGA MARK I 10/25/1975UNIV. ARIZONA TRIGA 100 TRIGA MARK I 12/6/19581.4 Summary of operationsThe UT TRIGA reactor has operated routinely since 1991 except for time required implementinga digital control system as a planned upgrade, and time to replace a failed reflector. Thenumber of days of reactor operation by year is provided in Fig. 1.4A, and the total energygeneration per year in Table 1.4B. The reactor is operated to meet demands of experimentalprograms and service work, with the only limit on operating time associated with personnelavailability.0250 10-z 200&150.0..-10050-z 1992 1995 1998 2001 2004 2007 2010Year of OperationFigure 1.4A, Days of Operation per Year3530025 i, .20 " ... .... .E.jý2015l101992 1995 1998 2001 2004 2007 2010Year of OperationFigure 1.4B, Burnup per Year1.5 Compliance with NWPA of 1982Compliance with NWPA of 1982 is assured by the Department of Energy. A copy of the fuelsassistance contract is provided in Chapter 15.Page 1-12 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 11.6 Facility history & modificationsThe Department of Mechanical Engineering of the Cockerel College of Engineering at theUniversity of Texas supports a Nuclear and Radiological Engineering program. Development ofthe nuclear engineering program was an effort of both physics and engineering faculty duringthe late 1950's and early 1960's. The program subsequently became part of the MechanicalEngineering Department where it currently resides. The program installed and operated thefirst UT TRIGA nuclear reactor in Taylor Hall on the main campus with initial criticality in August1963, rated for 10 kilowatts; the license was upgraded for 250 kilowatts operations in 1968.The Taylor Hall reactor operated for 25 years.In October 1983, planning was initiated for the NETL to replace the original UT TRIGAinstallation. Construction was initiated December 1986 and completed in May 1989. The NETLfacility operating license was issued in January 1992, with initial criticality on March 12, 1992.Dismantling and decommissioning of the first UT TRIGA reactor facility was completed inDecember 1992.The original computers supporting the control console have been replaced, and the operatingsystem changed from DOS to a Unix based system. In December 1999 a reflector failure wasidentified. The reflector was subsequently replaced.Page 1-13 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 22.0 SITE DESCRIPTIONThe site for the TRIGA reactor facility is located in the east tract of the AJ Pickle Research Campus,an area owned and operated by The University of Texas. The Research Center is located innorthern Travis County and the City of Austin about 11.6 kilometers north-northwest of TheUniversity of Texas at Austin campus. Fig. 2.1 thru 2.4 display the facility locations in relation tosurrounding areas. Located near the transition line between hill country and rolling plains, thesite is situated about 7.4 kilometers from where the flood controlled Colorado river crosses thetransition region and Balcones fault zone. The JA Pickle Research Campus east and west tractsspan part of the inactive fault zone. The east tract is within the transition region to rolling plains.Site location of the TRIGA reactor is in the northeast region of the research center east tract.Adjacent to the north boundary of the research center and near to the eastern boundary, the sitelocation is near the intersection of Braker Lane and Burnet Road. Fig. 2.4 shows the site locationwithin the Ai Pickle Research Campus.2.1 GENERAL LOCATION AND AREAMajor activities of The University of Texas at Austin, State of Texas government, and City of Austinbusiness district are centered at respective distances of 11.6, 12.6, and 12.9 kilometers to thesouth-southwest. Distances to air traffic landing facilities in the area are approximately 15kilometers to the Austin Executive Airport and 16 kilometers to the Breakaway Park Airport. Thenearest large commercial airport (Austin-Bergstrom International Airport)is approximately 22kilometers from the NETL building.A total area of 1.87 square kilometers is contained within the Research Center area east of Loop 1(Mopac). The east side of the Center is bounded by a State highway, FM 1325, known as BurnetRoad, and the west side is bounded by a Federal highway, US 183. The two tracts are divided by arail line, formerly the Missouri-Pacific, with 0.93 square kilometers in the east tract and 0.94square kilometers in the west tract of land. Highway intersections of US 183 with Burnet Roadand with Loops 1 and 360 are within two kilometers of the site.An area of about 9000 square meters in a rectangular shape of 120 meters by 75 meters willcomprise the general site location. The 120 meter length is along the north research centerboundary. Areas for parking, landscape and access roads are within the general site area. Abuffer zone exists between the site area and activities or structures to the east and west. To thewest the buffer zone is about 55 meters by 75 meters with parking also about 60 meters by 75meters. The east buffer region is primarily open space that will provide the access to otherdevelopment projects north of the general site area.Page 2-1 CHAPTER 2, SITE DESCRIPTION 12/2011...... ...:...*.......F" MT *mrcjd. Ufta Fana... .. .... ... .. .. i -" i....... ..A ...... .....1 "CoDo.Coo.. , "-Go~~~~~V Do, 4 a-o U,~~~~~~~~. ...... .... .........._ o .---, .,* ~ ~ ~ ~ ~ ~ ~ ............. .... = = N I ~ -K.,......." ... ..., ., ,.. ... _ __L _ _. ., !.. ...., .F, " "O " 8 " K ~ K.I --Figure 2.1, STATE OF TEXAS COUNTIESPage 2-2 THE UNIVERSITY OF TEXAS TRIGA Ii RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 2Figure 2.2, TRAVIS COUNTYPage 2-3 CHAPTER 2, SITE DESCRIPTIONI12/2011Figure 2.3, CITY OF AUSTINPage 2-4 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 2I12/2011O/14/4//rb I/r~it Ii,/5 1i/if<I'Pu I ~4< 7 <1I,~ 451 ii/ Ii;i~ /*1Pig 2Figure 2.4, JJ PICKLE RESEARCH CAMPUSPage 2-5 CHAPTER 2, SITE DESCRIPTIONI12/2011Figure 2.5, LAND USAGE AROUND JJ PICKLE RESEARCH CAMPUS, 2007Page 2-6 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 2Most areas adjacent to the Research Center are developed for mixed commercial and industrialactivities including warehouses, manufacturing facilities, and small business parks (see Figure 2-5).Mixed commercial and industrial areas south and east of the Research Center are bounded byhighway US 183, highway FM 1325 (Burnet Road), and the Texas New Orleans Railroad to theeast. Approximately 2.2 square kilometers of land are enclosed by the area. Much of theremaining area to the west of the Research Center is bounded by highway US 183 and Loop 1(Mopac) and is residentially and commercially developed, with the Gateway shopping center andmultiple apartment complexes. On the southwest side of the intersection of West Braker Roadand Loop 1 is the West Pickle Research Building, shown in Fig. 2.4. Immediately north of the AiPickle Research Campus east tract is a 2.3 square kilometer commercial complex. Residentialareas are located beyond adjoining areas around the Ai Pickle Research Campus with distancesfrom the reactor facility site of 1.2 kilometers to 2.0 kilometers. Few residential structures foreither multifamily or single family units are located within a radius of 1.2 kilometers of the reactorsite.2.2 POPULATION AND EMPLOYMENTAustin is composed primarily of governmental, business, and professional persons with theirfamilies. The city has substantial light industry with little heavy industry. Many of the persons inthe local labor force are related to activities of the City and its role as a State Capitol, theUniversity and its educational and research programs, or the growing computer-based industriesthat have established headquarters in the Austin metropolitan area. Travis county hasexperienced substantial and steady population growth rates over the last several decades.Information on population of the city of Austin and Travis county is contained in Table 2.1.Since this facility's first criticality in 1992, the Austin population has increased from 466,000 to790,000 in 2010, a 70% increase. The growth rate slowed down from 2000-2004, and steadilyincreased from then until 2009. However, according to the 2010 census and predictive data, thegrowth rate will decrease over the next decade. The 2012 predicted population is 826,235 inAustin and 1,076,119 in Travis County. The annual growth rate in 2010 was 2.11% for Austin and1.58% for Travis County.Land usage of the area around Ai Pickle Research Campus is shown in Fig. 2.5. The campus issurrounded by commercial use buildings, including multiple shopping centers. There are a smallamount of mixed living areas within several miles of NETL, including apartments and small homes.Population densities for Travis County are listed in Table 2.2 with a map of demarcation lines inFig. 2.6. Population density in the area containing NETL, zip code 78758, has an average of 5659people per square mile. This is high compared to other densities in the area because this zip codeincludes a large tract of residential areas on the far east side. The Research Campus is on the farwest side of the zip code, bordering zip code 78759 with 3415 people per square mile.Page 2-7 Table 2.1, AUSTIN AND TRAVIS COUNTY POPULATION TRENDSCity of Austin Annual City of Austin City of Austin Trvs Annual Five AnnualYar Total Area Growth Full Purpose Limited Cutravoth CunyisotPopulation Rate Population Purpose Pop Cut GRowth Couny GRowth1940 87,930 111,053 214,6031950 132,459 4.2% 160,980 3.8% 256,645 1.8%1960 186,545 3.5% 212,136 2.8% 301,261 1.6%1970 251,808 3.0% 295,516 3.4% 398,938 2.8%1980 345,890 3.2% 419,573 3.6% 585,051 3.9%1990 465,622 3.0% 576,407 3.2% 846,227 3.8%2000 656,562 3.5% 639,185 17,377 812,280 3.5% 1,249,763 4.0%2001 669,693 2.0% 654,019 15,674 830,150 2.2% 1,314,344 5.2%2002 680,899 1.7% 667,705 13,194 844,263 1.7% 1,353,122 3.0%2003 687,708 1.0% 674,382 13,326 856,927 1.5% 1,382,675 2.2%2004 692,102 0.64% 678,769 13,333 874,065 2.00% 1,419,137 2.6%2005 700,407 1.20% 687,061 13,346 893,295 2.20% 1,464,563 3.2%2006 718,912 2.64% 707,952 10,960 920,544 3.05% 1,527,040 4.3%2007 735,088 2.25% 724,117 10,971 948,160 3.00% 1,592,590 4.3%2008 750,525 2.10% 739,543 10,982 978,976 3.25% 1,648,331 3.5%2009 774,037 3.13% 765,957 8,080 1,008,345 3.00% 1,706,022 3.50%2010 790,390 2.11% 777,953 12,437 1,024,266 1.58% 1,716,289 0.60%2011 812,025 2.74% 799,578 12,447 1,049,873 2.50% 1,763,487 2.75%2012 826,235 1.75% 813,776 12,459 1,076,119 2.50% 1,811,983 2.75%2013 840,695 1.75% 828,223 12,472 1,103,022 2.50% 1,861,812 2.75%2014 857,508 2.00% 845,024 12,484 1,133,356 2.75% 1,917,667 3.00%2015 872,515 1.75% 860,018 12,497 1,164,523 2.75% 1,975,197 3.00%2016 887,784 1.75% 875,274 12,509 1,196,547 2.75% 2,034,453 3.00%2017 903,320 1.75% 890,798 12,522 1,232,444 3.00% 2,100,572 3.25%2018 919,128 1.75% 906,594 12,534 1,269,417 3.00% 2,168,841 3.25%2019 935,213 1.75% 922,666 12,547 1,307,499 3.00% 2,239,328 3.25%2020 949,241 1.50% 936,682 12,559 1,343,456 2.75% 2,306,508 3.00%2025 1,022,602 1.50% 1,009,984 12,618 1,538,624 2.75% 2,673,875 3.00%2030 1,101,633 1.50% 1,089,002 12,631 1,740,812 2.50% 3,062,318 2.75%2035 1,172,228 1.25% 1,159,584 12,644 1,719,686 2.25% 3,464,732 2.50%2040 1,232,023 1.00% 1,219,367 12,656 1,921,997 2.00% 3,920,026 2.50%NOTES: 1) The Five County Austin--Round Rock MSA wholly includes these counties: Bastrop, Caldwell, Hays, Travis and Williamson.2) Population figures are as of April 1 of each year.3) Historical and current period population figures for the City of Austin take into account annexations that have occurred.4) Forecasted population figures for the City of Austin do not assume any future annexation activity.Page 2-8 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 212/2011ZIP Code Population Dens(pop/square mile)78701 261178702 514178703 399978704 573778705 1248278.717 121578719 10978721 291878722 293878723 594578724 68978725 13378726 62178727 322578728 306578729 -38&578730 29478731 347378732 486787-33 58378734 96078735 60478736 21978737 17078738 29978739 87,678741 611078742 37278744 16778745 462778746 146878747 24878748 223478749 315378750 197-378751 660178752 619078753 457278754 53678756 494478757 500978758 565978759 3415Table 2.2, TRAVIS COUNTY 2009 AUSTIN POPULATION DENSITY DISTRIBUTION BY ZIP CODEPage 2-9 CHAPTER 2, SITE DESCRIPTIONI12/2011CHAPTER 2, SITE DESCRIPTION 12/2011Azip c* -OXWPlo* Rsnacti Curter iscoqtmnsd h787late dnaS~cet to 7875sFigure 2.6, 2009 ZIP CODE BOUNDARIESPage 2-10 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 2Research activities at Ai Pickle Research Campus are diverse, and have greatly expanded since theconstruction of NETL. Research ranges from archeological research on non-vertebrate andvertebrate paleontology to structural engineering to a center for energy and environmentalresources. A full list is compiled on the UT Library site1.It is difficult to put a number on howmany people work at AJ Pickle, since the majority of permanent staff have offices on UT maincampus, and most other staff are part time student research assistants. However, at the height ofa work day during a semester, there are upwards of 1500 people on Pickle Research Campus.Immediately adjacent to the NETL building is the geology building (see Fig. 2.4), which houses theInstitute for Geophysics, the Bureau of Economic Geology, other research groups, and someadministrative offices. Expansion of other activities near the NETL site is possible in the future.Consideration is being given to expanding utility services, and the Texas Advanced ComputingCenter is undergoing a major expansion near the NETL.2.3 CLIMATOLOGYAustin, capital of Texas, is located on the Colorado River where the stream crosses the BalconesEscarpment separating the Texas Hill Country from the Blackland Prairies to the east. Elevationswithin the City vary from 120 meters to 275 meters above sea level. Native trees include cedar,oak, walnut, mesquite, and pecan.The climate2 of Austin is humid subtropical with hot summers. Winters are mild, with belowfreezing temperatures occurring on an average of less than twenty-five days each year. Ratherstrong northerly winds, accompanied by sharp drops in temperature, occasionally occur duringthe winter months in connection with cold fronts, but cold periods are usually of short duration,rarely lasting more than two days. Daytime temperatures in summer are hot, but summer nightsare usually pleasant with average daily minima in the low seventies.Precipitation is fairly evenly distributed throughout the year, with heaviest amounts occurring inlate spring. A secondary rainfall peak occurs in September. Precipitation from April throughSeptember usually results from thundershowers, with fairly large amounts falling within shortperiods of time. While thunderstorms and heavy rains have occurred in all months of the year,most of the winter precipitation occurs as light rain. Snow is insignificant as a source of moisture,and usually melts as rapidly as it falls. The City may experience several seasons in succession withno measurable rain fall.Prevailing winds are southerly throughout the year. Northerly winds accompanying the colder airmasses in winter soon shift to southerly as these air masses move out over the Gulf of Mexico."Pickle Research campus." Univesity of Texas Libraries. Web. 09 June 2011. <http://www.lib.utexas.edu/blsc/>.2 NOAA -National Oceanic and Atmospheric Administration, Web, June 2011 <http://www.noaa.gov/>Page 2-11 CHAPTER 2, SITE DESCRIPTIONI12/2011Climatology data is summarized in Fig. 2.7. Typical Austin wind data are presented in Fig. 2.8'. Theaverage length of the warm season (freeze-free period) is 270 days. Climatology andmeteorological data is tabulated in Table 2.3 through Table 2.7 .CLIMATOLOGICAL DATATEMPERATURE (DEGREES FAHRENHEIT)4 ft^",1 ---.---- t -- --ILl0 Z AVG. HIGH40 AVG. LOWJ F M A M J J A S 0 N DTEMPERATURE EXTREMES: HIGH 108LOW -2SOLAR PATH DIAGRAMSUMMER\ ", ---"- EOUINOXiMAX. ALT. 31f"30 HUMIDITY (AVG. % RELATIVE)MIDNIGHT: 76* A.M.: 84NOON: 566 P.M.: 53PREVAILING WINDSSOUTHERLYAVERAGING 0 MPHSUNSHINE (% PossILE)6050 AVG.J F M A M 1 J A S 0 N DAVERAGE ANNUAL: 61PRECIPITATION (INCHES)42 AVG.1J F M A M J .A 5 0 N 0AVERAGE ANNUAL: 38.124 HOUR EXTREME: 19.03Figure 2.7, AUSTIN CLIMATOLOGY DATA3"Climatography of Texas; Wind Rose-Austin, Texas", National Weather Service, Austin, Texas.4 NOAA, op. cit.Page 2-12 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 2I12/20111.LEGEND1 2 3 4 5 6 78 9i l i * ' I I i IMiles per hourC 0-78-1213-18MD 19 and overe Less than 0.5%Annual Frequency in PercentPercentage frequencies of occurrence can be derived atany of the 16 compass points by using the scale andmeasuring from the perimeter of the Calm Circle. Toobtain true wind direction, add 1804 to those esbownin the easern seal-circle; subtrc.t 180 from thoseLhoun :n eetern seL-ciLrcle.Figure 2.8, AUSTIN WIND ROSE DATAPage 2-13 CHAPTER 2, SITE DESCRIPTION 12/2011Destructive winds and damaging hailstorms are infrequent. On rare occasions, dissipating tropicalstorms effect the City with strong winds and heavy rains. The frequency of tornado type activityand local sightings of tornadoes and funnel clouds is presented in Table 2.85. Recent tropic stormpaths are presented in Fig. 2.9 and Fig. 2.10'.2.4 GEOLOGYThe northwestern half of Travis county is part of the physiographic province of Texas known asthe Edwards Plateau. In Travis County, this is a highly dissected plateau with wooden hills rising insome places more than 150 meters above the drainage pathways. In marked contrast, thesoutheastern half of the county is gently rolling prairie land which is part of the physiographicprovince known as the Gulf Coastal Plain. These provinces are separated by the scarp of theBalcones fault zone, which rises 30 to 90 meters above the Coastal Plain. The scarp, however, isnot a vertical cliff; it is an indented line of sloping hills leading up from the lower plain to theplateau summit.The rocks that outcrop in Travis County are primarily of sedimentary origin and of Mesozoic(Cretaceous) and Cenozoic age. They consist largely of limestone, clay, and sand strata which dipsoutheastward toward the Gulf of Mexico at an angle slightly greater than the slope of the landsurface. Therefore, in going from southeast to northwest the outcrops of progressively olderformations are encountered, and the rocks lowest in the geologic column have the highesttopographic exposure.At the reactor facility site on the east tract, the geology is of the Austin Group defined as chalk,marly limestone, and limestone with light gray, soft to hard, thin to thick bed, and massive toslightly nodular character. On the west tract, the geology changes to the Edwards Formation oflimestone and dolomite with light gray to tan, hard to soft, thin to thick bed, and fine to mediumgrain character. The separate formations are, respectively, the up and down side of a segment ofthe Mount Bonnell Fault that passes approximately along the boundary of the east and westBalcones Research Center tracts. Distance to the fault is about 500 meters from the reactorfacility site.The Balcones fault zone, which extends from Williamson County to Uvalde County, extends thefull length of Travis County on a line passing through Manchaca, Austin, and McNeil. Here theorderly sequence of formations is replaced by an outcrop pattern controlled by the faults, most ofwhich are normal faults with the down-thrown side toward the coast. Most of the movement ofthe Balcones Fault zone occurred during the Miocene period. Since no movement has beendetected during modern times, this fault is no longer considered active7.Location of the BalconesFault zone and formations in the Austin area are depicted in Fig. 2.11.5 NOAA, op. cit.6 NOAA, op. cit.7 "Texas Earthquake Information." U.S. Geological Survey Earthquake Hazards Program, Web, June 2011,Page 2-14 (DLUU.S. Deliartinent of Commerce.National (kealc & Atamospheric Admulnli ra tionNational Enviytrwsnentul Satel~le, Data,and information ServiceClimatographyof the United StatesNo. 201971-2000National Climatic Data CenterFederal Building151 Patton AvenueAshtetlle, Not(h Carolhia 28801wWW.Iclicitoaa.govCOOP ID: 410428Station: AUSTIN CITY (CAMP MABRY),. TXClimate Division: TX 7NWS. Cal Sign:.AFTElevation:. 621 Feet Lat: 30°18NLon: 97°42Wt-T0CuCDN.Jz 0X r'.J>0(jG)Temperature (* F)".Mean () Extremes Degree.IDaystit Mean Number of1.Days OBase Temp 65MenN m e fD y (32a~ao LO~aO Max Max Max Max min NilnMonth Daily Daily Mean Year Day MAmd1) Year L Year Day %,,~1 Pear I at'ing n >"Mean Mean 100 90 50 32 32 0t1a,1 60.3 40.0 50.2 90 1971 30 .57.3 1990 ;2 1949 31 40.7 1979 47 7 .0 @ 24.1 .4 6.6 ,0Feb 65.1 54.6 99 1996 21 62.3 1999 7 1951 2 45.2 1978 319 18 .0 .3 24.4 3 3.5 .0Mar 725 o509 61.7 98 1971 20 66.9 2974 10+ 19410 12 57.2 1996 163 59. .0 .6 30.2 .0 .8 ,0Apr 78,9 57.6 68.3 98+ 2000 23 73.5 1972 3 1940 13 63.0. 1997 44 14' .0 1.6 30.0 .0 .0 .0May 54.8 65A 75.1 102 1998 7 80.6 1996 43 195- 4. '0.5 1976 2 323 .2 31.0 .0 .0 .0Jun 90.9 71.1 5.0 108 1998 14 .86:4 199q 53 1970 3 77.9 1983 0 495 10 268 30.0 .( .0 .0Jul 95.0 73.4 84.2 109 1954 26 88.0 1998 64+ 1970 23 80. 1 1976 0 605 4.3 20.0 31.0 .0 .0 .0Aug 95.6 73.3 84.5 107 2000 31 88-3 1999 61+ 196' 13 80.9 1992 0 610 5.6 28.2 31.0 .0 .0 .0Srp 90.1 68.8 79.5 ]12 2000 5 84.2 1977 .11 1942 27 '2. 1974 2 39 .18.2 30.0 .0 .0 .0Oct 81.4 59. -70.6 98+ 1991 12 73.9 1979 30 1993 31 61.2 1976 32 207 .0 4.4 39.9 .0 4 .0Nov.: 70.1 49.3 59.7 91 1951 13 65.6 1973 20 1976 29 52.2 1976 205 51 .0 .0 28.8 .0 .8 .01Dec 62.3 41.9 52.1 90 1955 25 .58.3. 1964 4 1989 23 41.8 19R5 406 13 .0 .0 26.2 .3 4.9 .0Sep Aug Jan JanAnn 75.9 50.0 :68.5 12 2 5 88.3 2999 -2 1949 1 20 19'9 228= 2972 11.8 1093 347.6 1.0 16.6+ Also.occurred on an earlier date(s)@ Denotes mean number of days greater than 0 but les thian .05Complete documentation available from: ,ww.ncdc.noaa.gov!oaiclimate/normalausnoimals.htmIIssue Date: February 2004 016-A.(1) From the 1971200(1 Monthly Normals(2) Derived from station's .available digital record: 1930-2(001(3) Derived:from 1971-2000 serially complete daily datal-"0 U.S. Department of CommerceNatiwfuil Oceanic & Almospjlieric Adin11strationNational Environmental Satellite, D'at.and Information ServiceClimatographyof the United StatesNo. 24)National Clhnatlc Dala CenterFederal BulldIng151 Patlton AvenueAsheille. North Carolina 28801wwictilc.,iOaa.gov1971-2000COOP ID: 410428:12.r'JCi~2friC.#30zStation: AUSTIN CITY (CAMP MABRY), TXClinmate Division: TX 7NWS Call Sign: ATTElevation: 621 Feet Lat: 30° 18NLon: 97°42W-4i0iCr'CrzlJ00= ->-rPrecipitation (inches)Precipitation Probabilities (11Precipitation Totals Mean Number Probability that the monthly/annual precipitation will be equal to or less than'theof Days (3) indicated amountMeansm. Daily Precipitatin Monthal'Annual 'reclpllallon vs Probablllty LevelsMdlans(l* These values were detcnlined irom the nt'omplele gamma dlsitilullonMonth Mean nipd .Yeai Day Y' ear 7-' Year 01 010 0.5 1.00 .05 .10 .20 .30 .40 .50 ý60 .70 so .90 .95la. 0.11301 &ManOt31 Mahl.w .0h9)05 10.lan 1.09 1.39 4AI 1991 9 9.21 1991 04 191. 7.7 ...3 12. 23 ..47 3 1.02 1.3.5 1.76 2,29 3 l 4.25 5,47Feb 1.99 2.00 3=05 1958 21 6.56 1992 .03 1999 7. 3.7 5 .3 ..21 .36 64 92 122 1.55 1.94 243 3.10 4.20 5.28Mai 2.14 2.09 2:69 1980 27 6.03 1983 .ofi 1972 7.9 4.4 1.4 .5 33 .61 .9' 12' 1.57 1.8 2.22. 2.63 3.17 4.01 4.81Apr 2.51 2.11 3.56 1976 18 813 1976 .06 19&4 7.2 3.9 1.7 .7 .28 .4'7 .82 1.t1 1.55f 1.97 2.46 3.07 390 5.28 6.62May 5113 5.38 5-S5 1979 21 9.49 1995 .73 1998 9.5 6.0 3.1 1.6 1.13 1.60 2.34 3.00 3.66 4.37 5.16 6.11 735 9.34 11.21.hm .3 3.05 8.040) 194 1 14.96 1981 .21 1974 7.5 5.2 2.1 ..3 .42 .70 1.23 1.76 233 2.97 3 .465 5.92. 8.02 10.0Jul 1.97 1.34 5.20 1936 16 10.54 1979 .00 1993 5.1 3.1 1.2 .5 .04 .15 40 .67 .99 1.36 1.31 238 3.20 4-58 5.96Aug 2.31 130. 5.68 1994 9. 8.90 1974 .1161: 19,7 5.2 3.3 1.4 .7 -.06 .15 .37 -66 1.01 1.44 1.99 2:72 3.78 ..63 7.52Sep 2.91 2.15 4.71 1973 26 7.44 1973 .27. 1989 7.2 4.4 1.9 .8 .44 .68 El 1.51 1.93 2.38 2.91 3.56 4.421 5.83 '19Orl 3.97 2.89 6.24 1998 17 12.39 1998 :31 1987 7:4 5.1 2.4 1.2 .39 .67 1.21 1.7' 2.36 3.04 3.04 4.84 621 3,48 10.?INov 2.60 2.4 '7.55 2001 is .7.95 2000 .1i 1999 8.2 4.3 1.7 .7 32 .53 90 1.28 1.68 2.12 241 :3.28 4.15. .5;8 6.598Dec 2.44 1'78 4.21 1991 20 14.16 1991 .14 1909 7.9 4:0 1.4 .' .18 .33 '64 .98 135 1.73 2.30 2.96 3.87 5,41 6.92Ann 33.65 33.90 8.00 .. 3 7 14.96 .Jim 001 Jul 8 I.8 S1.2 21:2 9.3 20.94 23.28 2634 28. 1 3o ,1 32.93 3511 37.55 40.55 44.9' 48.811941 1981 1993+ Also occurred on an earlier date(s)# Denotes amounts of a trace@ Denotes mean number of days greater than 0 but less than .05i Statistics not computed becaus9e less than six years out of thirty had measurable precipitation(1) From the 1971-201(. Monthly Normnals(2) Derived from station's availabldigital record: 1930-2001(3) Derived from 1971-2000 serially complete daily dataCompleie documentation available from:wsvw.ncdc.noaa:gov;oaiclimate/normalsiuanormals.htmlI0)fr'i-O0l016-Bt'Jk2 02a..JU.s. Department or Comn erceNationtal Oceanic & Atmospheric AdmnihlstratlonNational EmIvtronental Satellite, Data,andtInfor motion ServicesStation: AUSTIN CITY (CAMP MABRY), TXClimate Division: TX 7 NN'WS Call Sign. ATTClimatographyof the United StatesNo. 201971-2000National Climatic Data.CenterFederal Building151 Patton AvenueAsheville' North Carolina 28801w-ww.ncdc.noaa.govCOOP ID: 410428Lon: 97°42WElevation: 621 Feet Lat: 30°18NSnow (inches)-D00V,"r"zm--I--4m(j0r-0-4Snow Totals Mean Number of Days (i)Snow Fail Snow. DepthMeans/Medians (o) Extriemes t2)>= Thresholds >= ThresholdsHighest Highiest HighestSnow. Snow Snow Snow MonthlyMonth Fall Fali Depth DphMYer Day Year Year Day Mean Year 0.1 1.0 3.0 5.0 10.0 1 3 toMeau Median Mean Median Snow Snow Snow SowFall Fail DepthtDepthJan .4 .0 # 0 3.9 1985 2 75 1985 4 1985 13 F 1985 .3 1 1 0 0 2 1 0 0Feb .1 .0 # 0 1.2 1985 1 1:2 1985 ta 1985 2 # 1985 .2 .0 0 0 .0 .1 .0 .a .0Mar 1i .0 0 0. t 1994 9 hi .1994 0 0 0 0 0 .0 .0 .0 .0 .0 .0 .0 .0 .0Apr .0 .0 0 0 .0 0 0 .0. 0 0 0 0 0 0 .0 .0 .0 .0 0 0 .0 .0 .0May 0 .0 # 0 .0 0 0 ,0 0 0 0 I # 1994 ,0 .0 ,0 0 .0 0 .0 0 .0Jun .0 .0 0 0 .0 0 0 .0 0 0 0 0 0 0 .0 1 0 0 .0 .0 .0 .0 .0 .0Jul .0 .0 ( 0 .0 4 0 .0 0 0 0 0 0 .0 .0 ..00 .. .0 .0 .0Aug ,0 .0 P 0 :0 0 0 0 0 0 0 0 Pi 1997 .j .0 .0 .0 .0 :0. .0 .0 .0.0 0 .0 0 0 0 0 0 0 0 0 0 .0 .0 .0 .0 .0 .0 .0 .0 .0Ot .0 .0 0 0 .0 0 0 0 0 0 .0 .0 0 0. .0 A ,0 .0 .0 ,0 .0 ,0Nov 1 0 0 0 1.0 1980 25 2.0 1980 4-÷ 1980 26 0 0 i1 1 0 .0 .0 .0 .0 0 0Dec

• 0 0 0 .1998 24 .#I. 1998 .4 1996 16 0 0 0 .0 0 0. 0 .0 0 .0 .0Jan .jag jia Aug.Ann .6 0WA 39 2 7.5 4 13 1+ .0 .2 .i .0 .0 .3 .1 0 .1,98" 1085 1985 1997U0M-r0~TM-~(. Denotes mean number of days greater than 0. but less than .05-9.-9;9 represents missing vailuesAnnual statistics for MeanlMedian snow depths are not appropriate(1.LDerived tronm .now Climatology and 17i 1-2u0u dally 0a0t(2) Derived from 1971-2000 daily dataComplete documentation available from:'ww.ncdc.nona.gov/oa/clim ateinormalstusnormals.htm Ii...NJ06-C U.S. Deparimeni of CommerceNational Oceanic & Atmospheric AdministrationNational Eni~Arowmnenlal Satellite, Datu,and information ServiceClimatographyof the United StatesNo. 201971-2000National Climatic Data CenterFedcral Building151 Palon'AwlnaeAsheville, Norih Carolina 28801www.ncdc.noaa.govStation: AUSTIN CITY (CAMP MABRY), TXClimate Division: TX 7 NWS Call Sign: ATT" COOP ID: 410428Lat: 30° 18N Lon: 97°42WC/--Elevation: 621 Feet-n001( 0x w>0AG)4Freeze DataSpring Freeze Dates (Month/Day)Temp (F) Probability of later date in spring (thrujul 31) than indicated(*).10 .20 .30 A4 .50 .60 .70 .80 .9036 3;29 3/21 3/15 3/10 3/06 3i01 2!24 2/18 2/1032 3/15 3/06 ,2/28 2V22 2/17 " 2 2/!07 1/31 1/2328 3106 2/24 2/17 2A 0 Vi04 U/29 V22 1113 12/0224 2/19 2/09 2/io V125 2/14 1/2107 1W00 01/00 0/20 2/07 1/27 I/I18 U/O 12/23 0/00 0/00 0100 0/0016 1/05 0/00 T000 0/00 0700 0700 0/00 0!.(00 0900Fall Freeze Dates (Month/l)ay)Temip T) Probability of earlier date in fall (beginning Aug 1) than indicated(*)t10 .20 .30 .40 .50 .60 .70 .80 .9036 11,04 11109 11 13 11]/17 11/20 11V23 1 1/26 11V30 12706.32 11115 11/22 11/27 12/02 12f06 12/10 12/I5 12/20 12/2828 11/28 12/06 12/11 12/16 12121 12/26 12/31 1/02 1/2024. 1241 12122 1/01 1/09 1/19 2M02 1!00 0/00 0!0020 12/19 1/02 1i15 1;29 0/00 0/00 0/00 0/00 0/0016 1/02 0W00 0/00 0/00 0700 0/00 0/00 0100 0/00Freeze Free PeriodTemp (F) Probability of longer than indicated freeze free period (Days).10 .20 J30 .40 .50 .60 .79 .80 .9036 285 276 269 264 259 253 248 241 23232 323 312 304 297 291 285 278 270 259.28 >365 >365 341 328 319 311 303 294 28124 '365 >365 >365 .>365 >365 347 334 323 31120 >365 >365 >365 3365 >365 >365 >365 357 33516 >365 1 365 >365 >365 >31 635 >365 15 >365-.1frDt,000*Probability o1 observinga temperature as cold, or colder, later in the spnng or.earlier in the tall than the indicated date.O!06.1ndicates that the probability of occurrence of threshold temperature is less than:the indicated probability.Derived from 1971 2000 serially compl;ete daily data Complete'.documentation available from:www.ncdc.noaa.go~v/oa/cllniteinorm alsiýsnormials.htni016-Dc'J0',

03CDU.S. Department.of Commerce.Nalional Oceanle &.Atnaospherl. AdminlisrationNational ,EUnronflmental Satellite, Data.anad lrormatlon ServiceStation: AUSTIN CITY (CAMP MABRY), TXClimato graph yof the United StatesNo.. 201971-2000National Climatic Data CenterFederal Building151 alLton AvenueAsheville, North Carolina 28801i**.ncdc.noaa.govCOOP ID: 410428'712trjHOr-CClimate Division: TX 7NWS Call Sign: ATTElevation: 621 Feet Lat: 30°18NLon: 97042WDegree Days to Selected Base Tlemperatures (*F)0z--IM,>0-I030~--IM00r-0G--I3Base Heating Degree Days iI)Below ,Jan Feb Mar Apr Moy Jun Jul Aug Sep Oct. Nov DeW Ann65 475 319 163 44 2 0 0 0 2 32. 205 406 164860 342 "203 63 8 0 .0 0 0 0 4 118 271 101557 272 152 32 .0 0 6 0 0 1 7S 209 74655 231 122 19 0 0 0 0 0 0 1 56 171 60050 146 61 4 0 0 0 " 0 0 0 21 92 32432 8 0 0 0 0 .0 0 0 0 0 0 0 8Base Cooling Degree Days (i)Above Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec Ann32 566 636 924 1095 134- 1485 1628 1635 1428 11"O 833 628 1340455 65 110 .248 '109 034 795 915 922 738 490 202 81 560957 46 84 201 352 572 735 853 860 678 431 163 60 503560 26 53 138 271 430 645 760 767 539 343 113 36 422165 7 1 59 147 323 495 605 .610 439 207 51 13 297,170 I 4 20 61 L 85 .345 150 1 .157 298 104 15 2 1942.Growing Degree Units 12)Buse Growing Degree Unis-(Moanthly) Growing Degree Units (Accumulated Moodly)Jan Felb Mar Apr Mtay Ju Jul Aug Sep Oct .Nov Dec Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec40 357 451 684 867 1108 1253 1393 1396 1195 959 605 405 357 808 1,1921 2359 31467 4720 6113 7509 8701 9663 10268 1067345 233 324 534 717 953 1103 1238 1241 1045 804 459 272 233 557 1091 1808 2761 3864 5102 6343 7388 8192 8651 892350 137 214 390 568 798 953 1083 1086 895 650 324 164 137 351 741 1309 2107 3060 4143 5229 6124 6774 .7098 726255 69 123 257 421 643 803 928 931 745 .197 1209 82 69 192 449. 870 1513 2316 32,14 4175 MI92a 5417 5626 570860 29 63 143 279 488 653 773 776 596 353 118 41 29 92 235 514 1002 1655 2428 3204 3800 4153 4271 4312Base Growing Degree Units ror Corn (Monthly) Growing Degree Units for Corn (Aeccmulated Monthly)50'86 20,1 268 I124 1 571 1 770 1 872 1 937 1 9341 1 819 1 643 1 368 9 235 204

• 472 1 896. 1 1,167 1 2237 1 3109 1 4(136 1 4980 1 579 642 1 6810 1 7045:'2.>(-3c-Io(1) Derived fromthe 1971-2000 Monthly Normals(2) Derived from 1971-2000 serially complete daily dataNote: For corn, temperatures below 50 are set:to 50, and temperatures above 86 are. set to 86016-EComplete documentation available from:www.ncdc.noaa.govioa/climateinoimalsiusnormals.htm I0 CHAPTER 2, SITE DESCRIPTION _ __....12/2011Table 2.8, TRAVIS COUNTY TORNADO FREQUENCIESMag: MagnitudeDth: Deaths16 TORNADO(s) were reported in Travis County, Texas between Inj: Injuries01/01/1992 and 02/28/2011. PrD: Property DamageCrD: Crop DamageTexasLocation or County Date Time Type Mag Dth Inj CrD11 Lago Vista 109/20/1996 '07:58 PM 'Tornado IFl ;0 0 5K 10K2 Four Pts '05/27/1997 03:11PM Tornado ° F2 10 -0 150K ....13 Four Pts 105/27/1997 03:15 PM Tornado IF1 i0 J0 5K OK4Lakeway 105/27/1997 03:50 PM Tornado iF4 i1 i5 115.OM OK..... ......r 15 Lake Travis 108/29/1998 05:45 PM 1Tornado F1 i0 0 30K......16 Bergstrom Afb 03/16/2000 04:20 PM Tornado iFO 0 i0 10;7 Pflugerville 03/16/2000 05: 1PM Tornado F" 00.~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~~o ...... ... ..... .................. .. .......... ........ .. .... ................... ......18 Austin 11/15/2001 035 M IoraoF 01 OK 1Austin 11/15/2001 04:45 PM rTornado FO 0 IO[ 30K ý0 i...... .. .... ........ .o .... .....-10 Bergstrom Afb 11/15/2001 105:30 PM Tornadol 0 0 8K..... '" ............. .....................................11 Bergstrom Afb ....... 15/2001 -O5:44-P --I Tornado 1.FO 5K i012 Pflugerville 111/15/2001 06:02 PM oTornado F. .-... ... ........... I ...... ........ ............... ... ; i O i -; ; .....a ~ .... .i ...... ...... ) ... ....... .......... ... ....... ..... ..... .. .. ... ...._......113 [aor12/23/2002 070 MTornado:Fl10 1 12_0 0*14 Austin-bergstrom Arpt 07:45 PM !Tornado- IF0 0 10 150K .15 Beecaves 111/16/2004 04:52 PM ITornado FO [ F0 0...... ............... ........ ..... .. .... ..... ........ .... ........r .T o r a d ....16 Manor 103/25/2005 109:50 PM ITornado 1F1 0 0 100K 10TOTALS: 1 6 15.765M 10KPage 2-20 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 2I12/2011Figure 2.9, TROPICAL STORM PATHS WITHIN 50 NAUTICAL MILES OF AUSTIN, TEXAS(ALL RECORDED HURRICANES RATED Hi AND UP)Figure 2.10, TROPICAL STORM PATHS WITHIN SO NAUTICAL MILES OF AUSTIN, TEXAS(ALL RECORDED STORMS RATED TROP OR SUBTROP)Page 2-21 CHAPTER 2, SITE DESCRIPTION 12/20112.5 SEISMOLOGYThirty three earthquakes of intensity IV or greater have had epicenters in Texas since 18738. Theearthquake's intensities were characterized using the Modified Mercalli Scale of 1931. The scalehas a range of I thru XII, on which an intensity of I is not felt, an intensity of III is a vibration similarto that due to the passing of lightly loaded trucks, and intensity of VII is noticed by all as shakingtrees, waves on ponds, and quivering suspended objects but causes negligible damage tobuildings of good design and construction, and an intensity of XII results in practically all works ofconstruction being severely damaged or destroyed. The strongest earthquake, a maximumintensity of VIII, was in western Texas in 1931 and was felt over 1,165,000 square kilometers. Figs.2.12 and 2.13 show the locations and intensities of earthquakes in Texas. Of these, no damagehas ever occurred to local buildings in the Austin area.2.6 HYDROLOGYAlmost the entire county is drained by the Colorado River and its tributaries. Lake Travis, which isformed by the Mansfield Dam on the Colorado River, is part of the power, flood-control, waterconservation, and recreation project of the Lower Colorado River Authority. Other lakes are alsooperated by the Authority, such as Ladybird Lake and Lake Austin, and are created by Longhornand Tom Miller dams, respectively. Low level alluvial deposits of the river are commonlysaturated with water at relatively shallow depths. Recharge is primarily from the river and localsurface contaminations are easily transmitted to this shallow water table.Ground water from subsurface formation is found in basal Cretaceous sands referred to as the"Trinity" sands. Elevations of the Trinity aquifer range from depths commonly less than 300meters east of the Balcones Fault Zone to greater than 450 meters to the west of the zone. Eastof the Mount Bonnell Fault, dolomite and dolomite limestones provide a source of ground waterat shallower depths. Access to the Edwards aquifer ranges from 30 meters to 300 meters withnatural springs occurring in areas near the Colorado River. Minor aquifers associated with theGlenn Rose Formation supplies small quantities of water west of the Balcones Fault Zone. Waterbearing areas in the formation are at varying depths and literally discontinuous. On the PickleResearch Campus east tract, wells drilled for environmental monitoring have produced groundwater at depths of less than 15 meters. Fig. 2.14 shows the location of the ground water aquifers.Water supply for the research campus and wastewater treatment is provided by the City ofAustin. Although wells into the aquifers provide substantial water the city supply is filtered riverwater. Other area municipalities and organizations utilize aquifer water. Control of private wellsis the function of county and state Health Departments. Gross beta radioactivity of city water hasbeen measured and is reported in.Table 2.9.8 "Texas Earthquake Information." U.S. Geological Survey Earthquake Hazards Program, Web, June 2011Page 2-22 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 2U., upthrown sideD, downthrown sideColor shades indicatedifferent formationsFigure 2.11, BALCONES FAULT ZONEPage 2-23 CHAPTER 2, SITE DESCRIPTIONI12/2011CHAPTER 2, SITE DESCRIPTION 12/20111 o6W 104Vr 102'W 10OW98V9*W 94WW36*N34'N32"N3OdN28"N26"NW36N34"N-32"N~3ObN-28"N-26"N3502001601208060504030201816141210864201 06-W 104-W'102-W 100-W 98-W 96-W 94-WPeak AcceleraUion (%f) wilth 2% Probebifty of Exmedance In 50 Yearsaf NEHRP B-C boundaryNadonel Seimnic Immd MmppkMg Project (2006)2-8Figure 2.12, TEXAS EARTHQUAKE DATAPage 2-24 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 2I12/20111990-200MSeismicity of Texas-10o -104- -1or -100, -96, 4 9Mr36*34"32"30"2820'2W-2W'--106*-104 -1W2 .10r -98 46' -W4-Si,-71DFIII-38 0Depth is in kilometers.Purple Triangles: CitiesPurple Star: Capital CityCircles: Earthquakes (color represents depth range)Figure 2.13, TEXAS EARTHQUAKE DATAPage 2-25

• ..., .CHAPTER 2, SITE DESCRIPTION12/2011cI .,wf* .*~ ~ mw~ ~ -a-'.Figure 2.14, LOCAL WATER AQUIFERSTable 2.9GROUND WATER ACTIVITY9(gross beta)Travis County <4.0 pCi/I.JJ Pickle Research Campus <4.0 pCi/I.9 "Water Analysis Report." Texas Department of State Health Services; City of Austin Water and Wastewater, 28March 2005.-I'Page 2-26 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 22.7 HISTORICALRelocation of the UT TRIGA reactor and related facilities to the JJ Pickle Research Campus site,previously known as the Balcones Research Center, was to help accommodate growth ofprograms both at the University main campus and at the Research Center site. The facilitylocation at the Research Center is in the north-east corner of the research center site. Referenceguidance for site evaluation was ANS 15.710.The original research center site area was operated as a magnesium manufacturing plant by theFederal government in the 1940's. Subsequent arrangements and acquisition by the Universitywould determine activities of the site throughout the 1950's, 1960's and 1970's. Activities at thesite were not fully developed prior to the 1980's. University functions or research activities weremoved to the site when required accommodations were not available on the main campus. A fewfunctions of the University at the site had resulted in the construction of major facilities suitablefor long term use. Other activities at the site have utilized existing structures or other buildingsnot suited for long term use.A major program'1 was established in the 1980's to develop the Balcones Research Center siteactivities. As part of the first phase of development, several major research programs associatedwith energy and engineering were moved to facilities constructed at the site. Features of the site,before the development activities by the University and after initial development in the 1980's,are illustrated in Fig. 2.15 and 2.17.Several activities at the Research Center prior to 1980 had been associated with radioactivematerials. These activities ranged from the burial of low level radioactive waste materials such astritium and carbon-14 in the northwest corner of the site, to water transport studies performed in30 meter diameter surface tanks. Isotopes of cesium-137, cesium-134, and cobalt-60 werepresent in sludge samples of one of the tanks, but the surface tanks contaminated withradioactive materials used for water transport studies prior to the 1980s were decontaminatedand released for unrestricted use in January 1996. Subsequently, the tanks were demolished. Thelow-level radioactive waste burial site at Pickle Research Campus was released for unrestricteduse by the Texas Natural Resource Conservation Commission (now known as the TexasCommission on Environmental Quality) on 06 August 2001. Copies of pertinent documents are onfile with UT-Austin EHS12.Radioactive waste and other materials at the Research Center site are part of the University broadlicense for radioactive materials which is managed by the University Environmental, Health, andSafety Department and issued by the Texas Department of State Health Services.10 "Research Reactor Site Evaluation", American National Standard, ANSI ANS 15.7-1979 (N379)." "Balcones Research Center Project Analysis", Volume I, The University of Texas, 1981.12 "Environmental Health and Safety (EHS) I The University of Texas at Austin." The University of Texas at Austin,Web, June 2011, <http://www.utexas.edu/safety/ehs/>Page 2-27 CHAPTER 2, SITE DESCRIPTION 1 12/2011Figure 2.15, RESEARCH CAMPUS AREA 1940Page 2-28 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 2I12/2011,,,# ,EXISTING CONDITIONS© 0 200 000 FEETNORT 00 G I- -NORTH 100 800 I ACRFigure 2.16, PICLKE RESEARCH CAMPUS 1960Page 2-29 CHAPTER 2, SITE DESCRIPTIONI12/2011W, CMM.1C ftaL-u CSUOMG 0 I~ES C2*0 MAý-s 80.10 281* hOW0*WIHA CLLRSITTOCENTERTHE UNIVERSITY OF TEXAS AT AUSTINFigure 2.17, BALCONES RESEARCH CENTER 1990Page 2-30 UT TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORTCHAPTER 2 APPENDIX 1, SUBSURFACE EXPLORATION LOGSI12/2011SUBSURFACE EXPLORATION LOGP4oJ0ET NETL Building SORI,. No. B-2Balcones Research Center OATE 8/13/85Austin, Texas Jo01 NO. 5408TYPE oF BORNG Auger/Samole/Core SURFACE ELEY. 791 .0 ft.Z ttssn~waltod tube 91 00fltetwatk) twot of no ea .cey'L ~~~double tube c,, barre 0 dsstuwbed 1 sgl I IO E S C R I P T I O N O f S T R A T A .... ...(7870)-_ (787.0)B101l52025-(765.5)-3035123Light olive gray, clayey GRAVEL w/sand, medium denseto dense, calcareous. GC(Residual Soil)C,i i i ii .. .. .. .. i ii rVery pale orange, gravelly, lean CLAY, veryw/relict structure, dark brown clay lenses,completely weathered limestone gravel. CL(Completely Weathered Austin/Vinson)stiff, Na nd/I456910Very pale orange LIMESTONE, fine-grained, slightlyX weathered, low to moderate hardness, nodular, verythin bedded, w/shaly limestone layers and scattereddiscontinuities.(Austin/Vinson)discontinuity, dip angle 60*, at 23.4 ft.N14 _R90RQD75R83 -RQD 0R92RQD92R 9 2 -RQD92 -R19RQD 0R81 rRQD62R10011013O 0 "--~- '.4. 4Total depthNOTES: (1)of boring, 25.5 ft.Boring was advanced dry, and compressedair was used during the coring oper-ation.(2) The hole was open to 15.2 ft. and thewater surface was noted at the 10.0-ft.depth on 9/9/85.1R1100 tRQOB2 JFRANK 6. BRYANT & ASSOCIATES. INC.Austin. TexasPage 2.1-1 UT TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORTCHAPTER 2 APPENDIX 1, SUBSURFACE EXPLORATION LOGSI12/201140455560657075o885SUBSURFACE EXPLORATION LOGJoe NO. 5408 DATE 5/17/85 ORING NO. B-1P-JNOTES:(1) Boring was advanced dry to the 4.5-ft. depth,and no groundwater was encountered above thatdepth.(2) Upon completion of drilling, a piezometer(2-in. I.D. PVC pipe, capped on the bottom,w/the lower 10.0 ft. slotted) was installedw/the bottom at 30.9 ft.(3) On 6/4/85, the water surface was noted at the11.7-ft. depth, and the hole was bailed to30.7 ft.(4) On 6/5/85, the water surface was noted at the29.3-ft. depth, and the hole was again bailedto 30.7 ft.(5) On 6/7/85, the water surface was noted at the29.1-ft. depth.(6) On 9/9/85, the water surface9.3-ft. depth.was noted at theI-IFRANK G. BRYANT & ASSOCIATES. INC.Austun. TexaiPage 2-2 UT TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORTCHAPTER 2 APPENDIX 1, SUBSURFACE EXPLORATION LOGSI12/2011ISUBSURFACE EXPLORATION LOGPROJECT NETL Building BONIGNO. B-3Balcones Research Center OATE 8/15/85Austin, Texas JOB NO. 5408TYPE OF BORING AIIPr!Sampfinre SURFACE 789.5 ft.LEGEND MableZ W samgflletypethin-wals tube In Do.trotin test no recover -. N. oDESCRIPeTION OF STRATA z 2: 0"(785.6)(783.0)-(180.5-101520(766.5L2530352uUsky orown, Tat C.LATY, StiTT, noncalcareous, w/few fine gravels.. CHP4. t+P4.5+P4.5+(Residual Soil)I. -I ~4Very pale orange, gravelly, lean CLAY, very stiff,w/relict structure, dark brown clay lenses, andcompletely weathered limestone gravel. CLA(Completely Weathered Austin/Vinson)I-1-fl -gX;15'- Very pale orange LIMESTONE, fine-grained, slightlyweathered, low to moderate hardness, nodular, verythin bedded, w/shaly limestone layers and scattereddiscontinuities.RIO0ORQD81-6R100-RQD52-(Austi n/Vi nson)7R100RQD50Total depth of boring, 23.0 ft.NOTES: (1) Boring was advanced dry, and compressedair was used during the coring oper-ation.(2) The hole was open to 18.0 ft. and thewater surface was noted at the 9.0-ft.depth on 9/9/85.~1FRANK G. BRYANT & ASSOCIATMS. INC.Austin. Tea..Page 2.1-3 UT TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORTCHAPTER 2 APPENDIX 1, SUBSURFACE EXPLORATION LOGSI12/2011SUBSURFACE EXPLORATION LOGPROJECT NETL BuildingBalcones Research CenterAustin, TexasTYPE OF BORING .A naVt./K2mn oICarpBORNG NO- B-4DATE 8/15/85JOE NO. 5408SURFACE ELEV-709A Ff-6LEGEND -y ,gr{oundwater table0 13ntWelld tube permtrertion keel no recovetvJ 8 U 8013b tube core 0 dishr.ted 0'DE RIPTION OF STRATAF -_(789. 1)-10 =-15--(775.9) -20-(767.4)--2530 -351245KRnilVery pale orange, gravelly, lean CLAY, very stiff,w/relict structure, dark brown clay lenses, andcompletely weathered limestone gravel. CL(Completely Weathered Austin/Vinson)Dark yellowish brown, sandy, lean CLAY, hard, cal-'careous, w/some fine gravel. CLVery pale orange LIMESTONE, fine-grained, slightlyweathered, low to moderate hardness, nodular, verythin bedded, w/shaly limestone layers and scattereddiscontinuities.S c e (Austi n/Vinson)Lsoft, clayey layer from 11.5 to 12.2 ft..6k-discontinuities,60° at 17.1 ft.dip angle 400 at 16.8 ft. and78I-Total depth of boring, 25.0 ft.NOTES: (1) Boring was advanced dry, and compressedair was used during coring operations.(2) The hole was open to 24.4 ft. and thewater surface was noted at the 16.5-ft.depth on 9/9/85.FRANK G. BRYANT & ASSOCIATES. INC.APsta, Tea2sPage 2-4 UT TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORTCHAPTER 2 APPENDIX 1, SUBSURFACE EXPLORATION LOGSI12/2011SUBSURFACE EXPLORATION LOGPROJECT NETL BuildingBalcones Research CenterAustin, TexasTYPEOFBORtNG Auger/Sample/Core8ORING NO. B-IPDATE 5/17/85JOB NO. 5408SURFACEfEv. 791.3 ft.LEGEND _rIounldwater table* ....d tube pentraion test Cc nU double tube core barre 0 disturbod C 0DESCRIPTION OF STRAlA XW cDusky brown, fat CLAY, stiff, noncalcareous, w/fewfine gravels. CH(Residual Soil)I-* I , FVery pale orange, gravelly, lean CLAY, veryw/relict structure, dark brown clay lenses,completely weathered limestone gravel. CL(Completely Weathered Austin/Vinsonstiff,andg I gl I I. l pVery pale orange LIMESTONE, fine-grained, slightlyweathered, low to moderate hardness, nodular, verythin bedded, w/shaly limestone layers and scattereddiscontinuities. (Austin/Vinson)dark yellowish orange shale stringer fromIL10.6 to 10.8 ft.N17N46R96RQD75Ri100RQD64R100RQD86R100RQD74R100RQD100Light gray, SHALY LIMESTONE, fine-grained, lowhardness, fossiliferous, very thin bedded, w/scattered discontinuities.(Austin/Atco)pale yellowish orange, slightly weathered aboveboring, 31.5 ft.FRANK G. BRYANT & ASSOCIATES. INC.Austin. TaxesPage 2.1-5 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 33.0 DESIGN OF SYSTEMS, STRUCTURES AND COMPONENTSThe Nuclear Engineering Teaching Laboratory (NETL) was built in 1989-1993. The centerpieceof the NETL is a TRIGA Mark II nuclear research reactor. Structures, systems and components(SSC) required for safe operation of the reactor, safe shutdown and continued safe conditions,response to anticipated transients, responses to accidents analyzed in Chapter 13 (AccidentAnalyses), and control of radioactive material discussed in Chapter 11 (Radiation ProtectionProgram and Waste Management) are identified in Table 3.1. The NETL TRIGA Mk II reactorwas originally licensed to operate at power levels up to 1.1 MW, with routine operations up to950 kW and special operations as required up to 1 MW. Principal functions associated withnormal operations include reactor control, heat removal, radiation shielding, gaseousradioactive material control, and shielding. The spectrum of accidents identified for TRIGA andTRIGA fueled reactors in NUCREG/CR-2387 (PNL-4028)1 includes:* Excess reactivity additionBecause of the negative temperature feedback associate with the TRIGA fuel-moderator, core design bounds excess reactivity addition scenarios." Metal-water reactionsMolten metal is required to initiate metal water reactions with zirconium; zirconiummelting point (1823°C) exceeds TRIGA fuel temperature limits (1150'C) by a largemargin. The maximum temperature that can be achieved in a TRIGA reactor iscontrolled by design (limiting maximum excess reactivity).* Lost, misplaced, or inadvertent experimentThe introduction of a lost, misplaced or inadvertent experiment scenario is controlled bythe experiment process (section 10.6), and not by facility design." Mechanical rearrangement of coreMechanical rearrangement of the core can occur in one of two ways, core crushing ormechanical rearrangement of the core. Core crushing requires the introduction of alarge mass over the reactor capable of damaging the reflector and core, and isessentially an operational concern as opposed to a design constraint. Mechanicalrearrangement requires an external force (which could be an operationally driven event,or external such as a seismic event), and would result in a decrease in reactivity.Decreasing reactivity does not challenge fuel integrity." Loss of coolant accidentLoss of coolant accident could result from a loss of pool integrity, either a break in theliner or the beam tubes. The design basis for the pool cooling and cleanup systemincludes specifications to prevent the potential for a piping failure that could siphon aNUREG/CR-2387 (PNL-4028) Credible Accident Analyses for TRIGA and TRIGA Fueled Reactors (S. C.Hawley, R. L. Kathren, March 1982)Page 3-1 CHAPTER 3, DESIGN OF STRUCTURES, SYSTEMS AND COMPONENTS .12/2011significant amount of water out of the pool. The design basis for the fuel-moderatorelements assure that decay heat will not challenge cladding integrity." Changes in morphology and ZrH, compositionChanges in fuel morphology are driven by temperature changes; design bases to limitfuel morphology issues bound potential accident scenarios." Fuel handlingNUREG/CR-2387 identifies nominal core loading of 50 fuel elements; however the UT TRIGAinitial criticality required 87 fuel elements. A TRIGA element does not have positive reactivityworth after approximately 6 grams of 235U are burned; as a conservative measure, a maximumburnup of 10 grams is assumed in calculations.External event modes with potential challenges to each SSC are identified in the Table 3.1.Design criteria for each SSC are provided in section 3.1. Design criteria for and potential impacton required components which are vulnerable to meteorological conditions is provided insection 3.2. Designs to protect against water damage and the impact of potential flooding onstructures, system and components which are vulnerable to water intrusion effects areprovided in section 3.3. Design criteria for and potential impact on required components whichare vulnerable to seismic events is provided in section 3.4.Table 3.1, SSC VulnerabilityStructure, System, Component Potential VulnerabilityMeteorological Water SeismicFuel moderator elementsControl elementsCore structure XPool, pool cooling, pool cleanup X XBiological shieldingReactor Bay/Building X XVentilation Reactor bay vent, purge HVAC, X X XInstruments &Controls X XFacility sumps and drains X X X3.1 -Design Criteria for Structures, Systems and Components for Safe Reactor Operation3.1.1 Fuel Moderator ElementsThe TRIGA Mark II nuclear reactor was developed by the General Atomic Division of GeneralDynamics Corporation for use by universities and research institution as a general-purposeresearch and training facility. The TRIGA reactor design was based on four [interrelated]principles: safety, simplicity, utility, and cost. General Atomics developed a fuel matrixconsisting of zirconium hydride with uranium with a strong negative reactivity response totemperature used in fuel-moderator elements. Since temperature is a function of thermalpower and thermodynamic properties (including heat removal time constants), thePage 3-2 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 3temperature response is a feature that inherently limits the maximum achievable power levelsunder transient and steady state conditions. A complete description of the UT TRIGA Fuel isprovided in Chapter 5. The fuel-moderator matrix used at the UT TRIGA is enclosed in stainlesssteel cladding designed to prevent migration of fission products. The prototype TRIGA reactorattained criticality at General Atomics' John Hopkins Laboratory for Pure and Applied Sciencesin San Diego, California on May 3, 1958. The temperature response proved strong enough thatpulsing capabilities were developed, using step insertions of large amounts of reactivity throughpneumatic removal of a control rod. The performance of uranium-zirconium hydride fuel issubstantially independent of uranium content up to 45-w% uranium2, indicating uraniumloading (within a large nominal range of values) is not a design criterion.Cladding is the principal barrier to fission product release; therefore the design criteria forchemical, mechanical, and thermal conditions require fuel integrity under normal operating andpotential accident scenarios. Chemical degradation is limited by establishing a design basis forpool water quality that minimizes corrosion. Mechanical degradation from internal sources islimited by establishing a basis for acceptable morphology and the maximum acceptable internalpressure; mechanical degradation from external sources is limited operationally. The principlecladding failure mechanism is internal pressure generated by temperature; limitingtemperatures for pressure are much less than temperatures which could degrade the fuelmatrix or cladding directly.The design criteria for TRIGA fuel is based on pressure generated in the fuel-moderatorelement. If the cladding temperature is below 500°C, internal pressure will not exceed limits oncladding yield strength at fuel matrix temperatures below 1150°C. If the cladding temperatureis greater than 500'C, yield strength of stainless steel cladding is reduced and internal pressurewill not exceed limits on cladding yield strength at fuel matrix temperatures below 9500C.3.1.2 Control RodsReactivity is regulated by control rods loaded with boron, described in Chapter 5. Reactor coremechanical design permits control rods to operate in a small set of positions. The positions ofthe control rods in the core are manipulated by a control rod drive system. The control rodsand the control rod drives maintain and control reactor power (i.e., rate of fissions) fromshutdown to full power operation, including compensation for temperature increases andfission product poison generated during reactor operation.Design criteria requires control rods have reactivity capable of establishing and maintaining safeshutdown conditions with the most reactive control rod fully withdrawn, and overcomingnegative reactivity effects associated with operations. Design criteria for the control rod drivesystems include rod speed adequate to overcome temperature and xenon effects, and fail-safeoperation.2 NUREG-1282, Safety Evaluation report on High-Uranium Content, Low-Enriched Uranium ZirconiumHydride Fuels for TRIGA Reactor (Docket No. 50-163)Page 3-3 CHAPTER 3, DESIGN OF STRUCTURES, SYSTEMS AND COMPONENTS 12/20113.1.3 Core and structural SupportThe fuel-moderator elements and control rods are positioned by an upper and lower grid plate.The grid plates establish a geometric array designed to support water moderation and heatremoval, and the lower gird plate bears the weight of the fuel-moderator elements. Graphiteintegral to the elements and a separate, external graphite cylinder surrounding the grid platesreduce neutron leakage. A solid plate directly under the core limits control rod movementdown' from the fully inserted, preventing potential for the control rod falling out of the core.The reflector assembly rests on a rectangular core support platform fabricated from weldedstructural aluminum beams. The core support platform is welded to the reactor pool floor.Details of the reflector and core assembly are found in Chapter 5.Design criteria for the reflector and core array assembly includes mechanical support (stability,strength, and position) as well as cooling and neutronic geometry that assures safe operationsand adequate response to accident conditions (adequate cooling, maintenance of shutdownreactivity). Reactor cooling is analyzed in Chapter 5 for normal operations, and Chapter 13 foraccident scenarios.3.1.4 Pool and Pool Support SystemsThe reactor core operates by design near the bottom of a large pool of water (Chapter 5). Poolwater provides passive cooling for heat removal from the core, moderation of fission energyneutrons required to achieve criticality, and shielding from radiation (produced from the fissionprocess and materials neutron-activated in the core region). The amount of heat produced atthe rate of fission at operations below a few kW thermal-powers (and following shutdown) isadequately controlled by convection to pool water, with the heat removed from the pool waterby evaporation and conduction to the biological shield. Steady state operation at higher powerlevels requires active measures to control pool water temperature. A pool cooling system(Chapter 4, 5) is installed to remove heat from the pool water. A pool cleanup system assuresthe pool water chemistry does not degrade fuel elements.Design criteria for rector pool includes a depth of water to reduce radiation exposure toacceptable levels, (in conjunction with core cooling geometry) heat transfer characteristicsadequate to control pool water temperature during normal and accident conditions. Thedesign criterion for the pool cooling system requires the water temperature can be controlledduring operations, with potential for losing pool water inventory in a failure mode controlled.The design criterion for the pool cleanup system is that the water quality can be controlled toacceptable levels.3.1.5 Biological ShieldingThe reactor pool is surrounded by a large concrete biological shield (Chapter 5, 11). Theshielding design controls radiation hazard from the fission process (and activated materials).Access to high radiation fields is provided to support experimental programs with beam tubes(Chapter 5, 10) that penetrate the biological shielding. Internal shielding plugs control thePage 3-4 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 3 1...hazard when the beam ports are not in use, active measures provided by experiment controls(Chapter 10) compensate for the increased hazard during utilization.Design criteria for reactor biological shielding is control of area radiation levels to less than 1mrem/h.3.1.6 NETL Building/Reactor BayEngineering design, specifications, and construction for the building meet the State of TexasUniform General Conditions and The University of Texas at Austin Supplementing Conditions3.Provisions of the Uniform Building Code4 and other national codes for mechanical, electrical, andplumbing are applicable to this project. Equipment requirements will apply Underwriter'sLaboratories standards or labels, when appropriate, to a piece, type, class, or group ofequipment. Other specifications will conform to the standards of the American Society forTesting and Materials (ASTM). Provisions of the Life Safety Code are applicable. One code ofimportance, the National Fire Protection Code, will determine requirements that relate to firesafety for significant facility operation hazards.The building site is located on Soil tests of the subsurfaceset the load capacity . . Seismic design specifications are Uniform Building Code for zone 0. Normalbuilding loads from gravity and wind forces exceed the seismic accelerations for buildings inzone 0; therefore these specifications require no special provisions beyond those of standardbuilding load requirements.Wind load designs meet requirements of the Uniform Building Code for 70 mph (31.3 m/sec)winds. The specifications include factors for gusts in excess of the wind load criteria.Normal wind and storm conditions are within these design factors.Building and site draining system design specification were commercial grade, ASTMstandards. The sub draining system (French and storm drains) construction includes agranular drainage layer crushed stone meeting ASTM C-33, Grade 67 covering excavatedrock surfaces and in the sub-daring trenches for compacted Subdrainage systems were fabricated using American Societyof Testing and Materials (ASTM) 3 [A] Specifications for Nuclear Engineering teaching laboratory, Project No. 102-568, the University ofTexas at Austin (09/15/1986)[B] Construction Administration Manual for Nuclear Engineering teaching laboratory,, Project No. 102-568, the University of Texas at Austin (12/1986)[C] NETL Project Nos. 1, 2, & #, Project No. 102-568, Amendments; the University of Texas at Austin(12/1986)4 Uniform Building Code, International Conference of Building Officials (05/01/1985)Page 3-5 CHAPTER 3, DESIGN OF STRUCTURES, SYSTEMS AND COMPONENTS 12/2011and Vent Pipe and Fittings, and appropriate standards for joining A. BuildingArchitectural design of the building will develop two separate functional sections, the reactorbay wing and an academic and laboratory wing. Structural design of the building sections is ofconcrete columns and beams with steel reinforcement. Two floor levels will comprise theacademic and laboratory wing. The first level of the reactor bay wing is below the mean grade, while the academic wing entry level is 7 feet (2.1 meters) above themean grade.The entry floor level (second level) is an administrative and office section. Laboratories will beon the next level (third level). Construction of this wing is reinforced concrete pier andcolumns with poured beam and slab floors and roof. Exterior walls will consist of concretetilt panel, metal siding and window units. Interior walls are metal stud frames with gypsumboard panels. Doors are solid core wood. Entry way area and door is glass and metal frame.Stairwells at each end of the building wing will provide access to each building level.The reactor bay wing consists of three basic parts with several types of concreteconstruction. A 4-levelsection with the HVAC room, control room & offices, shops and facility service/equipmentrooms, and staging area are in a section adjacent to the reactor bay. A radiation experimentroom with is adjacent to the 4-level section. Exteriorwalls of the reactor bay are concrete and steel construction with tilt panels and attachmentcolumns. The combination of panels and columns set on top of the first level structure formsan integral unit by placement of the panels, then placement of the columns.Structural concrete and steel columns support slab and beam floors adjacent to the reactorbay. Interior walls are primarily concrete blocks with a few plaster board type walls. Theexterior construction of the reactor bay wing is completed by concrete and metal panels. Roofstructure is a steel joist system with metal deck, concrete slab, and built-up composition roofthat includes fire barrier and thermal insulation.A room of four walls and a roof of standard density concrete thick forms aradiation shield room to complete the reactor bay wing. The room is cast in place with keyjoints between concrete placements. Tilt panels and composition roof finish thestructure. All doors are of hollow metal construction.Page 3-6 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 3B. Reactor BayDesign of the reactor bay is specified by constraints on the function of the architecturedesign, access control for physical security, radiation protection for personnel safety, andapplicable building code standards.The reactor pool, shield and primary experiment facilities are located in a reactor bay areathat is about 18.3 meters on each side. A total of 4575 cubic meters of volume is enclosed inthe reactor bay above the 335 square meters of floor space. Operation control of reactor andof reactor experiment activities is provided by an area located adjacent to the reactor bay. Spacein the operation control area is divided into control room, conference room, office, and entryway. Total operation control area (7.3 by 18.3 m) is 134 square meter of floor space androughly 489 cubic meters of air space. The stairwell in the academic wing provides access tothe reactor bay and operation control areas. The remaining three sides of the reactor bay areaare enclosed by exterior walls. Both emergency exits and equipment bay doors on the firstlevel open into the adjacent area within the building from which building exits are accessible.Two rooms within the reactor bay will enclose reactor support systems. Pool water treatmentsystems for purification and cooling equipment are on the first level. Auxiliary equipment forexperiment systems, such as pneumatic systems, will be in the second level room. Otherfeatures of the reactor bay include a five-ton bridge crane and fuel storage pits. Thestorage pits and reactor shield structure are important systems to safely operate and storethe reactor fuel materials. However, only the ventilation design for the reactor bay is anengineering safety feature.3.1.7 Ventilation SystemsVentilation systems are provide to support general habitability, with two dedicated systemsdesigned to control the buildup of radioactive gas in the reactor bay (the confinementventilation system and the auxiliary purge system).Design criteria for the ventilation systems are to control radiation exposure from airborneradionuclides to within acceptable limits during normal operations, and to prevent reactor bayventilation systems from discharging unacceptable levels of radioactive effluent during accidentconditions. A secondary function of the system is to conserve energy required to condition theair when the reactor is not operating.Page 3-7 CHAPTER 3, DESIGN OF STRUCTURES, SYSTEMS AND COMPONENTS 12/2011A control system establishes and manages of differential pressures across spaces to maintain agradient that manages air flow. The control system is designed to ensure that any potentialreleases of radioactive materials is directed through a controlled discharge path (Chapter 9).The reactor bay ventilation system provides fresh air into and an exhaust stream from thereactor bay (Chapter 9, 11). This system has an operational mode that recirculates air if thereactor is not operating to reduce the energy consumed in conditioning the air.The auxiliary purge system exhausts atmosphere from experimental facilities, where gaseousactivation products are expected to occur (Chapter 9, 11).Effluent pathways for air, liquid, or solid releases of radioactive material provide control ofmaterial releases. Control pathways for air and liquid effluents are by way of two rooms,room 4.1M3 and room 1.108. Control of air releases from reactor experiment areas isprovided in room 4.1M 3, which contains the air, purge system isolation valve and filter bank.The filter bank normally contains prefilters and one high efficiency particulate filter. The filterbank is configured to accommodate a charcoal filter and additional high efficiency particulatefilters, if needed.There are two principle gases radionuclides produced as a byproduct of reactor operations inquantities of concern. Production of these radionuclides is addressed in Chapter 10.3.1.8 Instruments and ControlsReactor instrumentation and controls (including safety system, reactivity control systems, andprocess, radiation monitoring systems, and process monitoring systems) are designed to beoperated and monitored from a central control room.The design basis for the safety systems is to automatically terminate operations before a safetylimit can be exceeded. The design basis for the reactor controls system is to permit reactivitycontrol to (1) maintain safe shutdown under all license conditions, and (2) compensate fortransient changes in temperature and xenon over the full range of power operations.3.1.9 Sumps and DrainsControl of liquid releases that contain radioactive material is provided in room 1.108, whichcontains storage tanks for collection, processing, storage, or release of liquid effluents. Thereactor pool will not release liquid effluents as a part of normal operation.Design for water runoff in the project vicinity will provide for dispersal of water from localrainfall rates that are frequently sporadic but sometimes torrential. Drainage provisions forthe building roof, site landscape, access roadways and subsurface control local runoff. Local floodcontrol includes gravity flow drainage and collection sumps with dual operation pumps. Roofdrainage and site runoff are by gravity flow. Separate sumps with pumps control subsurfacedrainage at the building perimeter and beneath the reactor shield foundation.Page 3-8 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 33.2 Meteorological DamageNormal wind and storm conditions are within the design factors established in UniformBuilding Code for 70 mph (31.3 m/sec). Hurricanes are not likely to be a direct threat becauseof the natural dissipation of energy on land. However, tornados are a concern with their extremewind velocities. Tornado type activity is roughly one event per year per 1000 square miles (2590sq. kilometers) in the general site area. This activity represents a frequency of one per 2.5 x 105years for an area of a square with sides of 333 feet (31 meters) representative of the building.3.3 Water DamageGentle slope characteristics in the immediate site vicinity provide an ample gradient of about 3feet (1 meter) for surface water runoff. A concrete spillway has been constructed toassure drainoff does not concentrate. Mean elevation at the local site is 791 feet (241meters). Data from the National Flood Insurance Program indicates that no portion of theresearch campus site is within the 100 or 500 year flood zone. Thus, the only flooding likely willbe as a result of local runoff conditions.The facility has three collection sumps. One sump collects water from the radioactive wastecollection system which serves the radioactive labs in the laboratory and office wing, and doesnot play a role in protection form water intrusion. One sump collects water from French drainsinstalled around the reactor biological shielding/pool foundation. One sump collects water fromthe truck access ramp and French drains around the building foundations.Equipment providing services to reactor systems is located in two rooms on the lower level ofthe reactor building. Makeup water, compressed air, and HVAC chill water are provided from areactor building lower level room adjacent to the reactor bay. Pool cooling and cleanup arelocated in a room within the reactor bay structure.Makeup water is provided by potable water pressure. Service would still be available ifthe makeup water system were flooded, although water quality could not be monitored. Theloss of chill water to fan coil units affect habitability only. The ventilation system dampercontrols, pool cooling controls, and pulse rod operate using compressed air system. Thecompressors and air dryer would likely fail if the air compressor room were flooded. The pulserod would be inoperable with the control rod fully inserted in a safe condition. Pool coolingwould be inoperable. Reactor bay air dampers would fail closed. These systems are notrequired to maintain safe shutdown conditions, but the ventilation is required for reactoroperation.Pool cooling and cleanup pumps could be damaged or rendered inoperable by waterintrusion; however, the pool cleanup pump is not required for operation unless chemistryPage 3-9 CHAPTER 3, DESIGN OF STRUCTURES, SYSTEMS AND COMPONENTS 1 011control is required to maintain pH at acceptable levels, and the pool cooling pump is notrequired for operations as long as temperatures are acceptable (or operating at less than about100 kW) or while shutdown. The loss of pool cooling would affect the range of possibleoperations, but not reactor safety.In summary, massive water intrusion on the first floor could affect operability of thereactor but would not prevent maintenance of safe shutdown conditions.3.4 Seismic DamageThe potential for seismic damage is evaluated in three areas, (A) core and structural support,(B) pool and pool cooling, and (3) the building.A. Core and structural SupportGiven (1) the height of the reflector surrounded by a pool of water, (2) the distributed weight ofthe radial reflector around the core, and (3) the potential motion of fuel elements, hypotheticalseismic event is not likely to create any significant acceleration that would not be absorbed bythe pool water and/or mitigated by movement of the fuel elements followed by automatic re-centering of the elements in the lower gird plate. NUREG/CR-2387 (PNL-4028) analysisindicates that any disruption of the lattice by mechanical rearrangement would result innegative reactivity, increasing shutdown margin for a seismic event that dislocates, shifts, orotherwise moves fuel elements within the coreB. Pool and pool coolingAn aluminum liner is installed to provide integrity for the reactor pool. Beam ports penetratethe pool wall. However incredible, an earthquake has the potential to cause a loss of poolintegrity and therefore is postulated for analysis as a loss of cooling accident. Theconsequences of a loss of cooling accident are addressed in Chapter 13.C. BuildingA building of good construction should withstand an earthquake acceleration of about0.75 g. Ground accelerations that exceed this would be rare events in a region in whichearthquakes are already infrequent.Page 3-10 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 44.0 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: REACTOR12/20114.2.1 Reactor FuelThe TRIGA fuel system was developed around the concept of inherent safety, with fuel andcladding designed to withstand all credible environmental 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 of 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. Two 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 1.5, has a lower temperature limit than fuel with ahigher H/Zr ratio. Fig. 4.1 indicates that the higher hydride compositions are single phase andare not subject to the large volume changes associated with the phase transformations atapproximately 530°C in the lower hydrides. Also, it has been noted1 that the higher hydrideslack any significant thermal diffusion of hydrogen. These two facts preclude concomitantvolume changes. The important properties of delta phase U-ZrH are given in Table 4.1.1 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 4I12/2011Graphite 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 Table 4.1, TRIGA Fuel PropertiesPropertyDimensionsOutside 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 cm1 Kz)Heat capacity [T >0°C] (J cm-3 K')Mechanical properties of delta phase U-ZrHWElastic modulus at 200CElastic modulus at 6500CUltimate tensile strength (to 650°C)Compressive strength (20'C)Compressive yield (20'C)Mark IIIA (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 castin controlled atmosphere, high-temperature induction furnace.2 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 Harbonnier, Jean-Claude Ottone, CERCA,Proceedings of the 1997 TRTR Annual meetingPage 4-3 CHAPTER 4: REACTOR .12/2011fabricated 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 followers) respectively.A (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 betweenZrH1.64 and ZrH1.74 at room temperature, and closes at ZrH1.7 at 4550C. From 4550C 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 in determining physical 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 9557F (535°C) with large density 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 minimal in higher ratios aswell, minimizing potential for deformation from evolution 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 930°F (500°C).Table 4.2, Physical Properties of High-Hydrogen U-ZrHProperty Temperature Value UnitsThermal Conductivity 93°C -650°C 0.22 W cmloKz20°C 9.1x106 psiElastic Modulus 650°C 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 kcarg-mol1Page 4-4 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 4At 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 (0°C) to the maximum on the 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. Under non-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:log P = K1 + K-TWith:P = pressure (atm)T= temperature (K)K1= -3.8415 + 38.6433.X -34.2639_X2 + 9.28212.X3K2= -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 about1400°F (3000C). 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 change for further increases.Massively hydrided bulk density is reported to be about 2% lower than x-ray diffractionanalysis. For TRIGA fuel with a Zr:H ratio of 1:1.6, the uranium density, volume fraction, andweight fraction are related by:wUPU (A) Wu0.177-0.125. 'UandWU =0.177.pu(A)1 +0.125.pu(A)Page 4-5 CHAPTER 4: REACTORI12/2011CHAPTER 4: REACTOR 12/2011pu (A) = 19.07. V7)(A)wherepu(A) = Uranium densityWU = Uranium weight fractionVi = volume fraction of uranium in the U-ZrHI.6alloyThermal conductivity has been 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 cal-[ s-1 cm °C -1Volumetric specific heat is a function of temperature and composition. Table 4.3 lists values forvariations in Zr:H and w% U based on a 0°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 + 4.17x103 W .s .cm3U-ZrHl.720 2.17 + 4.36x103W .s .cm3U)00 02I I04 06 0B LO 12HYDFEXN C0OJTENT (H/Zt)Figure 4.1: H/Zr Phase DiagramI:jIILb -,m II IIL4(6Page 4-6 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 4A (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.A (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 with 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-energy 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.-14goo00 STAINLESS STEEL CLAD,0 -12 8.5 WT-% U-ZrHIo60 COREao _S-10060 20-S-400.01 0. 140NEUTRON ENERGY leVI -2 -0 200 400 600 800 1000 /200TEMPERATURE (EC)Figure 4.2A, Zr-H Transport Cross Section & TRIGA Figure 4.2B, Fuel Temperature Coefficient ofPage 4-7 CHAPTER 4: REACTOR 112/2011Thermal Neutron Spectra ReactivityA (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 inan increase in the partial pressure of hydrogen by as much as a factor of four.3The 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 4x10s to 2x10-8 cm2 s-1, and requiring days toequilibrate) lags the time constant 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 promotes outgassing. Surface cooling from endothermic gas emissionlowers the surface temperature and therefore tends to lower the diffusion constant at the fuelrod surfaces. Re-absorption occurs where hydride surfaces are at relatively lowertemperatures. 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 of HydrogenRedistribution in Zirconium Hydride under Temperature Gradient4 Congreso Internacional de Metalugia y Materiales, Primeras Jornadas Internacionales de MaterialesNucleares (19 al 23 de Octubre de 2009, Buenos Airesm Argentina; Some Peculiarities of HydrogenBehavior and Delayed Hydride Cracking in Zirconium Based Reactor Alloys, Shmakov, R.N. SinghPage 4-8 THE UNIVERSITY OF TEXAS TRIGA 11 RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 413 /02//ZrrH-~/~//7//7/7/IH IDATA FROM GA-8129AND NAA-SR-9374100 /1/600 700 000 9o0 1000 100 1200 1300TEHPERATURE (* )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 temperature-dependent fuel growth .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.A (6) Zr water reaction Among the chemical properties of U-ZrH and ZrH, the reaction rate ofthe hydride with water is of particular interest. Since the hydriding reaction is exothermic,water will react more readily with zirconium than with zirconium hydride systems. Zirconium isfrequently used in contact with water in reactors, and the zirconium-water reaction 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 air6.These tests have involved thequenching with water of both powders and solid specimens of U-ZrH after heating to as high as850TC, and of solid U-Zr alloy after heating to as high as 1200TC. Tests have also been made todetermine the extent to which fission products are removed from the surfaces of the fuelelements 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.A (7) Mechanical Effects At room temperature the hydride is like ceramic and shows little ductility.However, at the elevated temperatures of interest for pulsing, the material is found to be more ductile.5 General Atomics Technical Report E-117-8336 NUREG/CR-2387 Credible Accidents for TRIGA and TRIGA Fueled Reactors, S. C. Hawley,S. C. andKathren, R. L., PNL-4208 (1982)Page 4-9 CHAPTER 4: REACTOR .12/2011The effect of very large thermal stress on hydride fuel bodies has been observed in hot cell observationsto cause relatively widely spaced cracks which tend to be either radial or normal to the central axis anddo not interfere with radial heat flow. Since the segments tend to be orthogonal, their relative positionsappear to be quite stable. During fabrication, a molybdenum disk is placed between the lowest fuelmass and the lower axial-graphite reflector, minimizing potential for interaction that might affect thegraphite and cause position changes in fuel meat that has developed surface imperfections. Anticipatedmechanical effects from operation of the reactor are not expected to create conditions that challengefuel performance.A (8) Fission Product Release Early in development of U-ZrHx fuel, experiments wereperformed7 to determine the potential of the evolution of fission products from the fuel matrix.Zr-U-H alloy foils were irradiated in a materials test reactor and a post irradiation testconducted, with water flowing across the surface of the foil to remove fission products foranalysis. The test was performed for 1 day and for 8 days with the total fractional fissionproduct loss calculated to be between 10-7 and 10-5 from preferential leaching of radionuclides,with gasses evolving from depths of 2.6 iVm in the foil, and particulate from 22 A. Acceptable8upper values for release fraction are 1.0 x 10-4 for noble gases and iodine contained within thefuel, and of 1.0 x 10-6 for particulates (radionuclides other than noble gases and iodine).Experiments by General Atomics [Simnad et al., 1976] indicate a value of 1.5 x 10-5 for noblegases, which is in SARs 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:oE = P r/twhere:oe is the hoop stressP is internal pressurer is inside radiust is the wall thicknessGeneral 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 4I12/2011For stress is times the internal pressure. Figure4.4A provides temperature dependent ultimate strength and the 0.2% yield, and Figure 4.4Bshows where the hoop stress induced by the internal pressure intersects with ultimatestrength. This intersection corresponds to a fuel temperature of 9500C for claddingtemperatures greater than 500TC.Therefore, if fuel and cladding temperature remains below 9500C with cladding temperaturesgreater than 5000C, the stainless steel cladding will not fail from overpressure. For claddingtemperatures less than 5000C, hydrogen pressure from peak fuel temperature of 11500C wouldnot produce a stress in the clad in excess of its ultimate strength. The limiting fuel temperatureand pressure is therefore the design basis for the UT TRIGA fuel. TRIGA fuel with a hydrogen tozirconium ratio of at least 1.65 has been pulsed to temperatures of about 11500C withoutdamage to the clad9.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. Reabsorption takes place concurrently on thecooler hydride surfaces away from the hot spot. There is evidence for a low permeabilityoxide film on the fuel surface. Some local heat transfer does take place during the pulse timeto cause a less than adiabatic true surface temperature.9 "Annual Core Pulse Reactor," General Dynamics, General Atomics Division report GACD 6977 (Supplement 2),Dee. J. B., et. Al.Page 4-11 CHAPTER 4: REACTORI12/2011105U,C-U,LI,LaJI-U,10o4103 0.400500600 700 800 9001000 1100TEMPERATURE (-C)Figure 4.4A, Temperature and Cladding Strength for 0.2% YieldPage 4-12 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 412/2011Soo gooTEMPERATURE (*C)Figure 4.41, Temperature, Cladding Strength, and Stress4.2.2 Control Rods and Drive MechanismsThe control rods and drive mechanisms consist of (A) control rods, (B) standard control roddrives, (C) transient rod drives, (D) control functions, and (E) system operation. The UT TRIGAreactor was installed with 4 control rods, three standard rods magnetically coupled to thecontrol rod drive, and one pulse rod pneumatically coupled to the control rod drive. One of thestandard 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 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-ZrH1.6fuel, is immediately below the neutron absorberPage 4-13 CHAPTER 4: REACTORI12/2011of 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.The 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.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 cmWall 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-ZrHl.6OD 1.25 in. 3.18 cm 1.31 in 3.34 cmLength 20.88 in. 53.02 cmControl 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 PositionsPage 4-14 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 4I12/2011Control 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.Table 4.5, Control Rod InformationRod Location Diameter Estimated (1991) Current (2011)In. cm. % Ak/k $ $Transient 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-ZrHAirV2HiiLVF:16.5 cm38.1 cm38.1 cm1.6.5 cmFigure 4.6, Standard Control Rod ConfigurationA 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.Page 4-15 CHAPTER 4: REACTOR 12/2011The 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.The 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 be 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. The pinion gearengages a rack attached to the magnet draw tube. An electromagnet, attached to the lowerend of the draw tube, engages an iron armature. The armature is screwed and pinned into theupper end of a connecting rod that terminates at its lower end in the control rod. When thestepping motor is energized (via the rod control UP/DOWN switch on the reactor controlconsole), the pinion gear shaft rotates, thus raising the magnet draw tube, the armature andthe connecting rod will raise with the draw tube so that the control rod is withdrawn from thereactor core. In the event of a reactor scram, the magnet is de-energized and the armature willbe released. The connecting rod, the piston, and the control rod will then drop, thus reinsertingthe 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. DuringPage 4-16 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 4steady 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 (1.19cm/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.The 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 on 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 leverextending 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 full out position in less than I second.Page 4-17 CHAPTER 4: REACTORI12/2011D. 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 Ps -1 910 NPPPulse -1910 NPPPower Channel High power 110% 110%Detector High Voltage 80% 80%High Fuel Temperature 5500CMagnet 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.INTERLOCKSource InterlockPulse Rod InterlockMultiple RodWithdrawalPulse Mode InterlockPulse-Power Interlock2 cpsPulse roWithdrathan 1 rMode s'10 kWTable 4.7, Summary of Control Rod InterlocksSETPOINT FUNCTION/PURPOSEInhibit standard rod motion if nuclear instrumentstartup channel reading is less than instrumentsensitivity/ensure nuclear instrument startup channelis operatingPrevent applying power to pulse rod unless rodinserted/prevent inadvertent pulsePrevent withdrawal of more than I rod/Limitw snmaximum reactivity addition rate (does not apply in*odautomatic flux control)witch in Hi Pulse Prevent withdrawing standard control rods in pulsemodePrevent 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 safetysettings, 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:Page 4-18 THE UNIVERSITY OF TEXAS TRIGA 11 RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 41) 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 48.90C.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 assembly. The reflector assembly supports(A) an upper grid plate, (B) core barrel and reflector, and (C) lower grid plate, shown in Fig.4.7a/b. The upper and lower grid plates provide alignment and support for the fuel elements.-4 ...... I..--.I .........i" iFigure 4.7a, UT TRIGA Core Figure 4.7b, Core Top ViewPage 4-19 CHAPTER 4: REACTOR 12/2011A. 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 obtainflux 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.Table 4.8, Upper Grid Plate PenetrationsPenetration Function SizeFuel Elements 1.505 in. (3.8227 cm) diameter3-element 1.2 in. (3.048 cm) radius6/7-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) diameterThe 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.Fuel positions are indexed by letters denoting a "ring" where elements are collinear withrespect to the adjacent radial array fuel positions; A is the central ring.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). 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).Page 4-20 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 4I12/2011In 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 B. 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.B (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 thePage 4-21 CHAPTER 4: REACTORI12/2011lift 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. 4.9b),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 4.9a, Reflector Top AssemblyFigure 4.9b, 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. 4.10a) and a 10 in. (25.3 cm.) cylinder cut fromthe inner surface to allow a 3 inch wide experimental facility surrounding the core.0Figure 4.10a, Graphite Reflector, Through PortFigure 4.10b, Graphite Reflector Through port DetailFigure 4.10c, Graphite Reflector, Radial & Piercing-Beam PortsThe through port has a rectangular water-filled cut-out between the core shroud and the beamport penetration (Fig. 4.10b). Aluminum canisters that mate with the beam ports are nested inthe reflector in two of the beam ports, one radial and one tangential (Fig. 4.10c, Fig. 4.11a/b).The third beam port (radial) penetrates the core shroud (Fig. 4.11c).Page 4-22 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 4I12/2011Figure 4.11a, Tangential Beam Port InsertFigure 4.11b, Radial Beam Port insertrigure *.-.L., inner znroua ýurraceB (2) Graphite Rods. Graphite dummy elements may be used to fill grid positions not filledby the fuel-moderator elements or other core compounds. They are of the same generaldimensions and construction as the fuel-moderator elements, but are filled entirely withgraphite and are clad with aluminum.B (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. 4.12) provides alignment for fuel elements and control rods, and (inconjunction with the top fuel assembly) space for cooling flow. The lower (or bottom) gridplate is fabricated from a circular aluminum plate 1.75 inches (3.81 cm.), anodized to resistwear and corrosion. The top of the bottom grid plate is 9.9 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 the same center as the upper grid plate, but also containspenetrations that support alignment of the 3, 6, and 7 element facilities (Table 4.10). All but 11fuel penetrations in the lower grid plate are smaller than the diameter of the fuel element andchamfered to provide a surface supporting triflutes on the bottom of the fuel elementelements.Page 4-23 CHAPTER 4: REAUCIRI12/2011CHAPTER 4: REACTOR 12/2011Table 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 PlateReflector Canister Bottom View Grid Plate in Core ShroudFigure 4.12, 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 rodsare 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. 4.13).Figure 4.13, Fuel Element AdapterPage 4-24 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 44.2.4 Neutron Startup SourceThe reactor license permits the use of sealed neutron sources, including a 6is 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 structureThe core support structure includes (A) a platform supporting the reflector and core structure,and (B) a "safety plate" that prevents the control rods in a failure mode from falling out of thecore.A. Core Support PlatformThe reflector assembly rests on a platform (Fig. 4.14) 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 (Fig. 4.14a/b/c).Aluminum 6061-T651 plate is used for safety plate support pads (% in., 1.905 cm), cross braces(% in., 0.953 cm.), and platform support pads (Y2 in., 1.27 cm.). Angle aluminum is inserted 9in. (22.86 cm) from two edges to support the safety plate, with angle bracing on the edgesperpendicular to the safety plate supports.( 77-- --r -jCore Support Top View Core Support Side View Core Support Side ViewFigure 4.14, Core Support ViewsThe platform top surface is 30 Y4 in. X 30 Y4 in., with the top surface 16 /4 in. above the poolfloor. Surfaces are matte finished for uniform appearance with shot cleaning and peeningusing glass beads (MIL-STD -852).Page 4-25 CHAPTER 4: REACTORI12/2011Core and Support Structure AssemblyCore and Support Assembly IsometricFigure 4.15, Core and Support Structure ViewsB. Safety plateThe safety plate (Fig 4.16) 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 plate/2 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.4.__Da40Figure 4.16, 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-Page 4-26 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 4 1cylinders with a radius of 6 /2 feet separated by 6 Y2 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 m3 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)Volume (maximum) 11000 gal (41.64 Mi)Pipes Aluminum 6061Iron-Plastic Liner, 316 SSCoolant Lines ValvesBalndSeBall and StemFittings Aluminum (Victaulic)Type CentrifugalCoolant Pump Material Stainless SteelCapacity 250 gpm (15.8 Ips)Type Shell & TubeMaterials (shell) Carbon steelMaterials (tubes) 304 stainless steelHeat Exchanger Heat Duty Flow Rate (shell) Flow Rate (tubes) Tube Inlet Tube Outlet Typical Heat Exchanger Operating Parameters Shell Inlet48FShell Outlet The pool (Figs. 4.17a/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 Y in. (square) aluminumchannel was rolled and welded to the upper edge of the tank.Beam port penetrations are fabricated around the core to allow extraction of radiationbeams 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 sectionPage 4-27 CHAPTER 4: REACTOR12/2011CHAPTER 4: REACTOR 12/2011extending 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 / 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 Y2 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 ig through the core center, rotated 600 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 outerdiameter 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.Page 4-28 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 41. 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 asump, 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. 4.18) 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. Doselevels 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 4-29 CHAPTER 4: REACTOR12/2011A thick foundation pad supports the reactor pool and shield structure.Standard weight concrete, comprises the foundation pad. High densityconcrete, Five beam tubes at the 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.Parameter of InterestCONCRETE PADFLOORSAFETY PLATEGRID PLATECORE BOTTOMBEAM PORT CLCORE CLCORE TOPGRID PLATEMAIN LOWER SHIELDINGTRANSITIONAL CONCRETESHIFT TO HIGH DENSITY CiMIN CORE LEVEL (TS)VACUUM BREAKERSLOW POOL LEVEL SCRAMLOW POOL LEVELLOW POOL LEVEL ALARMNORMAL POOL LEVELHIGH POOL LEVELHIGH POOOL LEVEL TOP OF TOP LEVELTable 4.12, Significant Shielding and Pool LevelsLevel Notes(meters)pSTEP ONCRETE A ool4.5 Nuclear DesignThe characteristics and operating parameters of this reactor 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 samePage 4-30 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 4core 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. (0.381 m). Thecylinder radius is calculated as the average radius of a 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 4.16). 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 Ell, F13, and F14). Change in core configuration requirevalidation that control rod worth is not affected by the experiment facility, or re-establishmentof the control rod worth followed by verification that the limiting conditions for operation aremet.Table 4.13 Control Rod WorthControl Rod Reference Current (2011)Position Worth Position WorthTransient 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 $3.074.5.2 Nominal Reactivity Worth ValuesReactivity values for core components based on calculations and observations are provided inTable 4.14, with Technical Specifications values in bold face type. Current values are based onmeasurements; nominal values are calculations frOm indicated sources.Table 4.14, Reactivity Values$ Parameter TS CURRENT NOMINALLIMIT VALUE VALUE 0'0 Reactor Reference Data Notebook, Safety Analysis report Table 4-5; SAR Table 4-6 indicates CT Fuel $0.90, CTVoid -$0.15, PNT Void -$0. 10, RSR void -0.20Page 4-31 CHAPTER 4: REACTORI12/2011Table 4.14, Reactivity ValuesTC 1I D D IlT DIl K InKAIIAI$ Parameter I J .A.JI~r~LI'd I I'~.JIVIII'IIiLLIMIT VALUE VALUEUUUU UUUU UU UU UUUUUUUUUUUUUUUUMMME04.5.3 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. 4.19) 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.1 3-Element Experiment Authorization12 Significant deviation from values in 3-Element Experiment Authorization (cf. E-Ring -$0.50 & D-Ring $0.95)Page 4-32 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 412/2011Figure 4.19, Reactivity Loss with PowerA. Reference CalculationsA reference calculation of neutron flux distribution across the core was performed by GeneralAtomics13.The calculations were accomplished on an IBM-7090 using General Atomics(diffusion theory based) codes GAMBLE and GAZE, and GAM-I. GAM-I 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 ModelVolumeRadius Area Volume FratioFractionCell Region in. cm cm2 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 1.000013 GA-4361, Calculated Fluxes and Cross Sections for the TRIGA Reactors, G. B. West. August 1963Page 4-33 CHAPTER 4: REACTOR12/2011CHAPTER 4: REACTOR 12/2011Table 4.16, Selected TRIGA II Nuclear PropertiesNumber of cells 80 91Fuel Temperature 23°C 200°C1 eV to 10 MeV1a 0.00660 0.006757f 0.00135 0.00135Flux/watt 2.46x107 2.21x107p[1] 0.9405 0.94810 to 1 eV1a 0.0873 0.0794if 0.0526 0.0472Flux/watt 1.11x107 1.08x107% of fissions 94.6 94.5Vave cm/s 2.73x10' 2.94x10'Eave eV 0.0391 0.0455NOTE 1: Resonance escape probabilityTable 4.17, UT TRIGA DataCore ConfigurationRef Cold Clean Critical Loading Ref Operational Loading Actual Initial CriticalityFuel element pitch Coolant volume to cell ratio Fuel ElementsCladding SS304Fuel matrix U-ZrHl.6Fuel Mass Uranium fraction Enrichment Nuclear ParametersPrompt neutron lifetime ( e ) 41 pIsEffective delayed neutron 0.0070fraction (0)Prompt negative temperature 1x1O-4 Ak/k*Ccoefficient (a)B. Prompt Negative Temperature CoefficientGA 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 thePage 4-34 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 4temperature 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 neutron in the fuel element will gain energy from an excited state of an oscillatinghydrogen atom in the lattice. As the neutrons gain energy 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 spectrum to harden 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.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-314, GAZE-2's, 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 the14 General Atomics Report GA-7157, "Users and Programmer Manual for the GGC-3 Multigroup CrossSection Code," General Dynamics, General Atomic Division (1967)15 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 Atomics 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)Page 4-35 CHAPTER 4: REACTOR 12/2011resonance 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 Technologies2 , 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).Qualitatively, 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 -0.14 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 the transition 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.'8 "THERIMIDOR- A FORTRAN II Code for Calculating the Nelkin Scattering Kernel for Bound Hydrogen (Amodification of Robespierre),"Gulf General Atomic, Inc. (unpublished data) Brown, H. D., Jr.'9 "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 4-36 THE UNIVERSITY OF TEXAS TRIGA 11 RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 4 "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 S4 and S8 results tobe nearly identical. Group dependent disadvantage factors defined as Ogr/ cDgc (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.d. The cross sections for an equivalent homogenized core were used in a full reactorcalculation to determine the contribution to the temperature coefficient due to theincreased leakage of thermal neutrons into the reflector with increasing hydridetemperature. This calculation requires several thermal groups, but transport effects areno longer of major concern. Thus, reactivity calculations as a function of fuel 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. Figure 4.19 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 spatial distributionof 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.Page 4-37 CHAPTER 4: REACTOR 12/2011A. Core Peaking FactorsThe core is generally modeled as a right 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 not lend 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) isgiven22 by:1.202*(hRF, 2.4048However, TRIGA fuel elements are on the order of a mean free path of thermal neutrons, andthere is a significant change in thermal neutron flux across a fuel element. Calculated thermalneutron flux data23 indicates that the ratio of peak to average neutron flux (peaking factor) forTRIGA cores under a range of conditions (temperature, fuel type, water and graphite reflection)has a small range of 1.36 to 1.40. Therefore, actual power produced in the most limiting actualcase 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 iT/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*co 2j* 2*e+e. 'in which e = L / 2 and ge.Iis 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:22 Elements of Nuclear Reactor Design, 2nd Edition (1983), J. Weisman, Section 6.32' GA-4361, Calculated Fluxes and Cross Sections for TRIGA Reactors (8/14/1963), G. B. WestPage 4-38 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 412/2011Pq ' NI .DLB. Power distribution within a Fuel ElementThe radial and axial distribution of the power within a fuel element is given byq .'(r,z) = q1f(r)g(_)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:a + cr + er1 + br + dr2in which the parameters of the rational polynomial approximation are derived from flux-depression calculations for the TRIGA fuel24.Values for the coefficients are: a = 0.82446, b =-0.26315, c = -0.21869, d = -0.01726, and e = +0.04679. The fit is illustrated in Figure 4.20.1.31.21.11.0L"I I I I I I I I I I I I I I I I I I0.900.800.0 0.20 0.40 0.60 0.80 1.0 1.2 1.4 1.6 1.8 2.0r (cm)Figure 4.20, Radial Variation of Power Within a TRIGA Fuel Rod.(Data Points from Monte Carlo Calculations [Ahrens 1999a])C. Power per rodThe Bernath correlation25 calculates critical heat flux as: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)Page 4-39 CHAPTER 4: REACTOR 12/2011[A]BO -TBWhere the convection heat transfer coefficient for "burnout" condition is determined by:fDehBo = 10890o -D+/- ) + SLOPE -VWith two possible values for the "SLOPE" term:(1) IF De! 0.1 ft.,48SLOPE = 0.De.(2) IF De > 0.1 ft.:10SLOPE = 90 +And the burnout wall temperature term is calculated:P VTwBo = 57-In(P) -54 ---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 °C, D hydraulic diameter of the coolant passage in feet, D is the diameter of theeiheater surface (heated perimeter divided by T1) 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:D 48 f P V )-8 o =" D + D + ---- "D V) " 57 " In(P ) -54 p 1 5TWhere A is the flow area and WP the wetted perimeter, hydraulic diameter is calculated:25 ANL/RERTE/TM-07-01, Fundamental Approach to TRIGA Steady state Thermal-Hydraulic CHFAnalysisPage 4-40 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 44-ADe-=4*wP(1) Wetted perimeter:WP =- D f,,,,2(2) Flow area:A=PITCH2 ---f1 fu.l1TRACE calculations completed as described in section 4.6 calculation of thermal hydraulicparameters that are used to calculate critical heat flux using the Bernath correlation (and theratio of the heat flux to the critical heat flux, CHFR). TRACE calculates CHFR using the Biasicorrelation. The results of calculations using heat flux and temperature data for 490C water at6.5 m level is provided in Table 4.18. The minimum CHFR versus power level is provided inFigure 4.21. As illustrated, the CHFR values agree well and remain much greater than 2 atpower levels up to 22.5 kW per unit cell.Table 4.18, Critical Heat Flux ratio, Bernath CorrelationkW 1 2 3 4 5 6 7 8 9 10 11 12 13 14 151.5 106.2 93.5 83.3 74.9 68.6 66.1 63.7 61.5 63.5 65.7 67.2 72.5 78.9 86.8 96.83.o 61.0 53.4 47.2 42.1 38.5 37.0 35.7 34.4 35.5 36.6 37.2 39.7 42.9 46.8 51.94.5 44.3 38.6 34.0 30.2 27.6 26.4 25.5 24.4 25.3 26.0 26.2 27.8 29.7 32.2 35.56.o 35.4 30.9 26.9 23.6 21.6 20.8 20.0 19.2 19.7 20.3 20.3 21.3 23.9 24.6 26.77.5 29.9 25.8 22.5 19.7 17.9 17.2 16.5 15.8 16.3 16.8 16.6 17.4 19.2 19.6 21.39.o 26.0 22.3 19.4 16.9 15.3 14.7 14.1 13.5 13.9 14.2 14.0 14.6 16.0 16.2 17.610.5 23.1 19.8 17.1 14.8 13.4 12.9 12.4 11.8 12.1 12.4 12.1 12.5 13.6 13.8 14.812.0 20.8 17.8 15.3 13.2 12.0 11.5 11.0 10.5 10.8 11.0 10.7 10.9 11.8 11.8 12.613.5 19.1 16.3 13.9 11.9 10.8 10.3 9.9 9.4 9.7 9.9 9.5 9.7 10.4 10.3 10.915.0 17.6 15.0 12.8 10.9 9.9 9.4 9.0 8.6 8.8 9.0 8.6 8.6 9.2 9.1 9.516.5 16.4 13.9 11.8 10.0 9.1 8.6 8.2 7.8 8.0 8.2 7.8 7.8 7.9 8.0 8.318.0 15.4 13.0 11.0 9.3 8.4 8.0 7.6 7.2 7.4 7.5 7.1 7.0 7.0 7.1 7.319.5 14.5 12.2 10.3 8.6 7.8 7.4 7.1 6.7 6.8 7.0 6.5 6.4 6.4 6.4 6.521.0 14.0 11.8 9.9 8.3 7.5 7.2 6.8 6.5 6.6 6.7 6.3 6.2 6.1 6.1 6.122.5 13.7 11.5 9.7 8.1 7.3 7.0 6.6 6.3 6.4 6.6 6.1 6.0 5.9 5.9 6.0Page 4-41 CHAPTER 4: REACTORI12/2011Critical Heat Flux Ratio.24.46M -0041 -l0Bernath Correlation -BEAU--- -sorlt--16MO- -Biasi Correlation6.00M1.0M_2.50 4.00 6.50 900 22.5014.00 16.50 29.00 212-0Unit Cell Power (kW)Figure 4.21, 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 (PpEAK,ROD) is related to the core average power level(PAVE) by:PPEAKROD = PAvEN PF* NParametric 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. 4.22. 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 4-42 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 4I12/2011Limiting Core Configuration(Fuel Element Peak Power 45 kW)315W29W2653C.24W2150L9W1.37 1.44 151 125 1.65 1.72 L"79 1.86 1 2Peaking Factor-3-9M8M ENIS -gof --00 ELEMENTSFigure 4.22, 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 Appendix 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. 4.19. 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. 4.19 with an R2 value of 0.99999 is:8k = -1.75340-12.p4 + 6.06670"1O- 9p3 -8.77740"1 04.p2 +8.45380"1 O-3". 0.072937An approximation of the power coefficient of reactivity from 100 kW to 1 MW is therefore:A~ = -7.01360-12"pl +1l.82001"lO0-8.p2_1l.755488-10-6"P +8.45380-10-2dPPage 4-43 CHAPTER 4: REACTOR 12/2011Power Coefficiert of Reactivityjo m i -i : i i i i i i l f ,Figr 4.23, Poe Coficient of Recivity.... ....... .. ............. .[ ... ... ........ ... .ji ...i ... ... ... .. ..... .i ...l.!'iJJ ~~~~ ........... i i ........Power Level (kWI)Figure 4.23, 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.Note that prediction of the power coefficient of reactivity beyond the range is not supportedbecause the polynomial passes through a minimum above the maximum data point value.There appears to be a significant difference in response in the power level coefficientcomparing low power level data to high power level data.The operating limits on core reactivity are provided in Table 4.20.Page 4-44 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 4I12/2011Excess reactivityShutdown margin111Moveable experiment wcSingle experiment worthTotal experiment worthTable 4.20, Reactivity Limits% Ak/k4.90.2irth 0.71.752.10$7.000.1821.002.503.00NOTE [1]: most reactive rod fully withdrawn, moveable experiments in themost positive-reactive stateBased 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-i $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 -$2.854.6 Thermal Hydraulic DesignThis section provides an independent assessment of the expected fuel and cladding thermalconditions associated with both steady-state and pulse-mode operations with realisticmodeling of the fuel-cladding gap. Analysis is based on analysis of limiting conditions applied toa single fuel channel. The relation of the limiting channel to core average power is first definedas spatial power distribution.1Analysis of pulsed-mode behavior is provided in Appendix 4.1, revealing that film boiling is notexpected, even during or after pulsing leading to maximum adiabatic fuel temperatures.Appendix 4.1 reproduces a commonly cited analysis of TRIGA fuel and cladding temperaturesassociated with pulsing operations. The analysis addresses the case of a fuel element at anaverage temperature immediately following a pulse and estimates the cladding temperaturePage 4-45 CHAPTER 4: REACTOR 12/2011and surface heat flux as a function of time after the pulse. The analysis predicts that, if there isno gap resistance between cladding and fuel, film boiling can occur very shortly after a pulse,with cladding temperature reaching 470'C, but with stresses to the cladding well below theultimate tensile strength of the stainless steel. However, through comparisons withexperimental results, the analysis concludes that an effective gap resistance of 450 Btu hr-1 ft2OF-' (2550 W m2 Kz) is representative of standard TRIGA fuel and, with that gap resistance, filmboiling is not expected.Analysis of steady state conditions reveals maximum heat fluxes well below the critical heat fluxassociated with departure from nucleate boiling. The heat transfer model (A) is discussed,followed by (B) the results.4.7.1 Heat Transfer ModelThe overall heat transfer coefficient relating heat flux at the surface of the cladding to thedifference between the maximum fuel (centerline) temperature and the coolant temperaturecan be calculated as the sum of the temperature changes through each element from thecenterline of the fuel rod to the water coolant, where the subscripts for each of the AT'srepresent changes between bulk water temperature and cladding outer surface, (bro), changesbetween cladding outer surface and cladding inner surface (rori), cladding inner surface and fuelouter surface -gap (g), and the fuel outer surface to centerline (ricl):T = T b+ AT + ATr,. + AT9 + ATlTable 4.22: Thermodynamic ValuesParameter Symbol Value UnitsFuel conductivity kf 18 W mz K-114.9 W mz K' (300 K)Clad conductivity k9 16.6 W m1 K1 (400 K)19.8 W mz Kz (600 K)Gap resistance h9 2840 W m2 KzClad outer radius r0 0.018161 MFuel outer radius ri 0.018669 MActive fuel length Lf 0.381 MAxial peaking factor APF -,t/2 N/AA standard heat resistance model for this system is:T11q+ r° -In (+Iq"h k r,.hg 2.kPage 4-46 THE UNIVERSITY OF TEXAS TRIGA 11 RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 4 1in which r. and ri are cladding inner and outer radii, hg is the gap conductivity, h is theconvective heat transfer coefficient, and kf is the fuel thermal conductivity. The gapconductivity of 2840 W m-2 K-1 (500 Btu h-1 ft -2 OF-) is taken from Appendix A.The convective heat transfer coefficient is mode dependent and is determined in context.General Atomics reports that fuel conductivity over the range of interest has little temperaturedependence. Gap resistance has been experimentally determined as indicated in Table 4.25.Temperature change across the cladding is temperature dependent, with values quoted at 300K, 400 K and 600 K.4.7.2 ResultsTRACE was used to provide data supporting the analysis. TRACE models a unit cell which iscomposed of the area enclosed the pitch geometry. Since the UT TRIGA uses a hexagonalgeometry, the unit cell is an equilateral triangle. Three 300 segments of a fuel element fallwithin the unit cell, with calculations for heat generation corresponding to /2 of the element.For example, calculations assuming 10 kW for the unit cell give indication of thermal responseto an element output of 20 kW.The TRACE heat source was modeled as a 15 in. (38.1 cm) heat flux simulating fuel, exitingstainless steel with cladding dimensions. Heat distribution was modeled as sinusoidal variationfrom a maximum at the center to a minimum modified at the end by extrapolation length ofthermal neutrons in graphite. Data was calculated for 15 equally spaced nodes across the spanof 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). TRACE calculations provide claddingtemperatures directly, which is a significant portion of the standard heat resistance model,leaving only temperature differences across the gap and fuel matrix, reducing the model to:TC = Tcadd g +q". rO + re,r, .hg 2 k.ksConsidering that the terms in parenthesis are all constants, peak temperature at each nodeanalyzed is a function of cladding temperature and heat flux. Cladding temperature and heatflux at the surface of the cladding are both calculated directly by TRACE. Coolant temperatureis provided in Table 4.24, the heat fluxes at each of 15 nodes dividing the fuel section areprovided in Tables 4.25a/b.Table 4.24, Coolant Temperature for 49"C 6.5 m PoolUnit Cell 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15kW1.5 323 323 324 325 325 326 326 326 326 326 327 328 329 330 3313.0 323 324 326 327 327 328 328 328 329 329 330 332 333 335 336Page 4-47 CHAPTER 4: REACTORI12/2011CHAPTER 4: REACTOR 12/20114.5 323 325 327 3296.0 324 326 328 3307.5 324 326 329 3319.0 324 327 329 33210.5 324 327 330 33412.0 325 328 331 33513.5 325 328 332 33615.0 325 329 332 33716.5 325 329 333 33718.0 325 329 334 33819.5 326 330 334 33921.0 326 330 334 33922.5 326 330 335 340329 329 330 330330 331 331 332332 332 333 333333 334 334 335334 335 335 336335 336 337 337336 337 338 339337 338 339 340338 339 340 341339 340 341 342340 341 342 343340 341 342 343341 341 342 343331 331 333 335 337 338 340332 333 335 338 337 342 343334 334 337 340 340 345 347335 336 339 342 342 348 350337 337 341 344 345 350 353338 339339 340341 341342 343343 344344 345343344346347349350346 347 353 355348 350 355 358350 352 357 360352 356 360 363354 358 362 365355 360 364 367344 345 351 356 360 365 368344 345 351 356 360 365 368Table 4.25a, Heat Flux (Nodes 1-9) 49°C 6.5 Pool,Unit CellKw1.53.04.56.07.59.010.512.013.515.016.518.01 2-2.72E4 -3.06E4-5.44E4 -6.12E4-8.16E4 -9.18E4-1.09E5 -1.22E5-1.36E5 -1.53E5-1.63E5 -1.84E5-1.90E5 -2.14E5-2.18E5 -2.45E5-2.45E5 -2.75E5-2.72E5 -3.06E5-2.99E5 -3.37E5-3.26E5 -3.67E53-3.40E4-6.80E4-1.02E5-1.36E5-1.70E5-2.04E5-2.38E5-2.72E5-3.06E5-3.40E5-3.74E5-4.08E54 5-3.74E4 -4.08E4-7.48E4 -8.16E4-1.12E5 -1.22E5-1.50E5 -1.63E5-1.87E5 -2.04E5-2.24E5 -2.45E5-2.62E5 -2.86E5-2.99E5 -3.26E5-3.37E5 -3.67E5-3.74E5 -4.08E5-4.11E5 -4.49E5-4.49E5 -4.89E56-4.23E4-8.46E4-1.27E5-1.69E5-2.11E5-2.54E5-2.96E5-3.38E5-3.81E5-4.23E5-4.65E5-5.07E57-4.38E4-8.76E4-1.31E5-1.75E5-2.19E5-2.63E5-3.06E5-3.5OE5-3.94E5-4.38E5-4.82E5-5.25E58-4.53E4-9.06E4-1.36E5-1.81E5-2.27E5-2.72E5-3.17E5-3.63E5-4.08E5-4.53E5-4.99E5-5.44E59-4.38E4-8.76E4-1.31E5-1.75E5-2.19E5-2.63E5-3.06E5-3.50E5-3.94E5-4.38E5-4.82E5-5.25E5Table 4.25b, Heat Flux (Nodes 10-15) 49°C 6.5 Pool,Unit Cell 10 11 12 13 14 15Kw1.53.04.56.07.59.010.512.0-4.23E4-8.46E4-1.27E5-1.69E5-2.11E5-2.54E5-2.96E5-3.38E5-4.08E4-8.16E4-1.22E5-1.63E5-2.04E5-2.45E5-2.86E5-3.26E5-3.74E4-7.48E4-1.12E5-1.50E5-1.87E5-2.24E5-2.62E5-2.99E5-3.40E4-6.80E4-1.02E5-1.36E5-1.70E5-2.04E5-2.38E5-2.72E5-3.06E4-6.12E4-9.18E4-1.22E5-1.53E5-1.84E5-2.14E5-2.45E5-2.72E4-5.44E4-8.16E4-1.09E5-1.36E5-1.63E5-1.90E5-2.18E5Page 4-48 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 4I12/2011Table 4.25b, Heat Flux (Nodes 10-15) 49°C 6.5 Pool,Unit Cell 10 11 12 13 14 15Kw13.5 -3.81E5 -3.67E5 -3.37E5 -3.06E5 -2.75E5 -2.45E515.0 -4.23E5 -4.08E5 -3.74E5 -3.40E5 -3.06E5 -2.72E516.5 -4.65E5 -4.49E5 -4.11E5 -3.74E5 -3.37E5 -2.99E518.0 -5.07E5 -4.89E5 -4.49E5 -4.08E5 -3.67E5 -3.26E519.5 -5.50E5 -5.30E5 -4.86E5 -4.42E5 -3.98E5 -3.54E521.0 -5.92E5 -5.71E5 -5.23E5 -4.76E5 -4.28E5 -3.81E522.5 -6.34E5 -6.12E5 -5.61E5 -5.10E5 -4.59E5 -4.08E5Table 4.26, Peak Fuel Centerline Line Temperature (K) 49°C 6.5 Pool,UnitCell 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15Cell1.5 363 369 374 380 385 388 390 393 391 389 387 383 379 375 3703.0 400 410 420 431 441 445 449 454 450 446 443 434 426 418 4094.5 434 449 463 478 491 498 504 511 504 499 494 482 470 458 4466.0 467 485 505 524 541 549 556 562 556 550 543 528 512 496 4807.5 500 523 545 567 584 591 598 606 598 591 584 568 551 533 5149.0 531 559 583 603 623 631 640 648 640 631 623 603 584 565 54510.5 563 592 616 639 661 671 680 690 680 671 661 639 616 594 57112.0 595 622 350 673 699 710 721 733 721 710 698 673 648 623 59813.5 622 651 680 709 737 750 762 775 762 750 737 709 680 651 62315.0 648 680 712 743 775 789 803 816 802 789 775 743 711 680 64816.5 673 709 743 778 813 828 843 859 843 827 813 777 743 709 67318.0 698 737 775 813 850 866 883 901 883 866 850 813 775 737 69819.5 725 765 806 847 888 906 924 942 924 906 887 847 806 765 72421.0 750 793 838 881 926 945 964 984 964 945 925 881 838 793 75022.5 775 822 869 916 963 984 1005 1026 1005 984 963 916 869 822 774TRACE calculation provides thermal response for only a single unit cell which is /2 of a fuel rod,and neutron flux distribution causes power level to vary across the core. For instance, flow rateversus power is provided for each fuel element in Fig. 4.25. Core flow can be calculated bysumming the flow rates of individual fuel rods operating at specific power levels, with peakingfactors as identified above used to calculate the element power level. Unit cell temperaturesare provided in Fig. 4.24 for two unit cell power levels, 10.5 and 22.5 kW.Page 4-49 CHAPTER 4: REACTORI12/201110.5 kW Unit Cell Axial Temperature Profile... ..... ..- " .... ..... .. ......... .... .: .... .. ............ .i -...k ..625 = ....: ... ..... ...... ................. .. ... ... ,. _ _ = .575 .. .£25475&~ 425E -- -- --- -375275.... .... ..-1 3 s 7 9 U 13 15Node-- Temp --- Inne1 Cd Temp -Out, Clad Ie~n -C-olan Tem-pE1222.5 kW Unit Cell Axial Temperature Profile975875i. .. ... ........... ... .. ..... .. ...... i .............. ..77 5 . ' ' ' " ' ' "' ' ' ' " "" "" "' " .... ........ ..." " [ " ' "{ ' " ":... .. ." " "."..'" ..... ..:... ... ...... .. ." ..... -." -..: .-.. '... .r '...... .. ... .. ...675 -- -----5757 ................... ..1 3 S 7 9 11 13 15-F e n p --- I--f -ladlemp -----d ---O-te --Temp --Coolant Ten-NodeFigure 4.24, Unit Cell Temperature DistributionUnit Cell Flow Rate Verus Power0.120.11 ..... ....1_ 0.1 -0 .0 9 ........... .. ........... ..... ............... ....... ..... i...... ..0.070.040 .0 4 .. ... ....... .... ......... ..........i ........ ......... .......... ............ .. ....1.5 45 7-5 10.5 13.5 16.5 19.5 22.5Unit Cell Power (kW)Figure 4.25, Single Rod Flow Cooling Flow Rate versus Power Level 49°C 6.5 Pool,Page 4-50 UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT 12/2011APPENDIX 4.1, PULSING THERMAL RESPONSEThis discussion is reproduced from Safety Analysis Reports for the University of Texas Reactor Facility(UTA 1991) and the McClellan Nuclear Radiation Center (MNRC 1998).The following discussion relates the element clad temperature and the maximum fueltemperature during a short time after a pulse. The radial temperature distribution in the fuel elementimmediately following a pulse is very similar to the power distribution shown in Figure 4A.1. This initialsteep thermal gradient at the fuel surface results in some heat transfer during the time of the pulse sothat the true peak temperature does not quite reach the adiabatic peak temperature. A largetemperature gradient is also impressed upon the clad which can result in a high heat flux from the cladinto the water. If the heat flux is sufficiently high, film boiling may occur and form an insulating jacket ofsteam around the fuel elements permitting the clad temperature to tend to approach the fueltemperature. Evidence has been obtained experimentally which shows that film boiling has occurredoccasionally for some fuel elements in the Advanced TRIGA Prototype Reactor located at GATechnologies [Coffer 1964]. The consequence of this film boiling was discoloration of the clad surface.Thermal transient calculations were made using the RAT computer code. RAT is a 2-D transientheat transport code developed to account for fluid flow and temperature dependent materialproperties. Calculations show that if film boiling occurs after a pulse it may take place either at the timeof maximum heat flux from the clad, before the bulk temperature of the coolant has changedappreciably, or it may take place at a much later time when the bulk temperature of the coolant hasapproached the saturation temperature, resulting in a markedly reduced threshold for film boiling. Dataobtained by Johnson et al. [1961] for transient heating of ribbons in 100°F water, showed burnout fluxesof 0.9 to 2.0 Mbtu ft-2 hr1 for e-folding periods from 5 to 90 milliseconds. On the other hand, sufficientbulk heating of the coolant channel between fuel elements can take place in several tenths of a secondto lower the departure from nucleate boiling (DNB) point to approximately 0.4 Mbtu ft2 hr1.It isshown, on the basis of the following analysis, that the second mode is the most likely; i.e., when filmboiling occurs it takes place under essentially steady-state conditions at local water temperatures nearsaturation.A value for the temperature that may be reached by the clad if film boiling occurs was obtainedin the following manner. A transient thermal calculation was performed using the radial and axial powerdistributions in Figures 4A.land 4A.2, respectively, under the assumption that the thermal resistance atthe fuel-clad interface was nonexistent. A boiling heat transfer model, as shown in Figure 4A.3, wasused in order to obtain an upper limit for the clad temperature rise. The model used the data ofMcAdams [1954] for subcooled boiling and the work of Sparrow and Cess [1962] for the film boilingregime. A conservative estimate was obtained for the minimum heat flux in film boiling by using thecorrelations of Speigler et al. [1963], Zuber [1959], and Rohsenow and Choi [1961] to find the minimumtemperature point at which film boiling could occur. This calculation gave an upper limit of 760°C cladtemperature for a peak initial fuel temperature of 10000C, as shown in Figure. 4A.4. Fuel temperaturedistributions for this case are shown in Figure 4A.5 and the heat flux into the water from the clad isshown in Figure 4A.6. In this limiting case, DNB occurred only 13 milliseconds after the pulse,conservatively calculated assuming a steady-state DNB correlation. Subsequently, experimentaltransition and film boiling data were found to have been reported by Ellion [9] for water conditionssimilar to those for the TRIGA system. The Ellion data show the minimum heat flux, used in the limitingcalculation described above, was conservative by a factor of 5. An appropriate correction was madewhich resulted in a more realistic estimate of 470°C as the maximum clad temperature expected if filmboiling occurs. This result is in agreement with experimental evidence obtained for clad temperatures ofPage 4.1-1 UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORTAPPENDIX 4.1, PULSING THERMAL RESPONSE12/2011400°C to 500°C for TRIGA Mark F fuel elements which have been operated under film boiling conditions[Coffer et al. 1965]..1.3 -1.21.00.90.80Figure 4A.1.RADIUS (IN.)Representative Radial Variation of Power Within the TRIGA Fuel Rod1.10.00.9N.0.80.70.60.50 1 2 3 4 5 6 7 8AXIAL DISTANCE FROM MID-PLANE OF FUEL ELEMENT (IN.)Figure 4A.2, Representative Axial Variation of Power Within the TRIGA Fuel Rod.Page 4.1-2 UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORTAPPENDIX 4.1, PULSING THERMAL RESPONSE12/2011-S -~ flat- .- -~-- -____' I1o5, 10gCURVE USEDIN ANALYSISI I I I I I 1 I103* q .....I I L 1 I 1 I10102i03TW- TSAT (F)Figure 4A.3, Subcooled Boiling Heat Transfer for Water.z0.8RADIUS (IN.)Figure 4A.4, Fuel Body Temperature at the Midplaneof a Well-Bonded Fuel Element After Pulse.Page 4.1-3 UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT IAPPENDIX 4.1, PULSING THERMAL RESPONSE12/20111061010to3 _0.0010.01 0.1 l.0 10ELAPSED TIME FROM ENO OF PULSE (SEC)I00Figure 4A.5, Surface Heat Flux at the Midplane of a Well Bonded Fuel Element After a Pulse.10.*00010001010 ". l -I I ., totol O.l o. I I. t0.001 0.01 0.1 1.0 toELAPSED TIME FROM END OF PULSE (SEC)I00Figure 4A.6, Clad Temperature at Midpoint of Well-Bonded Fuel Element.The preceding analysis assessing the maximum clad temperatures associated with film boilingPage 4.1-4 UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT 12/2011APPENDIX 4.1, PULSING THERMAL RESPONSEIassumed no thermal resistance at fuel-clad interface. Measurements of fuel temperatures as a functionof steady-state power level provide evidence that after operating at high fuel temperatures, apermanent gap is produced between the fuel body and the clad by fuel expansion. This gap exists at alltemperatures below the maximum operating temperature. (See, for example, Figure 16 in the Cofferreport [1965].) The gap thickness varies with fuel temperature and clad temperature so that cooling ofthe fuel or overheating of the clad tends to widen the gap and decrease the heat transfer rate.Additional thermal resistance due to oxide and other films on the fuel and clad surfaces is expected.Experimental and theoretical studies of thermal contact resistance have been reported [Fenech andRohsenow 1959, Graff 1960, Fenech and Henry 1962] which provide insight into the mechanismsinvolved. They do not, however, permit quantitative prediction of this application because the basicdata required for input are presently not fully known. Instead, several transient thermal computationswere made using the RAT code. Each of these was made with an assumed value for the effective gapconductance, in order to determine the effective gap coefficient for which departure from nucleateboiling is incipient. These results were then compared with the incipient film boiling conditions of the1000°C peak fuel temperature case.For convenience, the calculations were made using the same initial temperature distribution aswas used for the preceding calculation. The calculations assumed a coolant flow velocity of 1 ft persecond, which is within the range of flow velocities computed for natural convection under varioussteady-state conditions for these reactors. The calculations did not use a complete boiling curve heattransfer model, but instead, included a convection cooled region (no boiling) and a subcooled nucleateboiling region without employing an upper DNB limit. The results were analyzed by inspection using theextended steady-state correlation of Bernath [1960] which has been reported by Spano [1964] to giveagreement with SPERT II burnout results within the experimental uncertainties in flow rate.The transient thermal calculations were performed using effective gap conductances of 500,375, and 250 Btu ft-2 hr1 'F-. The resulting wall temperature distributions were inspected to determinethe axial wall position and time after the pulse which gave the closest approach between the localcomputed surface heat flux and the DNB heat flux according to Bernath. The axial distribution of thecomputed and critical heat fluxes for each of the three cases at the time of closest approach is given inFigures 4A.7 through 4A.9. If the minimum approach to DNB is corrected to TRIGA Mark F conditionsand cross-plotted, an estimate of the effective gap conductance of 450 Btu ft2 hr- 'F- is obtained forincipient burnout so that the case using 500 is thought to be representative of standard TRIGA fuel.The surface heat flux at the midplane of the element is shown in Figure 4A.10 with gapconductance as a parameter. It may be observed that the maximum heat flux is approximatelyproportional to the heat transfer coefficient of the gap, and the time lag after the pulse for which thepeak occurs is also increased by about the same factor. The closest approach to DNB in thesecalculations did not necessarily occur at these times and places, however, as indicated on the curves ofFigures 4A.7 through 4A.9. The initial DNB point occurred near the core outlet for a local heat flux ofabout 340 kBtu ft-2 hr' OF- according to the more conservative Bernath correlation at a local watertemperature approaching saturation.This analysis indicates that after operation of the reactor at steady-state power levels of 1MW(t), or after pulsing to equivalent fuel temperatures, the heat flux through the clad is reduced andtherefore reduces the likelihood of reaching a regime where there is a departure from nucleate boiling.Page 4.1-5 UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORTAPPENDIX 4.1, PULSING THERMAL RESPONSE12/2011From the foregoing analysis, a maximum temperature for the clad during a pulse which gives a peakadiabatic fuel temperature of 1000'C is conservatively estimated to be 470°C.As can be seen from Figure 4.7, the ultimate strength of the clad at a temperature of 470°C is59,000 psi. If the stress produced by the hydrogen over pressure in the can is less than 59,000 psi, thefuel element.will not undergo loss of containment. Referring to Figure 4.8, and considering U-ZrH fuelwith a peak temperature of 1000°C, one finds the stress on the clad to be 12,600 psi. Further studiesshow that the hydrogen pressure that would result from a transient for which the peak fuel temperatureis 1150°C would not produce a stress in the clad in excess of its ultimate strength. TRIGA fuel with ahydrogen to zirconium ratio of at least 1.65 has been pulsed to temperatures of about 1150°C withoutdamage to the clad [Dee et al. 1966].7I-XI.I..LL.654~38 9 10 11 1213DISTANCE FROM BOTTOM OF FUEL (IN.)Figure 4A.7, Surface Heat Flux Distribution for Standard Non-Gapped (hgap = 500 Btu/h ft2 OF) Fuel Element After aPulse.Page 4.1-6 UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORTAPPENDIX 4.1, PULSING THERMAL RESPONSEI12/20117RI ICAL MEAT FLUXRI T CAL MEAT FLUX'U.LI. 1AT FLUXELAPSE97TIMME FROMNO OF PULSE IS0.316 SEI:4.321 a 9 IQ 1 12 13 15DISTANCE FROM 50TT01 OF FUEL (IN.)Figure 4A.8, Surface Heat-Flux Distribution for Standard Non-Gapped Fuel Element (hgap = 375 Btu/h ft2 OF ) After aPulse.aI.-I-0'47 CRtI T ICAL HEAT FLUX65ELAPSED TIME FROM END14 OF PULSE IS 0.1440 SEC3ACTUAL HEAT FLUX2[II .i t i i i --7 8 9 10 II 12 13DISTANCE FROM BOTTOM OF FUEL (IN.)114 15Figure 4A.9, Surface Heat-Flux Distribution for Standard Non-Gapped Fuel Element (hgap = 250 Btu/h ft2 OF ) After aPulse.Page 4.1-7 UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORTAPPENDIX 4.1, PULSING THERMAL RESPONSE12/2011106'.4'I-'4-z0-z'4.I-=L.A'4.U,0.1ELAPSED TIME FROM END OF PULSE (SEC)1.0Figure 4A.10, Surface Heat Flux at Midpoint vs. Time for Standard Non-Gapped Fuel Element After a Pulse.Page 4.1-8 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 55.0 REACTOR COOLANT SYSTEMSThe TRIGA is designed for operation with cooling provided by natural convective flow ofdemineralized water in the reactor pool. The suitability of this type of cooling at the powerlevels for this TRIGA has been demonstrated by numerous TRIGA installations throughout theworld.5.1 Summary DescriptionThe cooling system is composed of three subsystems: the reactor pool, pool cooling and poolcleanup.The principal function of the reactor pool is to remove fission and decay heat from the fuel, butpool water also serves to:" provide vertical shielding of radiation from the reactor," moderate fission energy neutrons, and" allow access to the reactor core for maintenance, surveillance, and experimental activities.Reactor pool functions are accomplished passively. Heat removal occurs by natural circulation.Shielding is provided by the height of the water above the reactor core. Shielding aspects ofthe pool are discussed in Chapter 11. Approximately 1/3 of the core volume is water,contributing to moderation of fission energy neutrons. Core physics are addressed in Chapter4. Maintenance, surveillance, and experiment activities are typically performed remotely (i.e.,from the pool surface, through the pool water) with long-handled or tethered tools.When the pool cooling system is operating, pool temperature is controlled by transferringheat from the pool water to a campus chill water system through heat exchanger. The poolcooling system is designed to maintain a higher pressure in the chill water system comparedto the pool cooling system, assuring pool water cannot leak into the chill water system. Poolcooling piping is designed with vacuum breakers to prevent potential siphoning through thepool cooling system.As described in Chapter 4, the fuel is encapsulated in a sealed stainless steel cladding; poolwater quality is controlled to assure cladding integrity by the pool cleanup system. The poolcleanup system recirculates pool water through a filter and ion exchanger to removesuspended solids and chemical impurities.5.2 Reactor PoolThe reactor pool is a tall tank formed by the union of two half-cylinders with a radius of 6separating the half-cylinders asPage 5-1 CHAPTER 5, REACTOR COOLING SYSTEMS 1 12/2011shown in Fig. 5.1A. The bottom of the pool is at the reactor bay floor level. The reactor coreis centered on one of the half-cylinders. Normal pool level is 8.01 meters above the bottomof the pool, with a minimum level of 6.5 m required for operations. The volume of water inthe pool (excluding the reflector, beam tubes and core-metal) is 40.57 m3 and 32.50 m3 forthe nominal and minimum-required levels. Basic reactor coolant system data is provided inTable 5.1.Table 5.1, Reactor Coolant System design SummaryMaterial Aluminum plate (6061)Reactor Tank Thickness 1/4 in. (0.635 cm)Volume (maximum) 11000 gal (41.64 M3)Pipes Aluminum 6061Iron-Plastic Liner, 316 SSCoolant Lines ValvesBalndSeBall and StemFittings Aluminum (Victaulic)Type CentrifugalCoolant Pump Material Stainless SteelCapacity 250 gpm (15.8 Ips)Type Shell & TubeMaterials (shell) Carbon steelMaterials (tubes) 304 stainless steelHeat Exchanger Heat Duty Flow Rate (shell) Flow Rate (tubes) Tube Inlet FTube Outlet69'Typical Heat Exchanger Operating Parameters Shell Inlet S h e ll O u t le t 5.2.1 Heat LoadThe reactor pool is open at the top (with an argon purge system normally drawing air across thesurface) surrounded by concrete. Conduction of heat through the concrete combined withforced convection and evaporation provide ambient cooling adequate to control pool watertemperature at low power operations and decay heat removal. At 1 MW operation the reactoris capable of heating up the pool under nominal level of 8.1 m at 20.7°C per hour, and at 2 MWapproximately 41°C per hour.Page 5-2 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 55.2.2 Pool FabricationThe pool is fabricated from sheets of 0.25 in. (0.635 cm) 6061 aluminum in 4 vertical sectionswelded to a Y2 in. thick aluminum plate. Full penetration inspection was performed on tankcomponents during fabrication, including 20% of the vertical seam welds, 100% on thebottom welds (internal and external to the pool volume), and 100% on the beam port weldexternal to the pool volume. A single floor centerline seam weld was used; a sealed channelwas welded under the seam and instrumented through a /4 in. NPT threaded connection toperform a leak test during fabrication. A 2 in. X 2 in. X /4 in. (square) aluminum channel wasrolled and welded to the upper edge of the tank.5.2.3 Beam PortsBeam port penetrations are fabricated around the core to allow extraction of radiationbeams 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 sectionextending outside of the pool.In pool sections are 0.1524 m (6 in.) in diameter, with a 0.00635 cm (0.25 in.) wall thickness.The in pool section for BP 1 and 5 is 6 in. while the remaining in-pool beam port sections arelonger. Supports (2 in. X 2 in. X Y4 in. aluminum angle bracket) are welded at one end to thebottom of the pool and at the other end directly to BP 2, 3, and 4 to support the weight ofthe extended lengths. BP 2 and 4 terminate at the outer surface of the reflector, while BP 3extends into the reflector, terminating at the inner shroud.BP 2 terminates in an oblique cut, extending approximately 43 cm (16.94 in.) into the poolwith the support 12.7 cm (5 in.) from the in-core end. BP 3 extends 73 cm (28.75 in.) into thepool with the support 37.62 cm (14.8125 in.) from the in-pool end. BP 4 extends 43 cm intothe pool (16.94 in.) with the support 7.62 cm (3 in.) from the in pool end. Beam port 1 and 5are aligned in a single beam line. A flight tube inserted into BP 1/5 extends through thereflector near the core shroud; BP 1 and 5 are equipped with a bellows to seal a neutronflight-tube. Beam ports 2, 3, and 4 are sealed at the in-pool end. BP 2 is tangential to thecore shroud, offset 34.29 cm (13 Y2 in.) from core center rotated 30' with respect to BP 3.Beam port 3 is oriented 90' with respect to BP 1/5, aligned to the center of the core.Alignment of BP 4 is through the core center, rotated 600 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 outerdiameter to the pool wall. The inner diameter of the flange is welded to the beam port tube.Page 5-3 CHAPTER 5, REACTOR COOLING SYSTEMSI12/2011The beam ports extend approximately 15.4 cm (6 in.) outside of the area define by the poolwalls. A stainless steel (304) ring is machined for a slip fit over the 6 in. (15.24 cm) aluminumtube extension. The ring is welded to 6 5/8 in. diameter stainless steel pipe (SST 304W/ASTM312) extending the flight tube for the beam port into the biological shielding.Four pads are welded to the pool floor reinforcing the floor for the core and supportstructure. As noted, the in-pool beam port supports are welded to the pool floor... ...... ... ...."Figure 5.1A, Pool FabricationIC?!lrl, CIt"CWOI IC AIN tIIIS'UFigure 5.1B, Cross SectionFigure 5.C, Beam Orientation5.3 Pool Cooling SystemThe pool cooling system is shown in Fig. 5.2.Figure 5.2, Pool Cooling System5.3.1 Reactor PoolThe reactor pool is open at the top (with an argon purge system normally drawing air acrossthe surface) surrounded by concrete. Conduction of heat through the concrete combinedPage 5-4 THE UNIVERSITY OF TEXAS TRIGA 11 RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 5with forced convection and evaporation provides ambient cooling adequate to control poolwater temperature at low power operations. At 1 MW operation the reactor is capable ofheating up the pool 20.7TC per hour, at 2 MW approximately 410C per hour. As noted above,fuel element cooling analysis assumed a maximum temperature of 48.9°C, which could beachieved after operating at the maximum power level for short periods. Therefore a poolcooling system is installed to control pool temperature.Historically the maximum pool temperature of 48.9°C was established to protect the integrityof ion exchange resin. The reactor pool is normally controlled at about 200C. In the absence ofpool cooling, a temperature rise of 28°C (from 200C to the maximum permissible pooltemperature of 48.9°C) could occur in 1.35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br /> at 1 MW, or about 40 minutes at 2 MW. Evenwithout pool cooling, time-limited support for experimental program is possible while stillmaintaining pool temperature below the limiting value used in analyses.5.3.2 Pool Heat ExchangerA tube and shell heat exchanger is installed for heat removal from the reactor pool to theavailable chilled water system. Design and operating parameters for the heat exchanger areprovided in Table 5.1. Heat exchanger capacity is designed to maintain reactor pooltemperature at or below the maximum temperature used in heat transfer analysis, 120°F(48.90C). The stable temperature is maintained by a heat exchanger capacity equivalent to thereactor core thermal output capacity. Other heat losses such as evaporation, or heat gains fromthe pump, are considered negligible. Heat transfer is defined by:q =U.A.6Twhere U overall heat transfer coefficient (watt/m2 -0C)A -surface area for heat transfer (M2)6Tin = true mean temperature difference (°C)The overall heat transfer coefficient of a tube and shell heat exchanger is composed of threeterms, the convective heat transfer from the fluid in the tubes to the tube walls, the conductiveheat transfer thru the tube wall, and the convective heat transfer from the outside tube wall tothe fluid in the shell of the heat exchanger. Based on the outside tube area for heat transfer,the overall heat transfer coefficient is defined as':Heat Transfer, Holman, JH. P., McGraw-Hill, 4th Edition (1976) pp386-391Page 5-5 CHAPTER 5, REACTOR COOLING SYSTEMSI12/2011u= + +A-Rkh hWherer:Ao is the total outside tube area (m2)Ai is the total inside tube area (M2)ri is the tube inside radius (m)ro is the tube outside radius (m)hi is the convective heat transfer coefficient between fluid in tubes(W/m2-oC)ho is the convective heat transfer coefficient between fluid in shell(W/m2-oC)k is the conductive heat transfer coefficient in the tube wall (W/m2-oC)I is the total tube length in heat exchanger (m)and tube walland tube wallA correction is applied for fouling of heat exchanger caused by buildup of various deposits. Theoverall heat transfer coefficient for a fouled heat exchanger is determined by:1U- 1where Rf is the fouling factor, (non-dimensional). The convective heat transfer coefficient isdefined ash Nu '-khe-dWhere:Nu is the Nusselt Numberk is the thermal conductivity of the fluid evaluated at the appropriate averagetemperature (W/m-°C)d is the tube diameter or applicable hydraulic diameter (m)Page 5-6 THE UNIVERSITY OF TEXAS TRIGA 11 RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 5The complicated nature of turbulent flow heat transfer is described by a Nusselt numberdetermined by experimental correlation with the Reynolds and Prandtl Numbers. Dittus andBoelter2 recommend the following relation for fully developed turbulent flow in tubes:Nut = 0.023. Re'

• Prnwhere parameters are measured inside the tubesRe is the Reynolds Number based on tube diameter,Pr is the Prandtl Number at average fluid temperature,n is 0.4 for heating, 0.3 for cooling.The relation for the shell side of a baffled cross flow heat exchanger is suggested by Colburn3 asfollows:Nut = 0.33 -Re0.6 Pr0.33where parameters are measured outside the tubes andRe is the Reynolds Number based on tube outside diameter and velocity at minimumshell cross sectional area,Pr is the Prandtl Number at average fluid temperature.The product terms, 6Tin, are defined consistent with the definition of U and heat exchangerdesign. The total cross sectional area of the tubes is represented by the heat transfer area, A, asspecified by the heat transfer coefficient, U. The true mean temperature difference, 6Tm, isrelated to the heat exchanger type by a correction factor, F, and a log mean temperaturedifference, LMTD 4. The correlation relates a simple single pass heat exchanger with morecomplex multiple pass baffled units. A relation is defined by6Tm = F -LMTDwhereF is the correction factor5, 6,2 University of California (Berkeley) Pub. Eng, Dittus, F. W and Boelter, L. M. K., Vol 2, pp 443 (1930)A method of Correlating Forced Convection Heat Transfer Data and Comparison with Fluid Friction, Colburn, A. P.,Trans. AlChE, Vol 29, pp 174-210 (1933)4 Heat Tansfer, White, op. cit.s Mean Temperature Difference in Design, Bowman, R. A., Mueller, A. C., and Nagle, W. M., Trans. ASME, Vol 62(1940) pp283-2946 Standards, TEMA 3rd Ed., Tubular Heat Exchanger Manufacturers Association New York (1952)Page 5-7 CHAPTER 5, REACTOR COOLING SYSTEMS12/2011T -TbLMTD = bIn TaTbFor a counter flow heat exchangerTa is (T hot fluid in -T cold fluid out)Tb is (T hot fluid out -T cold fluid in)Actual heat exchanger capacity is calculated using an energy balance on either the shell or tubefluid. The heat transfer is defined as:Tb = Thot-l~uia-out -Thotrluid-iaTa = Tftotaludin -Tofttluid-outq = C (Tin -Tout)whereC = m Cpm is the mass flow rate,cp is the fluid specific heat,Ti, is the temp of fluid entering heat exchanger,Tout is the temp of fluid exiting heat exchanger.In the current case Tout of either fluid is not known. Only Tin (100*F pool water, 480F coolantwater) and the mass flow rate of both fluids are known. To determine Tout theeffectiveness/NTU method7'8 is used. The dimensionless parameter called the heat exchangereffectiveness E is defined asActualHXMax_HXwhere the maximum possible heat transfer isMaXHX = Cmin " (r ot_,n -Told_in)Substituting (11) for each fluid and (13) into (12) results inChot (Thot, in -Tcoldan)cmin (Thotin -Tcoa i)7Heat Transfer, White, F. M., Addison-Wesley (1984) pp 512-5138 Compact Heat Exhcangers,2 nd Ed., Keys, W. and Landon, A. L., McGraw-Hill (1964)Page 5-8 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 5I12/2011for the hot fluid andCcold '(Tcold-in -Tcold in)Cmin *( Thot- -clinfor the cold fluid. The heat exchange effectiveness determinedexchanger with one shell pan and any multiple of tube passes= z+ e-N[B -1E~z- l~r+B.1e---N-.B]WhereR is Cmin/ CmaxU is the overall heat transfer defined in (2)A is the surface area for heat transferB is( 1+r2)12by9for a shell and tube heatOnce the effectiveness is calculated and the above used to determine Thot out and TcoId out.These may then be used to determine the capacity of the heat exchanger.Table 5.2, Heat Exchanger, Heat Transfer and Hydraulic ParametersComponent/Parameter Specification Value UnitsOutside Diameter in. (cm)Tubes Wall Thickness in. (cm)Thermal Conductivity Btu/hr-ft-°FTube Side in2 (cm2)Flow Area Shell Side in2 (cm2)Heat transfer Surfaces Na ft2 (M2)Average Prandtl No. Tube naShell naTube ft2/s (m2/S)Average Kinematic Viscosity Shell ft2/s (m2/s)Reynolds No. Tube naShell naCorrective Heat Transfer Tube Btu/hr-ft2--F (W/m2-°C)Coefficients Shell Btu/hr-ft2-°F (W/m2-°C)Overall Heat Transfer Tube Btu/hr-ft2 -F (W/m2-°C)Coefficient Shell Btu/hr-ft2 -F (W/m2-°C)Effectiveness Clean naFouled naLMTD Na F ('C)Corrective Factor F Na naCapacity Clean kW9 Compact Heat Exchangers op. cit.Page 5-9 CHAPTER 5, REACTOR COOLING SYSTEMS 12/2011Fouled 1070 kWHeat removal capacity and thus pool heat rate is specified by analysis of a tube and shell heatexchanger. Heat removal rate of 1140 kW is expected at a flow rate of 400 gal/min (25.2liters/sec) of chilled water at 48°F (8.89*C). The presence of fouling in the heat exchanger isconsidered minimal based on the purity of the two heat exchanger fluids. Capacity is reduced to1070 kW for a fouling factor of 0.0004.5.3.3 Secondary CoolingWhen the pool cooling system is operating, pool temperature is controlled by transferring heatfrom the pool water to a campus chill water system in a heat exchanger. The chilled watersystem is operated by the University for cooling of Pickle Research Campus buildings andequipment through a campus supply loop. At the time of NETL construction, chilling capacitywas provided by multiple 1200 ton (4220 kW) units, with 25% of the chilling system capacity ofone unit allocated to pool cooling. Construction is currently underway to remove a majorload/demand from the shared system; the Texas Advanced Computing Center is expanding, andinstalling a dedicated cooling system. The PRC chill water supply is also currently planning forsystem renovations which will expand capacity to meet campus growth and development.Chill water pumps in the NETL building draw from the campus supply loop and direct flow tothe loads at the NETL, including two installations (2 pumps each) supporting building ventilationand air conditioning, and a single pump providing chill water flow for the pool cooling systemheat exchanger.5.3.4 Control SystemChill water flow is normally about 500 gpm. Chill water flow through the heat exchanger isregulated to control pool temperature, sensed in flow to the heat exchanger. If temperature islower than the control setpoint, a bypass line diverts chill water flow around the heatexchanger. Conversely, if temperature rises above the setpoint, the bypass flow is reduced sothat more flow passes through the heat exchange.The pool cooling system is designed to maintain a higher pressure in the chill water systemcompared to the pool cooling system, assuring pool water cannot leak into the chill watersystem. Pool cooling is normally 250 gpm; if pressure at the chill water outlet rises above thepressure at the pool inlet to the heat exchanger, the pool outlet inlet is throttled by a controlvalve.Page 5-10 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 55.4 Primary Cleanup SystemThe primary cleanup system (Fig. 5.3) is designed to use filtration and ion exchange to controlwater quality for corrosion control, radioactivity control, and optical clarity of the coolantwater. Water purity is monitored by analysis of the water conductivity.The purification skid is located in room at about the same level as the reactor core. Theskid consists of a pump, flowmeter, filter, resin bed, and instrumentation. The cleanup systemis normally operated continuously to provide removal of suspended particles and soluble ions inthe coolant water. The system flow rate is about 10 gpm (0.6 Ips).Suction of water from the pool is provided by two inlets in the reactor pool, neither of whichextends more than 2 meters below the top of the reactor tank. Valves at the pool surface allowsuction from either a subsurface inlet or from a surface skimmer designed to collect andremove floating debris. Accidental siphoning of reactor pool water is prevented by siphonbreaks similar to those on the coolant piping. Return flow to the pool is through a subsurfacedischarge pipe. Valves are provided for isolation of the suction or return lines, and for isolationof system components for maintenance or resin replacement.--T@PFILTERr -ET OR T A 1I INO MAL IZERLOCAL AMIGJTINSTRUFMB4ATIONG)9 C.VLTIV1TY OSiiFigure 5.3, Pool Cleanup SystemPage 5-11 CHAPTER 5, REACTOR COOLING SYSTEMS 1 12/2011Purification functions of the loop are generated by two components, a filter for removal ofsuspended materials and a resin bed for removal of soluble elements. Typical filtration isprovided with 25 micron filters. Typical ion exchange is provided by 0.85 cubic meters of mixedcation and anion resin. Resin historically used is rated to 120°F; therefore the maximum pooltemperature used in analysis is 120°F (48.9°C). Resin performance is monitored as the decreasein conductivity across the demineralizer, measured by inline conductivity cells. Measurementsof water conductivity as low as 2.0 micromho per centimeter (or resistance of 1 megohm percentimeter) are maintained by filtration and ion exchange. The conductivity is reduced furtherby control of materials exposed to the reactor coolant, minimizing dust settling to the poolsurface, and occasional cleaning of pool surfaces. Experience has shown that conductivities of5.0 pmho/cm are sufficient to maintain acceptable limits on corrosion plus good water opticalquality and removal of activation products in the water.Should radioactivity be released from a clad leak or rupture of an experiment, detection of therelease would be signaled by the continuous air monitor or by the reactor room areamonitors. Based on coolant transport time calculations in the safety analysis section, thesemonitors should register an increase in coolant radioactivity within approximately 60 secondsof the time of radioactivity release. The transport time is estimated from the time for thecoolant exposed in the core to reach the surface of the water where the continuous air monitorwill detect a release of radioactivity from the pool water. An alternate indication of radioactiverelease is provided if a water activity monitor is installed or by a GM detector area monitor.Experience with this purification equipment in other TRIGA systems has shown that coolantconductivity can be easily maintained at levels of less than five micromhos per centimeter usingthe materials contained in the coolant system design. Furthermore, this experience has shownthat no apparent corrosion of fuel clad or other components will occur if the conductivity of thewater does not exceed five micromhos per centimeter when averaged over a 30-day period.5.5 Makeup Water SystemA connection from the domestic (potable) water system to the pool cleanup system providesmakeup water to replenish pool inventory losses from evaporation. The potable water headersupplies a mechanical filter and a bank of 4 deionizers. Each deionizer is capable of beingbypassed, and is instrumented with an indicator that energizes a white lamp if conductivity isgreater than 200 kmhos per cm, and a red lamp if conductivity exceeds the setpoint. Thedeionizers supply lab-spaces and makeup water to the pool cleanup system. A pumprecirculates water through the final deionizer and the laboratory distribution header.A line from the deionizers is routed through shutoff valves and a check valve to a flexibleextension in the water treatment room. The flexible extension is equipped with a conductivitymonitor and terminated in a quick disconnect fitting that allows physical separation of the twosystems except during periods in which the makeup process is operating. When the poolPage 5-12 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 5I12/2011inventory has decreased from evaporation, the quick disconnect is made up at the suction ofthe cleanup pump to provide makeup water through the cleanup filter and demineralizer.5.6 Cooling System Instruments and ControlsNumerous cooling and cleanup system parameters are measured by local sensors in the systemlines. Transmitters provide some of the parameters remotely to the control room.Temperature and pressure probes are located on the inlet and outlet lines of the pool waterside and chill water side of the heat exchanger. A local indication of flow in the coolant loop isprovided by the pressure drop across a venturi in the flow path. Purification loop flow ismeasured by an in line flow meter. Water pressure before and after the filter in thepurification loop is measured locally for indication of filter condition. Parameter monitoringpoints are illustrated in Fig. 5.2 and 5.3. The parameters that are considered part of the watersystem instrumentation system are presented in Figure 5.4.I----IE~Ir~'E2~Iui~Ir~CONTROL CONSOLEIsix MM meIREACTOR POOLFUL wS=10 VAIua iMVVMFM IMISCRVIZ VAMM 7zWMnM IlMKDMMUMMET COuRCM11YFDMwMIL-Uauez cazrinw1-TN9ZJ-IW~ PuL. VAD&ME PIMM TAPI]M 041LUJ VAIU!GmnZr M TAPI'Ir-3DUU VA M M" ~ n" I LIPCOOLANT TREATAENT AREAFigure 5.4, Cooling and Cleanup InstrumentationThe cooling system parameters normally available in the control room include coolanttemperatures, flow rates, differential pressure status, and pool level. Two temperature probes,one in the pool suction line and one in the line, allow monitoring of heat exchanger coolingfunction. Typical temperature probes used are resistance temperature detectors (RTD's).Two flow meters, one in the chilled water line and one in the pool water line provideinformation on system flow rates. A differential pressure monitor provides an alarm if thePage 5-13 CHAPTER 5, REACTOR COOLING SYSTEMS 12/2011pressure at the high pressure point on the heat exchanger tube side is not less than the lowpressure point on the shell side. The differential pressure is designed for a differencesubstantially greater than 7 kilopascals (1 lb/sq. in.).Water quality is measured by two conductivity cells in the purification loop. The cells arelocated on inlet and outlet lines of the demineralizer that readout locally in the control room.Typical conductivity cells are composed of two parts, titanium electrodes shielded by thyton forconductivity measurement, and a thermister for temperature compensation. A Wheatstonebridge circuit on the purification skid is connected to the cells. A switch allows selection ofeither inlet or outlet conductivity.Page 5-14 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 66.0 ENGINEERED SAFEGUARD FEATURESAs (1) discussed in Chapter 13, Chapter 5, (2) identified in previous analysis, and (3) identifiedfrom experience at other TRIGA reactors, emergency core cooling is not required for operationsat steady state thermal powers below 1900 kW. No engineered safety features are required forthe UT TRIGA II research reactor because the steady state power limit is 1,100 kW.6.1 ReferencesTRIGA Reactor facility, Nuclear Engineering teaching Laboratory, The University of Texas atAustin SAFETY ANALYSIS REPORT Submitted May 1991NUCREG-1135, Safety Evaluation report related to the Construction Permit and OperatingLicense for the Research Reactor at the University of Texas, Docket 50-602, May 1985NUREG-1282, "Safety Evaluation Report on High-Uranium Content, Low-Enriched Uranium-Zirconium Hydride Fuels for TRIGA Reactors," U.S. Nuclear Regulatory Commission, 1987.

THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 1 12/2011SAFETY ANALYSIS REPORT, CHAPTER 7 I7.0 INSTRUMENTATION AND CONTROL SYSTEMDesign of the instrumentation and control system was intended for TRIGA reactor facilities as areplacement of analog reactor consoles. Initial verification and testing of the design by themanufacturer was a requirement prior to installation at The University of Texas at Austin. Anevaluation by the University of the instrument and control console for the TRIGA was part of theinitial installation of the console by the vendor. The system development, installation, and initialtesting were the responsibility of the vendor, General Atomics.The system described in this document is a microprocessor-based instrumentation and controlsystem developed by the General Atomics (GA) TRIGA Reactor Division. This system incorporates(1) a digital wide-range neutron power monitor, (2) two analog power safety channels, (3) avariety of state-of-the-art signal conditioners and process controllers, plus (4) a digital dataacquisition and control system incorporating a PC compatible computer.There has been ample historical testing of the digital control system used at this facility. Digitalcontrol of research reactors has been accomplished by over twenty facilities across the UnitedStates for a number of years. The University of Texas digital TRIGA control system has beenoperating since 1992.7.1 DESIGN BASESThe design and manufacture of this system complies with the guidance given in American NuclearSociety and the American National Standards Institute Guide Criteria for the Reactor SafetySystems of Research Reactors (ANSI/ANS 15.15-1978)1,2. This standard has served the researchreactor community in lieu of the ad hoc application of similar standards for power reactors. Evenif single-failure criteria for plant protective actions -not deemed mandatory by ANSI/ANS 15.15for negligible risk reactors -were applied, the standard has allowed the use of simple redundancy,i.e., the monitoring of the same reactor parameter using independent, redundant equipment, tosatisfy the single-failure criteria for the reactor safety system.There are several advantages in a microprocessor-based system which enhances system safety,reliability, and maintainability over the analog control system used in previous TRIGA reactors:1. The use of microcomputers allows data (operator input as well as output) to be moreefficiently and systematically processed.2. Several data reductions (such as on-line calculation of the prompt period during a pulse)can be done in near-real-time.S"Criteria for the Reactor Safety Systems of Research Reactors", American Nuclear Society, American NationalStandard, ANSI/ANS-15.15-1978.2 "Microprocessor Based Research Reactor Instrumentation and Control System", INS-27, Rev. A., GA Technologies,August 1987.Page 7-1 CHAPTER 7, INSTRUMENTS AND CONTROLS j 12/20113. On-line self-diagnostics can be performed to determine the state of the system at alltimes.4. Operational surveillance and operations data are accommodated with all informationgathering and processing done routinely and regularly by the console computers.The Instrumentation and Control System for the TRIGA reactor3 is a computer-based designincorporating the use of one multifunction, NM-1000 microprocessor neutron flux monitoringchannel and two companion current mode neutron-monitoring safety channels (NP-1000 andNPP-1000). The combination of these two systems provides an independent operating channeland the redundant safety function of percent power with scram. The NM-1000 provides widerange log power and multi-range linear power from source level to full power. The control systemlogic is contained in a separate control system computer (CSC) with graphics and text displayswhich are the interface between the operator and the reactor. Another system for dataacquisition and control (DAC) functions as the interface point for interface circuitry, processsignals and communications. The multifunction NM-1000, NP-1000 and NP-1000 units, and twosystem microprocessors, the control system computer (CSC) and data acquisition and controlsystem (DAC) are development products of General Atomics. The basic system configuration isshown in Fig. 7.1.Information from the NM-1000 channel is processed and displayed by the CSC. The NP-1000 andNPP-1000 are independent channels that deliver steady state power level data to the safetysystem scram circuit, hardwired analog indicators, and to the CSC for processing and display. TheNPP-1000 also covers the pulse range. Operating ranges for the neutron channels are shown inFig. 7.2.The NM-1000 digital neutron monitor channel was developed for the nuclear power industry andis fully qualified for use in the demanding and restrictive conditions of a nuclear power generatingplant. Its design is based on a special GA-designed fission chamber, and low noise ultra-fast pulseamplifier. The NP-1000 and NPP-1000 were developed specifically for use with research reactorsafety systems and include several features not usually found in this type of application.The CSC and its acquisition system, the DAC, manage all control rod movements, accounting forsuch things as interlocks, and choice of particular operating modes. It also processes and displaysinformation on control rod position, power level, fuel and water temperature, and can displaypulse characteristics. The CSC also performs many other functions, such as calibrating controlrods, monitoring reactor usage, and historical operating data can be saved for replay at a laterdate.3 'Safety Analysis of Microprocessor Reactor Control and Instrumentation System", The University of Texas at Austin,1989.Page 7-2 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 712/2011FIGURE 7.1, CONTROL SYSTEM BLOCK DIAGRAM7.1.1. NM-1000 Neutron ChannelThe NM-1000 nuclear channel has multifunction capability to provide neutron monitoring over awide power range from a single detector. The selectable functions are any or all of the following:a. Percent power.b. Wide-range log power.c. Power rate of change.d. Multi-range linear power.For the TRIGA ICS, one NM-1000 system is designated to provide the wide-range log powerfunction and the multi-range linear power function. The wide-range log power function is a digitalversion of the patented GA 10-decade log power system to cover the reactor power range frombelow surface level to 150% power and provide a period signal. For the log power function, thechamber signal from startup (pulse counting) range through the Campbelling (root mean square[RMS] signal processing) range covers in excess of 10-decades of power level. The self-containedmicroprocessor combines these signals and derives the power rate of change (period) through thePage 7-3 CHAPTER 7, INSTRUMENTS AND CONTROLSI12/2011full range of power. The microprocessor automatically tests the system to ensure that the upperdecades are operable while the reactor is operating in the lower decades and vice versa when thereactor is at high power.2000 IM200 M6120 ON214W200 kWl20 kWt2 kit200 W20 W2 W0.2 W0.02 10.002 W0.0002 WTPULSENM41000 I .A 8CNPIOGO NPPIOOO% POWERI kWINTERLOCKTRIP-- ~~~,.SOURCE LEVELSOURCE INTERLOCK TRIP100110%1110-210,3%10-6%10,7%A a Wide Range Log Channel, B a Wide Range Linear Channel, NMIO00C

• Manual, Automatic, and Squarewave Modes,t4PlOOOFigure 7.2, NEUTRON CHANNEL OPERATING RANGESFor the multi-range function, the NM-1000 uses the same signal source as for the log function.However, instead of the microprocessor converting the signal into a log function, it converts itinto 10 linear power ranges. This feature provides for a more precise reading of linear power levelover the entire range of reactor power. The same self-checking features are included for the logfunction. The multi-range function is either auto-range or slave to a position switch on theoperator's console via the control system computer. A linear power level signal is available for thepercent power safety function for 1 to 125%.The NM-1000 system is contained in two National Electrical Manufacturers Association (NEMA)enclosures, one for the amplifier and one for the processor assemblies. The amplifier assemblycontains modular plug-in subassemblies for pulse preamplifier electronics, bandpass filter andRMS electronics, signal conditioning circuits, low voltage power supplies, detector high-voltagepower supply, and digital diagnostics and communication electronics. The processor assembly ismade up of modular plug-in subassemblies for communication electronics (between amplifier andprocessor), the microprocessor, a control/display module, low-voltage power supplies, isolated 4Page 7-4 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 712/2011to 20 mA outputs, and isolated alarm outputs. Outputs are ClassCommunication between the amplifier and processor assemblies is viacables.1E as specifiedtwo twisted-pairby IEEE.shielded0 0LUA".log PaSelFigure 7.3, Auxiliary Display PanelThe amplifier/microprocessor circuit design employs the latest concepts in automatic on-line selfdiagnostics and calibration verification. Detection of unacceptable circuit performance isautomatically alarmed. The system is automatically calibrated and checked (including the testingof trip levels) prior to operation. The checkout data is recorded for future use, and operationcannot proceed without a satisfactorily completed checkout. The accuracy of the channels is + 3%of full scale, and trip settings are repeatable within 1% of full-scale input.The neutron detector uses the standard 0.2 counts per "nv" fission chamber that has providedreliable service in the past. It has, however, been improved by additional shielding to provide agreater signal-to-noise ratio. The low noise construction of the chamber assembly allows thesystem to respond to a low reactor shutdown level which is subject to being masked by noise.7.1.2. NP-1000 Power Safety ChannelThe NP-1000 Power Safety Channel is a complete linear percent power monitoring systemmounted within one compact enclosure which contains current to voltage conversion signalconditioning, power supplies, trip circuits, isolation devices, and computer interface circuitry. Thepower level trip circuit is normally hardwired into the scram system and the isolated analogoutputs are monitored by the CSC as well as being hardwired to a bar graph indicator.Page 7-5 CHAPTER 7, INSTRUMENTS AND CONTROLSI12/2011A special version of the safety channel, the NPP-1000, provides measurement functions for peakpulse power, total pulse energy, automatic gain change and related trip points. The controlsystem automatically selects proper gain setting for steady-state or pulse mode when theoperator determines the reactor operating mode. Peak pulse power and total pulse energy arealso set by the pulse operation mode.Both safety channels, the NP-1000 and the NPP-1000, are identical except for the peak and energycircuits. The detector for each safety channel may be either an ionization chamber or self-powered in-core detector.7.1.3. Reactor Control ConsoleThe layout of the control console is shown in Fig. 7.3. The reactor control console contains severalcomponents needed by the operator for reactor control. Included are the following:a. Reactor control panels.b. Control System computer (CSC).c. Two color graphics monitors.d. Power and temperature meter panels.e. Disk drive storage and a graphics printer.MODE AND STARTUP WORKSTATIONSWITCHES CABINETASSEMBLYREACTOR INFORMATION MONITOR-POWER TRENDRECORDERSSTORAGE DRAWER'PRINTER DRAWERROD POSITION ANDCONTROL SWITCHESCONTROL SYSTEM*--COMPUTERKCOMPUTEREXPANSIONCHASSISBLANK PANEL>"19 IN. RACK MOUNTPEDESTAL ASSEMBLYFigure 7.3, LAYOUT OF THE REACTOR CONTROL CONSOLEPage 7-6 THE UNIVERSITY OF TEXAS TRIGA 11 RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 7A keyboard interface to the system computer is provided for operator control of several systemfunctions. As previously mentioned, the power and period information from the NM-100channel and power levels from the NP-1000 and NPP-1000 channels are processed and displayedby the CSC. However, several wide-range channel parameters are also present on linear bar graphmeter displays at the console. The NP-1000 and NPP-1000 safety systems are independent, withtheir own output displays, and connected directly to the control system scram circuit. Thus, wide-range log power, period, multi-range linear power and both percent power channels, have theiroutput displayed on meters as well as on the monitors. This is also true of fuel temperature.Typical layouts of the console panels and video displays are shown in Fig. 7.4 and 7.5.Functions of the rod control panel are represented in Figure 7.3 and are presented as:a. Key switch for rod magnet power (also operates "Reactor On" lights).b. Rod control switches and annunciators.c. SCRAM-switch for safety function.d. Annunciation is also provided for reset of the audio channel, as well as for reset ofthe alarm indicator following alarm clearance.The CSC provides all of the logic functions needed to control the reactor and augments the safetysystem by monitoring for undesirable operating characteristics. It displays reactor operationalinformation in a color format on a high-resolution LED monitor for ease of comprehension.Essentially all of the control systems logic contained in previous TRIGA reactor control systems isincorporated into the CSC. However, instead of using electronic circuits and electrical relaycircuits, the logic is programmed into the computer.The availability of the computer allows great versatility and flexibility in operationally-relatedactivities aside from the direct control of rod movements. Many other functions can also beperformed by the CSC, such as monitoring reactor usage, storing pulse data, reactor operatinghistory and logging operator usage.Two auxiliary cabinets can be provided to the console for the addition of process instrumentreadout.7.1.4. Reactor Operating ModesThere are four standard operating modes: manual, automatic, pulse, and square-wave. Themanual and automatic modes apply to the steady-state reactor condition; the pulse and square-wave modes are the conditions implied by their names and require a pulse rod drive. Manual andautomatic reactor control modes are used for reactor operation from source level to 100% power.These two modes are used for manual reactor startup, change in power level, and steady-stateoperation. The pulse mode generates high-power levels for very short periods of time. High-Page 7-7 CHAPTER 7, INSTRUMENTS AND CONTROLSI12/2011power and low-power pulse mode options are available. The square-wave operation allows thepower level to be raised quickly to a desired power level.aC'SII..'rnI-IwII!illDDQK-1.lo DI.PlY P-.1Mo. Control PV~e231. MWJ MW.A MW1--I MWAMW 10-23 BMW 10-4A MW.IMW is-2 Nw -_.1NMW 3 U23 MW -PWE -u igraphic displayFigure 7.4, CONSOLE CONTROL PANELSPage 7-8 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 712/2011Figure 7.5, TYPICAL VIDEO DISPLAY DATAROD POSITION CONTROL_7I II1MAGNET1POWETRANSIENT SHIM I SHIM 2 REGv v v v09890000]93Figure 7.6, ROD CONTROL PANELManual rod control is accomplished by the lighted push buttons on the rod control panel. The toprow of annunciators, when illuminated, indicate magnet contact with the armature and magnetcurrent. Depressing any one of the AIR-MAGNET push buttons will interrupt the current to thatmagnet and extinguish the magnet current on indication. If the rod is above the down limit, thePage 7-9 CHAPTER 7, INSTRUMENTS AND CONTROLS 1 12/2011rod will fall back into the core and the AIR-MAGNET light will remain extinguished until themagnet is driven to the down limit where it again contacts the armature.The middle row of pushbuttons (UP) and the bottom row (DOWN) are used to position the controlrods. Depressing the pushbuttons causes the control rod to move in the direction indicated.Several interlocks prevent the movement of the rods in the up direction under conditions such asthe following:a. Scrams not reset.b. Magnet not coupled to armature.c. Source level below minimum count.d. Two UP switches depressed at the same time.e. Mode switch in one of the pulse positions.f, Mode switch in AUTO position (regulating rod only).There is no interlock inhibiting the down direction of the control rods except in the case of theregulating rod while in the AUTO mode.Automatic (servo) power control can be obtained by switching from manual operation toautomatic operation. All the instrumentation, safety, and interlock circuitry described aboveapplies and is in operation in this mode. However, the regulating rod is now controlledautomatically in response to a power level and period signal. The reactor power level is comparedwith the demand level set by the operator and is used to bring the reactor power to the demandlevel on a fixed preset period. Logic for the automatic control operation by proportional, integral-differential (PID) control is contained within the digital algorithms of the control system. Thepurpose of this feature is to automatically maintain the preset power level during long-termpower runs. The function of automatic control is provided by the regulating rod with a steppingmotor drive.Reactor control in the pulsing mode consists of establishing criticality at a flux level below 1 kW inthe MANUAL mode. This is accomplished by the use of the motor-driven control rods, leaving thetransient rod either fully or partially inserted. The mode selector switch is then depressed. TheMODE selection switches automatically connect the pulsing chamber to monitor and record peakflux (nv) and energy release (nvt). Pulsing can be initiated from either the critical or subcriticalreactor state.In a square-wave operation, the reactor is first brought to criticality below 1 kW, leaving thetransient rod partially in the core. All of the manual instrumentation is in operation. Thetransient rod is ejected from the core by means of the transient rod FIRE pushbutton. When thepower level reaches the demand level, it is maintained much the same as in the automatic mode.Two rods are used, the transient rod to achieve power and the regulating rod to maintain power.Page 7-10 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 77.1.5. Reactor Scram and Shutdown SystemA reactor protective action4 interrupts the magnet current and results in the immediate insertionof all rods under any of the following:a. High neutron fluxes from either NP-1000 or NPP1000.b. High-voltage failure on the NM-1000, NP-1000, or NP1000.c. High fuel temperature (one out of two).d. Manual scram.e. Peak neutron flux or energy (pulse mode).f. Minimum period (available for use as desired).g. External safety switches (for experiments).h. Loss of electrical power to the control consolei. Watchdog circuits for each computer to monitor computer status by updatingtimers.All scram conditions are automatically indicated on the monitor. A manual scram will also insertthe control rods and may be used for a normal fast shutdown of the reactor. The scram circuitsafety function is an independent system that depends on wiring independent of the digitalcontrol system functions.Several conditions of the digital processing system will cause the scram mode condition. Amongthese are the loss of communication between the two computers, a database timeout conditionor failure of a digital input scanner. By updating dual programmable timers, watchdog circuits atperiodic intervals, determine the execution status of key elements of the computer digitalprogram.Two options for reliable operation performance may be installed as necessary. One option forconditions requiring long-term, high power steady-state operation, is configuration of three safetychannels with 2 out of 3 logic, allowing one channel to be out of service without requiring reactorshutdown. Another option is an uninterruptable power supply as auxiliary power for the reactorcontrol and monitoring systems for intermittent power failures of periods up to 15 minutes.4 "Safety Analysis of Microprocessor Reactor Control and Instrumentation System", The University of Texas at Austin,1989.Page 7-11 CHAPTER 7, INSTRUMENTS AND CONTROLS 12/20117.1.6. Logic FunctionsA simplified control system logic diagram is shown in Fig. 6.7. The two separate flux monitoringsafety channels ensure safe operation of the reactor by monitoring the power level and actindependently to shut the reactor down if a potentially dangerous condition exists. They provideinformation to the control system, which consists of three major parts: a reactor control console(RCC), Control System Computer (CSC) and Data Acquisition Computer (DAC). In addition, thereare two high resolution LED monitors and a graphics printer. The left-most display monitorcontains basic reactor operation control data. The second display monitor provides informationon annunciators and special control features. Data from both displays is available for log-records.The CSC provides the operator with immediate information concerning reactor conditions visuallyon the monitors. At the same time, the DAC is collecting data from the reactor system andconcentrating it into a permanent data base, which is transmitted to the CSC on request andmaintained for historical purposes.During operation of the reactor, the operator's commands to adjust control rod positions aretransmitted from the CSC to the DAC to the drive mechanisms. In the automatic mode the DACcontrols the position of the rods. The rod control program for automatic operation appliesproportional-integral-differential control logic. Digital rod position indication is shown in inches,with a resolution of < 0.1 in. and accuracy equal to or better than + 0.2% of indicated position.The control rod interface accepts the digital commands from the data acquisition and controlsystem (DAC) to operate the control rod motors. It contains the opto-isolation circuits which sendthe up-down limits and loss of contact signals to the control rod logic system. An excitation powersupply provides a stable reference voltage for the rod position indicator system.The magnet supply furnishes the required 200 mA needed for the rod magnets to hold controlrods in contact with the armature. An opto-isolator detects the absence of magnet current toeach drive magnet.A gamma chamber provides the signal for peak power (nv) and energy release (nvt) in the pulsemode. The nv/nvt amplifier provides the high impedance interface, high voltage and calibrationcircuits for the pulsing detector.All of the analog signals and digital signals are routed to the DAC chassis. However, the primereactor operating signals are also sent directly to the control room. These signals include logpower, period, percent power (2), fuel temperature (2), and pulse mode signals for peak andenergy.The DAC system converts the analog signals to a digital equivalent for transmitting along with thedigital signals to the CSC in the control room. The DAC chassis receives control instructions fromthe CSC, via the communication link, which in turn moves the control rods as requested by theoperator and causes the individual subsystems to go to the calibrate mode when commanded bythe system or operator.Page 7-12 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 712/2011The fuel temperature transmitters are accurate, highly stable units which convert the 0-600°C fueltemperature into a 4-10 mA output signal. A level comparator is included which provides scramcapability through an isolated contact state change when the preset level is exceeded.The water temperature transmitters are standard Resistance Temperature Detector (RTD)transmitters which convert the 0 to 100°C temperature into a 4-20 mA signal. The transmittershave a self-contained power supply.External switches are provided with terminalthe DAC chassis (beam port open-close, etc.)strips to terminate and connect various switches toCONTROL ROOMFigure 7.7, LOGIC DIAGRAM FOR CONTROL SYSTEM7.1.7 Mechanical HardwareTypical reactor installation are contained in two NEMA enclosure junction boxes, one electronicequipment cabinet, separate stepping motor power supplies installed in the reactor bay, andreactor operator console components installed in the reactor control room.Page 7-13 CHAPTER 7, INSTRUMENTS AND CONTROLS 12/2011The control console consists of the components needed by the operator for reactor control.These components include rod control switches and annunciators, the digital rod positionindicators, on-line reactor status meters (power and temperature), the control system computer(CSC), reactor operating mode switch panel, LED monitors, printer, disc drives (2) and externalswitch annunciators (beam port open-close, reactor access, etc.).Enclosure 1 contains NM-1000 high and low voltage power supplies, a pulse pre-amp withdiscriminator, an RMS Campbell convertor and a communications module.Enclosure 2 contains the NM-1000 microprocessor selected to provide the 10-decade log signaland the multi-range linear function from the information provided by the circuits in enclosure 1.The information processed by the microprocessor is 10-decades of log power, rate of powerchange (period), multi-range linear function, linear percent power from 1 to 125%, level trips fromthe log and linear percent power, calibrate and failure signals.Enclosure 3 is a standard rack type equipment enclosure for electronic components. Space in theenclosure provides the terminal strips for connections to the various signal detection systems andthe communications to the RCC. The cabinet enclosure includes eight shelves with functionalseparation between shelves. Power supplies for subsystems are on shelf 1. Shelves 2 and 3contain, respectively, ac digital and dc digital circuits for processing input or output circuits. Shelf4 provides several special modules for signal processing. The two power safety channels arepositioned on shelf 5. Shelves 6-8 contain the computer. The regulating rod drive translator forthe stepping motor drives is contained in a separate, fourth enclosure.7.2 DESIGN EVALUATIONThe TRIGA reactor console5'6'7 [6,7,8] has developed through the successful operation of manyinstalled facilities throughout the world. Design of the ICS unit incorporates similar basic logicfunctions proven effective in prior designs. Incorporation of digital electronic techniques in thedesign to replace analogue circuits is justified by improved performance. Functional self-checks,circuit calibrations, and automated data logging are implemented effectively and efficiently.A multiphase design, development and installation program by the system manufacturer providedthe initial demonstration of the system acceptance by analysis and review. No licensemodification was found necessary to implement the new digital system in place of the old analogsystem. The analysis by both the manufacturer and during operation on site determined 1) thatthere was no increase of the probability of occurrence or the consequences of an accident ormalfunction of equipment important to safety, 2) that the system does not create the possibilityof an accident or malfunction of a different type and 3) no reduction occurs in the margin ofsafety as defined in the basis for any technical specification.5 "Operation and Maintenance Manual Microprocessor Based Instrumentation System for the University of TRIGATexas Reactor", E117-1004, General Atomics 1989.6"Operation and Maintenance Manual NM1000 Neutron Monitoring Channel", E117-1000, General Atomics 19897 "Operation and Maintenance Manual NP1000/NPP1000 Percent Power Channel", E117-1010, General Atomics 1989.Page 7-14 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 88.0 ELECTRIC POWER SYSTEMSElectric power on the Pickle Research Camus is distributed underground.The main breaker for the NETL is 3 phase, 480 volts AC (with a 277 tap) rated at 600 amps perphase. 480 VAC power is supplied to:* HVAC Fans" Chill water pumps" Pool cooling pump* Laboratory vacuum pump" Laboratory air compressor" Instrument air compressor* Crane* Elevator277 VAC power is supplied to the reactor bay lighting transformer.Motor control center and load control center panels are located in a machine room adjacent tothe reactor bay on the middle level and upper levels.An emergency diesel generator operated and maintained by the facilities maintenance on thePRC provides backup power for lighting and sump pumps.The reactor safety and control systems are failsafe, in that a power supply failure causes thereactor to shutdown. The underground distribution system prevents the potential for mostexternal events affecting the power supply, with exceptions that damage the distributionstation.

THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 99.0 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, approximately precast concrete tilt panels 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, interfacing 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.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 on-site boiler unit with a designcapacity set by local building (HVAC) requirements. The cooling system is a PRC chilled watertreatment plant with design capacity 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 SYSTEMSI12/2011The 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.Figure 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 illustrated in Fig. 9.1, the auxiliary purge system discharge is within the HVACexhaust stack. The auxiliary purge 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 basisThe design goal for HVAC system is to control the reactor bay, adjacent zones and academicwing of the building at a negative pressure difference relative to ambient atmospheric pressureduring routine operations. The differential pressures are 0.06: 0.04: 0.03 in. water (0.15: 0.10:0.80 cm of water). This pressure gradient assures that any radioactive material released duringroutine operations is discharged through the stack and does not build up in the reactor bay.Release of airborne radioactivity consists mostly of activated 41Ar from routine operation.Page 9-2 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 9 1During potential accident conditions, sensors initiate confinement system isolation when highlevels of radioactivity are detected in reactor bay air, e.g. If a fuel element failure releasesfission products or if an experiment with sufficient inventory of radioactive material fails. Thereactor room confinement is designed to control the exposure of operation personnel and thepublic from radioactive material or its release caused by reactor operation. Release criterion isbased on Title 10 Chapter 20 of the U.S. Code of Federal Regulations.Confinement system ventilation has three modes of operation. When the reactor is notoperating (quiescent mode), the ventilation system is operated to minimize requirements forconditioning incoming air, in a recirculation mode with a minimal exhaust flow rate and freshair intake as required to maintain a negative pressure in the reactor bay with respect toadjoining spaces. When the reactor is operating (reactor run mode) the system is operated togenerate a rate of air exchange exceeding 2 air volumes (4120 M3) per hour, maintain a stackvelocity, and regulate negative pressure in the reactor bay. In the event that airborneradioactive material exceeding a trip set point is detected, the system is designed to establish ashutdown and isolated condition.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 confinement system uses heating and cooling in a singleunit, the remainder of the building HVAC system has air conditioning split into separate hot andcold decks.~~4T ~ 0" 6OILT STACKDAMPER AEKYVE ROOFReturnFun ISOLATIONFan :DA~APEXSAYSLPPLYYAIRFILTER a TEVIPEMTRA'l ISOLATION4CONRO CONTROL DAMOPESFigure 9.2A, Main Reactor Bay HVAC SystemPage 9-3 CHAPTER 9, AUXILIARY SYSTEMS 12/2011MINAIR W-A TCTAMIENT DAY STP..Y S2rl-HVAC OPERATION P'MODESU AIR ER DAAP"S' Y. SWITCHY FANLSml S ~HVAC ~OPEATIO AODE"RDDk AIR PR CfLSFANBY S.LFP FAN SPEFAN ~~*fM AIR PRESUR CUNl RCLA.By SUPLY FANr vrFigure 9.2B, Main Reactor Bay HVAC Control System ControlTable 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/s9.2.3 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 down the wall opening near the floor, enhancingmixing and preventing stratification. Air is discharged in parallel duct work (near the ceiling,near the floor) to an exhaust fan. In the reactor run mode the confinement system exhaust fanis controlled to maintain stack velocity designed to exceed the minimum air changespecification. Control dampers are located at the supply fan inlet (fresh air intake) and theexhaust fan outlet (discharge to stack), and in a line between the inlet and outlet ducts.Confinement system ventilation discharge is through a stack on the reactor building roof.Schematics of the ventilation system for the reactor bay area and a logic diagram of theventilation control system sensors and controls are provided in Fig. 9.2A and B.Page 9-4 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 9In the reactor run mode, confinement ventilation is balanced to provide at least reactor-bay 2air changes per hour and a preset stack flow rate:(1) One control damper (inlet/outlet duct cross connect) is closed(2) Supply and exhaust control dampers are open(3) The exhaust fan is controlled to provide a specified stack velocity(4) The supply fan is controlled to maintain the reactor bay at nominal 0.06 in. waterIn the quiescent mode, the confinement ventilation system is balanced for recirculation flowwith a small amount of effluent:(1) One control damper is throttled open across connecting inlet and outlet ducts,(2) Inlet and outlet control dampers are set to a minimum open position,(3) The outlet fan is operated at a constant, minimal speed(4) The supply fan is controlled to maintain the reactor bay at nominal 0.06 in. water.In confinement isolation mode:(1) All isolation dampers are closed(2) Supply and exhaust fans are securedAtmospheric dispersion using a stack model requires stack discharge 60 (18.23 m) feet abovethe ground, and at least 2 and Y2 times the height of adjacent structures. The nearest structureis approximately 80 meters 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 meters) above the roof level so that the stack discharge is 63 feet(19.202 m). The effective release point above the exhaust stack can be calculated from theBryan -Davidson equation:(vs) 1.4Ah -DWhere:Ah is the height of plume rise above release point (m)D is the diameter of stack (m), confinement vent 0.4012 M2, auxiliary purge 0.152 m2P is the mean wind speed at stack heght (m/s)V, is the effluent vertical efflux 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 velocity} m above the stack, and the effective stack height for theauxiliary purge system is 22.25/{wind velocity} 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.Page 9-5 CHAPTER 9, AUXILIARY SYSTEMS 12/2011Pneumatically operated isolation dampers in the confinement system ventilation are located atthe supply fan outlet (supply to the reactor bay) and the exhaust fan inlets (return from thereactor bay) near the reactor bay wall penetrations as indicated in Fig. 9.1. Controls close thedampers and secure the fans in response to manual or automatic signal initiated by highairborne particulate radioactivity. Loss of instrument air or loss of control power will cause thedampers to isolate the 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.Figure 9.3, Confinement System Ventilation ControlsAlarm indicators on the control panel provide indication that the differential 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 10CFR20. 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.9.2.5 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 materialPage 9-6 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 9in the reactor bay. If the 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 such 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.9.3.2 System descriptionMIN AIRARG PURGE LP. W5VtVm WAET1 R SAY PUIRGE FILTEJR CAISSI13/I RH UG 98-957!P fLEA FILTER E-P. S :7C2FAN I E, -31 2 99.977. 'EPA FILTERE~~hUST ~3 FUTURE O*HARI. FILTER9PE A7S 4 FUTURE. I.PA FILTER MIN .AM4U-P. 1dZ8tIKATICPOEILT1ER Sml, DASam leT RISMATI ON Ta ..DZ* POLSRAE L2I C3- sxx ^m\ ,REACTOR BAYLWV URmFigure 9.4A, Purge Air System Figure 9.4B, Purge Air Controls9.3.3 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 isolation damper in thepurge system is actuated manually, using the fan control switch. Automatic isolation of thePage 9-7 CHAPTER 9, AUXILIARY SYSTEMS 12/2011system is generated by the same particulate radiation monitor as is used by the HVACconfinement ventilation system.9.3.4 Instruments and controlsThe auxiliary purge system is controlled from the same panel as the confinement ventilationsystem. A flow gage indicates purge stack velocity at the panel. The exhaust point is concentricto the center 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.9.3.5 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 0.9 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 descriptionThere is space for a large number of fuel racks inside the reactor pool. The racks are aluminum,suspended from the pool edge by connecting rods. Page 9-8 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 9To facilitate extrastorage, 2 racks may be attached to the same connecting rods by locating one rack at adifferent vertical level and offsetting the horizontal position slightly. Outside the reactor pool, rack design isintended to fit in special storage wells (Fig. Watermay be added for shielding or cooling. Outside the reactorpool, supplemental fuel storage is planned fortemporary storage of elements transferred toor from the facility, for isolation of fuelelements with clad damage, emergencystorage of elements from the reactor pool andcore assembly and routine storage of otherradioactive materials. Temporary storage forreactor components or experimentsalso use the fuel storage racks in thereactor pool. Other locations not in the poolwill also provide storage for radioactive non-fuel materials.Page 9-9 CHAPTER 9, AUXILIARY SYSTEMSI12/2011CHAPTER 9, AUXILIARY SYSTEMS 12/2011A 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 elementsA 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 byTRGIA 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 failurePage 9-10 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 99.4.4 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.9.5 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.1Code 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 9-11 CHAPTER 9, AUXILIARY SYSTEMS 12/2011Basic 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.Automatic 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 wing 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, andPage 9-12 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 9other building safety features. An emergency shower and eye wash are available in the hallwayadjacent to laboratory areas.Each of the three components of the fire protection program is applied to the design, operationand modification of the reactor facility and components. Fire prevention 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 halon system is equipped with a local alarm to promptevacuation of the control room prior to system actuation; a manual override can defeat thesystem if the nature of the event does not require actuation of the system.Fire suppression is used only in areas where there are no significant quantities 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 telephone 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.Page 9-13 CHAPTER 9, AUXILIARY SYSTEMS12/20119.5.2 System descriptionThe telephone system is installed 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 locations 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 prompts personnel to initiate protective actions.An 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 specifically to 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 and office wing.Page 9-14 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 99.6.1 Design basisThe design basis of the NETL laboratories is to allow the safe and controlled use of radioactivematerials.9.6.2 System description (drawings, tables)Strategic lab and office wing rooms are equipped with fume hoods and ventilation to controlthe potential for release of radioactive materials. One room 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 10.9.6.3 Operational analysis and safety functionEngineered controls permit safe handling of radioactive materials.9.6.4 Instruments and controlsAn installed radiation monitor ensures personnel handling 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.Page 9-15 CHAPTER 9, AUXILIARY SYSTEMS12/20119.7.2 System descriptionThe 3 element facility, 6 element facility, pneumatic in-core terminals, and central 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. Concrete blocksprovide temporary shielding as needed.9.7.4 Instruments and controlsInstruments and controls associated with specific facilities are addressed in Chapter 10.9.7.5 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.Page 9-16 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 9 .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.9.8.4 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 Specifications specifically associated with the compressed air systems.Page 9-17 THE UNIVERSITY OF TEXAS TRIGA 11 RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 1010.0 EXPERIMENTAL FACILTIES AND UTILIZATION10.1 Summary DescriptionThe Nuclear Engineering Teaching Laboratory (NETL) experimental facilities support teaching,research, and service work. Multiple courses are taught at NETL that focus on reactoroperations, radiation detection, radiochemistry, and health physics. With the reactor facilitybeing the center focus of NETL, many of the nuclear analytical techniques utilize neutrons formaterials probing or activation. Isotope production is performed largely for detectorcalibrations and specialized experiments. In-core experimental facilities are used mostly foractivations for neutron activation analysis and for detector calibration related isotopeproduction. Beam port facilities utilize neutrons for either activation or imaging. Laboratoryfacilities are utilized for radiation detection and measurement along with radiochemistry. Theneutron generator facility contains a D-T neutron generator utilized for neutron activationstudies. The subcritical assembly is utilized for teaching and neutron source basedexperiments. The UT Austin TRIGA does not have thermal columns or irradiation roomsassociated with the reactor.List of experimental facilities1. In core facilitiesa. Central Thimbleb. Fuel element positionsc. Pneumatic transfer systemsd. Three element Facilitye. 6/7-Element Facility2. In reflector facilities/Rotary Specimen Rack3. Automatic transfer facilitiesa. Manualb. Automatic4. Beam ports5. Cold neutron source6. Non-reactor experiment facilitiesa. Neutron generator roomPage 10-1 CHAPTER 10, EXPERIMENTAL FACILTIES AND UTILIZATION12/2011b. Sub critical assemblyc. Laboratoriesi. Radiochemistry laboratoryii. Neutron activation analysis laboratoryiii. Radiation detection laboratoryiv. Sample preparation laboratoryv. General purpose laboratoryThe facility runs experiments in three basic categories: 1) in core irradiations, 2) beam portexperiments and non-reactor experiments. The majority of in core experiments are irradiationsfor neutron activation analysis. Other common in core experiments are irradiations to producesources for detector calibrations and irradiations for either materials damage or electronicsdamage studies. Beam port experiments utilize the neutrons for various nuclear analyticaltechniques from neutron depth profiling to prompt gamma activation analysis to neutronradiography. Non-reactor experiments include those that utilize the D-T neutron generator orother radiation sources.Experimenters work with licensed reactor operators for experiment planning, facility access,and facility utilization. Radiation monitors are placed near unloading points for in coreexperiments and near beam port facilities. Reactor operators watch neutron monitors adjacentto the reactor core monitor for reactivity perturbations resulting from in core experiments. An41Ar system monitors the activation of air within the core and beam ports. A continuous airmonitor tracks radioactive aerosols that may be produced from experiments or fuel leakage.Access to beam port facilities is directly displayed on the reactor console.Reactor based experiments and other experiments utilizing radiation sources such as the D-Tneutron generator are reviewed by the Reactor Oversight Committee (ROC). A safety analysis iswritten by the experimenter and often presented in an oral format to the ROC. A ROCsubcommittee is nominally formed to review the written safety analysis document. Evaluationcriteria include but are not limited to a radiation exposure assessment, core reactivity effects,radiation levels produced, chemical nature of experiment, and heat transfer effects. Thesubcommittee members then make recommendations to the ROC Chair regarding approval,denial, or recommended changes to the experiment. After a positive review process, theexperiment then becomes an approved experiment. Experimenters schedule reactor timeutilizing Operation Requests that are reviewed by a senior reactor operator to ensure that thework is an approved experiment.Page 10-2 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 10I12/201110.2 In-Core FacilitiesIn core irradiation facilities include a central thimble, penetrations for flux probes along twoperpendicular axes, and four facilities that displace (3, 6, or 7) fuel elements. Cutouts in theupper grid plate accommodate removable plates that position fuel elements when the facilitiesare not in use. Page 10-3 CHAPTER 10, EXPERIMENTAL FACILTIES AND UTILIZATIONI12/201110.2.1 Central Thimble (In-Core Facility)A. DESCRIPTION.The central thimble provides access to the maximum neutron flux in the reactor. The centralthimble has two modes, normal and beam operation.Experiment objectives for normal operations maximize activation, gamma irradiation, orreactivity. The central thimble is used to enhance activation or radiation damage. Enhancedactivation supports isotope production or neutron activation analysis, while enhanced radiationdamage supports radiation damage studies. Experiments or research in reactor kinetics may beperformed with the central thimble.The design of the central thimble permits extraction of a neutron or gamma beam to thebridgework over the pool. Typical beam experiments such as radiography and prompt gammaanalysis may be accomplished using the central thimble in the beam mode.The central thimble consists of an. aluminum tube extending through the core. The centralthimble provides access to the maximum neutron flux available in the core for samplePage 10-4 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 10irradiation or beam experiments. Samples are placed (normally in an aluminum canister) intothe central thimble from the bridge. A threaded cap covers the top of the central thimble whenthe facility is not in use. Water can be displaced in the central thimble volume above the corewith pressurized air to use the central thimble as a beam.B DESIGN & SPECIFICATIONSThe central thimble is approximately 7.2 m long. The central thimble is assembled fromthree sections of tubing with the bottom tube sealed on the lower end. Sections are joinedwith water-tight aluminum or stainless steel connectors with the tube and a sealing sleevejoined and sealed on each side by a large aluminum nut. The bottom two sections (originallyused at the UT TRIGA I reactor) are 10 ft. long (3.048 m).The central thimble extends from the reactor bridge through the (radial) center of the coreto approximately 7.5 in. (0.19 m) below the lower grid plate and 8.7 in. (0.22 m) above thesafety plate. The central thimble tube outer diameter is 1.5 in. (3.81 cm.), with 1.33 in. (3.38cm.) inner diameter. There are. four /4 in. (0.00635 m) holes in the central thimbleapproximately 3 in. (0.762 m) above the upper grid plate to ensure the volume in the core ismaintained in a flooded condition. Figure 10.3 illustrates the central thimble union assembly.wFigure 10.3: Central Thimble Union AssemblyThe central thimble tubing is aluminum alloy 6061. The alloy is a precipitation hardeningaluminum alloy, containing magnesium and silicon as its major alloying elements. It has goodmechanical properties and exhibits good weldability.The mechanical joint at the lower junction is prefabricated aluminum with a stainless steelsleeve. The upper mechanical joint may be either aluminum or stainless steel.Page 10-5 CHAPTER 10, EXPERIMENTAL FACILTIES AND UTILIZATION 1 12/2011Aluminum 6061 is a widely used material in aircraft and structural applications. Typical densityfor 6061 alloy is 2.7 g cm-1.Table 10.1 provides the material composition of Aluminum 6061.Table 10.1: Composition of Al 6061Component Wt. %Al 95.8 -98.6Cr 0.04 -0.35Cu 0.15-0.4Fe (Max) 0.7Mg 0.8 -1.2Mn (Max) 0.15Si 0.4-0.8Ti (Max) 0.15Zn (Max) 0.25Other, total (Max) 0.15Other, each (Max) 0.05The 6061 alloy has excellent joining characteristics, and good acceptance of applied coatings.The alloy combines relatively high strength, good workability, and high resistance to corrosion.Aluminum 6061 has a high resistance to corrosion. The central thimble tubing is anodized tofurther control potential corrosion.C. REACTIVITYThe original Safety Analysis Report for the UT at Austin TRIGA reactor provided data indicatingthat replacing the central thimble with a standard fuel element would result in a reactivitychange of 0.90% Ak/k ($1.29), and that replacing the central thimble with a void would result ina reactivity change of -0.15% Ak/k (-$0.21). As noted above, voiding of the portion of thecentral thimble in the core region is prevented passively by design.D. RADIOLOGICAL ASSESSMENTActivation of argon dissolved in water will occur in the central thimble region whether thecentral thimble is installed or not. Radioargon is considered as a normal byproduct of reactoroperation. Calculation of argon production and the consequences from normal operations isconsidered in Chapter 11.Portions of the central thimble in the core area will become activated, principally minorconstituents of 6061 aluminum alloy. A conservative irradiation scheme of 60 years at 2X 1013 ncm 2 s-1 followed by a week of decay using nominal values of 0.7% Fe, 0.4% Cu, 0.35% Cr, and0.25% Zn results in specific activities identified in Table 10.2.Page 10-6 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 10I12/2011Table 10.2: Activation Products in Central Thimble 6061 Aluminum Alloy after 60 Year IrradiationElement Target o IsotopeIsotope Concentration Produced Half Life ActivityIron Fe-54 392.1 pg/g Fe-55 2.7 years 4.889 mCiFe-58 21.78 I'g/g Fe-59 44.53 days 35.7 pCiCopper Cu-63 2.740 mg/g Cu-64 12.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> 5.625 GCiChromium Cr-50 146.2 pg/g Cr-51 27.7 days 6.969 mCiZinc Zn-64 1.187 mg/g Zn-65 243.9 days 6.255 mCiThe central thimble is normally installed for all operations, and does not create any increasedradiological hazards during operations unless the volume above the core is voided for beamexperiments. If the central thimble is used in a beam experiment the experiment proposal,review, and approval process will evaluate the need for additional radiological controls.Portions of the central thimble in the core area will become activated, principally minorconstituents of 6061 aluminum alloy. Using values for activation previously calculated, doserate from the 15 in. (0.381 m) of the tube adjacent to the active fuel region using a point sourceapproximation is estimated approximately 150 mR h-1.However the central thimble can besuspended in the reactor pool indefinitely or removed from the pool using a shielded container.Radiological hazards associated with materials to be irradiated in the central thimble areevaluated as part of the experiment review and approval process.E. INSTRUMENTATIONThere is no instrumentation associated with the central thimble. Instrumentation that might beused as part of an experiment program will be evaluated as part of the experiment review andapproval process.F. PHYSICAL RESTRAINTS, SHIELDS, OR BEAM CATCHERSThe central thimble facility is shielded by the same materials that shield the reactor core. Theseinclude water and concrete.G. OPERATING CHARACTERISTICSIsolation from the control rods prevents any potential interaction between the control rods andthe central thimble. Maintaining the volume in the core flooded by passive means preventslarge reactivity changes associated with voiding and flooding.Page 10-7 CHAPTER 10, EXPERIMENTAL FACILTIES AND UTILIZATION 12/2011The central thimble in the core is a static volume, open to the pool only through thepenetrations above the core. The penetrations in the central thimble tube above the uppergrid plate core eliminate any possible impact on loss of cooling potential or consequences.Cooling for material in the central thimble is principally through conduction to the water in thecore with some thermally induced circulation inside the central thimble.H. SAFETY ASSESSMENTThe central thimble facility does not carry much risk during reactor operation. Reactivitychanges could occur as a result of sample introductions. Such reactivity changes would have tobe assessed on an individual basis as part of the experimental review process. However,maximum reactivity additions would not likely be more than a fuel element which would bewithin the realm of allowed reactivity for a fixed experiment. Negative reactivity changes mayoccur due to sample introduction into the facility or for water introduction into the facilitywhen it is voided. Sample reactivity has to be assessed on an individual basis and must complywith Technical Specifications. Reactivity changes from water leakage into a voided centralthimble facility were calculated to be -0.15% Ak/k (-$0.21) which would not appreciably affectreactor safety.10.2.2 Fuel Element Positions (In-Core Facilities)Fuel element positions can be used for in-core irradiation facilities including single fuel elementpositions and multielement positions incorporated in the grid plate design. Standard facilitiesused in fuel element positions include in-core terminals for a pneumatic sample transitsystem and two types of multielement-position irradiation facilities (displacing 3 fuel elements,6 elements and the central thimble, or 7 elements). Proposals for any other in-core facilities orirradiation of materials in existing facilities are evaluated as part of the experiment review andapproval process.10.2.2.1 Pneumatic Sample Transit SystemA. DESCRIPTION.The pneumatic transit system is used to support neutron activation analysis and isotopeproduction. Major components of the pneumatic sample transit system include:" In-core terminus assemblies* Receiver assemblies" Blower-and-filter assembly" Valve assemblyPage 10-8 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 10* Control assembly" Specimen capsulesThree different in-core terminals are available for insertion into a core fuel position. Receivingstations are available in the reactor bay and in an adjacent laboratory (either in a fume hood orin an automated counting and analysis station); an additional sample line is available fordevelopment. Two capsule sizes are available, a large capsule with an internal volume of 25cm3, and a small capsule with an internal volume of 5 cm3.B. DESIGN & SPECIFICATIONS.The system design is a modification to the original, standard General Atomics pneumaticterminal system. Transit lines connect unions at the reactor pool-side to a mechanical switch(used to select the receiving station). Samples can be loaded from and delivered to receivingstations in the reactor bay, a fume hood in 3.102, or a counting station in room 3.102. A line isinstalled for an additional receiving station, not currently developed; the mechanical switchselects the receiving station. Idle sample transit line unions are capped to prevent intrusion offoreign material into the system. Three in core terminals are available for use in core positionG-34; the original, large capsule terminus, a small capsule terminus, and a cadmium lined smallcapsule terminus.Sample movement between the loading port and core terminal is provided by a motor-blowerassembly, and four valves for air flow direction control (components of the original GA PNTsystem). Gas flow is designed to recirculate within the system, with losses only at loadingstations or system connection points. The large and small systems have separate sample transitlines with a single gas supply and return line supporting boththe large and small sample transitsystems. Air displacement by CO2 gas reduces 41Ar production in the system. An air filter inthe flow system controls the amount of circulating particulates.The specimen capsule or "rabbit" is made of polyethylene. The effective available space insidethe capsule is 0.56 inch (14.2) in diameter by 3.95 inches (100 cm) in length giving a usablevolume of 0.97 cubic inches (15.9 cm3). The capsule is designed to pass freely in a tube with acurved section no smaller than 2 feet (61 cm) in radius and with an inside tube diameter nosmaller than 1.08 inches.Table 10.3 shows the pneumatic transit system dimensions. The A (Large) system is theoriginal General Atomics pneumatic terminal system. The B(Small) system is the modifiedsystem.Page 10-9 CHAPTER 10, EXPERIMENTAL FACILTIES AND UTILIZATION 12/2011Table 10.3 Characteristic Dimension of UT-TRIGA PTS..Transport System A (Large) B (Small)Terminal point OD 1.25 0.875Terminal point ID 1.085 0.685Terminal point tube Thickness 0.0825 0.095Transport tube OD (aluminum) 1.25 0.875Transport tube ID (aluminum) 1.12 0.745Transport tube Thickness 0.065 0.065Transport bends OD (polyethylene) 1.5 1Transport bends ID (polyethylene) 1.25 0.75Transport bends Thickness (polyethylene) 0.125 0.125Polyethylene transport capsule: 0.985 d X 4.75 I 0.650 d x 2.1S ITotal transport tube length (feet) 90 90Transit time (seconds) 6 6One terminal is an aluminum 6061 alloy with the associated radioactive nuclides. The otherterminal contains a cadmium liner (for thermal neutron filtering) in addition to the normalaluminum alloy radioactivity.The pool assembly consists of irradiation terminal and transport tubes to the pool surface. Poolassembly components are made of aluminum (alloy 6061). Tube connections in the pool are nutand ferrule type (aluminum Weatherhead) to seal against water leakage. The standardinstallation of the GA PNT design consists of aluminum and polyethylene tube. Straighttransport sections are 1.25 inch diameter (OD) aluminum (6061) tube. Transport bends are 1.5inch diameter (OD) polyethylene tubing with two-foot radius curves. Tube connections at theload port in the fume hood are also nut and ferrule type (stainless steel Swedgelock). Tubeconnections between aluminum and polyethylene transport sections are made with band stylehose clamps.Air lines for the transport system are made of 1.25 (OD) diameter aluminum tube for straightsections and 2.25 (OD) diameter flexible plastic hose for bend sections. All connections aremade with band style hose clamps.Large capsules are high-density polyethylene. High density capsules are reusable several times.Small capsules are fabricated from low-density polyethylene capsule without any reuse totransport the sample capsule.C. REACTIVITYCalculations and experiments show that the reactivity effects of the unlined pneumatic transitsystem are negligible and close to zero. The cadmium lined pneumatic transit system has areactivity of -0.21% Ak/k (-$0.30). Samples introduced to the pneumatic transit system arePage 10-10 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 10evaluated with regard to reactivity and must be less than the values stated in the TechnicalSpecifications.D. RADIOLOGICAL ASSEMENTThe pneumatic transit system is constructed of aluminum 6061 alloy. One terminal has anadditional cadmium liner. Activation calculations show similar levels to that of the centralthimble facility. The cadmium liner activated to 107Cd (6.52 day half-life), 10'Cd (461 day half-life), nlmCd (48.5 minute half-life), n3mCd (14.1 year half-life), n3Cd (7.7 x 1015 year half-life),n15mCd (44.6 day half-life), "nCd (2.228 day half-life), 17M Cd (3.4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> half-life), and n7Cd (2.49hour half-life). The Cd liner consists of two sheets of 0.020 inch thick sheets. They line theinterior of the irradiation terminal that has in inner diameter of 0.685 inches and a height of 20inches. This equates to 77.7 g of Cd utilized as a liner in the pneumatic transit system. Table10.4 lists the activity of the Cd liner after a 30 year irradiation at a flux of 1012 n cm-2 s-1 and a 1year decay. The dominant activity results from 109Cd. With a half-life of 464 days, 109Cd couldbe allowed to decay on-site for a number of years prior to disposal.Table 10.4: Activation of Pneumatic Transit System Cadmium LinerIsotope Activity (Ci) Half LifeCd-107 0 6.490 hCd-109 0.04173 464.0 dCd-lllm 0 48.5 mCd-113m 0 14.1 aCd-113 13.33e-15 7.7e15 aCd-115m 0 44.6 dCd-115 0 53.46 hIn-li5m 0 4.486 hIn-115 4.252e-15 5.100e15 aCd-117m 0 3.4 hCd-117 0 2.49 hIn-117m 0 116.5 mIn-117 0 43.80 mSn-117m 868.4e-15 13.61 dSample activation levels are assessed on an individual basis.E. INSTRUMENTATIONInstrumentation supporting the pneumatic transit system includes a control system (located inboth the control room and in the laboratory associated with the system) and a radiationmonitor in the fume hood near the end terminal. The control system allows the system to bePage 10-11 CHAPTER 10, EXPERIMENTAL FACILTIES AND UTILIZATION 12/2011turned on and off, includes manual and automatic send/retrieve controls, and is attached to atimer. The radiation monitor assesses the activity of samples irradiated in the pneumatictransit system and the readings are displayed in both the laboratory and the control room. Analarm is set to warn experimenters and reactor operators if a high activity samples aremeasured.F. PHYSICAL RETRAINTS, SHIELDS, OR BEAM CATCHERSNo special restraints or shields are in place for the pneumatic transit system. The transit linehas a bend to prevent streaming. The in-core facility utilizes the same shielding that is in placefor the reactor core. Shielded areas are available in the laboratory for sample deposition afterirradiation.G. OPERATING CHARACTERISTICSThe unlined pneumatic transit system may be operated at any licensed power level. However,the cadmium lined pneumatic transit facility is limited to a power of 500 kW due totemperature constraints. This limit is to prevent the polyethylene sample rabbits fromsoftening in the facility and becoming fixed in place. Temperature measurements in theterminals at 500 kilowatts and 950 kilowatts were made with a thermocouple. Approximately30 minutes is required to create steady-state temperatures. Peak temperatures in the standardterminals are 52.5°C and 72 °C at the two respective power levels. Higher temperatures of 83°Cand 120 °C occur in the Cd version of the irradiation terminal.Neutron flux measurements with gold foils and three threshold foils were made to characterizethe facility. Results of the measurements. are shown in Table 10.5 and demonstrate theoperational difference between the two irradiation terminals. Absorption of neutrons by the Cdliner changes the cadmium ratio for a sample from 5.06 to 0.99.Table 10.5: Flux Measurements in Pneumatic Transit System at 100 kW(n cm- 's-)Thermal Epithermal Cd RatioNo cd 7.8 x 10"l 1.3 x 10"° 5.06W cd 1.80 x 10' 1.1 x 101° 0.99H. SAFETY ASSESSMENTAir displacement by CO2 gas reduces 41Ar production in the system. An air filter in the flowsystem controls the amount of circulating particulates. As a result, operation of the systemwithout samples causes a minimal radiological risk. Samples need to be evaluated on a case bycase basis. In the event of a sample with unexpected high radiation levels, a radiation monitorwith an automated alarm will alert experimenters.Page 10-12 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 10With regard to nuclear reactivity, the facility itself is well within Technical Specificationrequirements. Calculations and measurements on routine samples show reactivity levels lessthan 0.035% Ak/k ($0.05).10.2.2.2 Three Element IrradiatorA. DESCRIPTION.The three element facility is typically used to generate radioisotopes for research orneutron activation analysis.The three element experiment facility displaces thee fuel elements. The three elementfacility consists of modifications to the upper and lower grid plate, a fixture for aligning andmanipulating the three element canister, and the three element canister. Since the bulk ofthe upper grid plate supporting the position of three fuel elements is removed, an adapter isrequired to position fuel elements the facility is not in use. The three-element facility isdesigned to be rotated (either manually or motor driven) to minimize spatial variations influence when required, using a reach-rod or other attachment extended to the bridge.The facility requires ballast in the form of a metal liner. A lead liner is used for a normal,predominantly thermal neutron irradiation. A cadmium liner is used when reducedthermal neutron flux and enhanced epithermal irradiation is desired.B. DESIGN & SPECIFICATIONS.B (1) Upper and Lower Grid Plate Modifications. The upper grid plate has two positionswhere a three element irradiation canister can be inserted. The positions are fabricatedby machining a 2.062 in. (0.052375 m) diameter hole in the upper grid plate centered at acenter point between three fuel elements. A hole is fabricated in the lower grid platecentered on the three fuel element positions for alignment.The alignment fixture is composed of two plates (that interface with the upper and lowergrid plates) attached by rods. The lower plate is a disk with lobes corresponding to each ofthe three fuel element positions. A pin extends through the plate. On the bottom, the pinfits into the centered-penetration in the lower grid plate previously described. A recess inthe bottom of the three element canister fits over the pin in the top of the plate. The plateacts as a bearing surface for rotation of the canister. The upper plate is roughly triangularwith truncated apexes, and is machined into two separate thicknesses. The thicker, centersection of the upper plate has extrusions that mater with vacant fuel penetration holes inthe upper grid plate around a center hole for insertion of the canister. The triangular sectionPage 10-13 CHAPTER 10, EXPERIMENTAL FACILTIES AND UTILIZATION 12/2011extends over fuel positions adjacent to the three element vacancy, and circular cutoutsprovide clearance for adjacent fuel element cooling channels. Additional holes are drilledaround the central hole to provide cooling flow for the three element canister.B (2) Alignment Frame. The three element facility uses an alignment frame that fits into thecore grid location. The alignment frame provides position control, vertical and lateral support,of the irradiation canister. Components of the grid alignment frame consists of the base plate,an alignment pin for the canister, three vertical rods for the frame structure, a top plate for theplacement of the irradiation canister, and a fitting for use when the canister is not present. Thethree element assembly rests on the lower grid plate, and is ballasted to be negatively buoyant.The submerged weight of the three element facility is less than the weight of the threeelements it displaces. Although theoretically the all of the three element space could be fullyoccupied by sample material, flux depression considerations prevent such usage.Structure rods of the alignment frame prevent the irradiation canister from contacting theadjacent fuel elements during insertion and removal of the irradiation canister into the frame.A 0.5-inch diameter pin.at the base of the frame aligns the irradiation canister and provides abearing for the rotation of the canister. The rods are welled into the upper and lower plates. Atthe top of the frame is a 2-inch diameter hole within which the canister rotates. Coolant holesin grid alignment frame provide for cooling of the irradiation canister. A closure fitting is placedon the irradiation assembly frame when a tube is not in place. This fitting minimizes coolant bypassing the fuel and prevents inadvertent reactivity insertion into the three-element gridlocation in the reactor core.The three element facility positions are in fuel element positions D-05, E-06 and E-07 and fuelelement positions D-17, E-22, and E-23. The three element facilities are isolated from potentialcontrol rod positions by at least one fuel element position from traditional positions for thepulse and regulating rods, and two fuel rods in the case of the shim rods. One three elementfacility is two elements from the outer edge of the core, the other is one fuel element from theouter edge of the core.The D05/E06/E07 three-element facility is close to the radial extension form the center to apower level channel. Experiments have demonstrated that the facility is sufficiently isolatedfrom the leakage neutron path reaching the detector as to not excessively affect the powerlevel signal.The D17/E22/E23 three-element facility lies in a quadrant of the core between two power leveldetectors, is closer to the core center, and is sufficiently isolated as to have a minimal effect onleakage neutrons.B (3) Three Element Facility Canister. The facility uses a sealed canister with a usablespace 1.527 in. (0.038786 m) in diameter. A component assembly diagram is provided inFigure 10.4; a rod with an end fitting similar to a fuel element is secured to the top cap forPage 10-14 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 10handling with the fuel tool, and a rod with a tapered end is secured to the bottom foralignment in the lower grid plate penetration. The three element canister outer diameteris 1.875 in. (0.047625 m). The canister wall is 0.1 in. (0.000254 m). The inner liner is 1.625in. (0.041275 m) outer diameter, and 1.527 in. (0.038786 m) Inner diameter. Overall lengthfrom the bottom of the canister to the top of the threaded fitting at the top of the canister(i.e., excluding the handling and alignment pins and the end cap) is 50.375 in. (1.279525 m)with the length of the usable volume 48.125 in. (1.22375 m).A threaded cap for the top fitting contains two o-ring seals, a pressure relief valve, a gas valveor vent port, and an attachment anchor for remote handling of the canister. Seals for theprotection of both expansion and compression pressures in the canister consist of two o-ringseals, one a radial seal and one an end seal. The double seal design should provide extraprotection against water leakage into the canister. Two holes in the cap allow venting andpurging of the canister gases. One cap hole is the vent line. The other hole contains a pressurerelief valve set at a differential pressure of about 2 psig. During sample irradiation the positionof the canister is at a depth of about 20 feet of water. At the irradiation position the canisterpressure with 20 feet of water will change about 12 psig relative to the loading condition at apressure of one atmosphere. A threaded hole at the center of the canister cap is for theattachment of a canister-handling device. The type of attachment rod utilized depends oncanister handling requirements. One type of attachment is a rod with a fitting for remoteattachment with the fuel-handling tool. Routine movements of the canister in the reactor pooland core can then be made with the fuel-handling tool.When the facility is in use, lobes of the vacated fuel element position are open. Thegeometry of fuel elements surrounding the three-element facility causes significant potentialfor variations in exposure based on the position of the material to be irradiated. Thereforethe capability to rotate the canister was designed into the facility alignment fixture.The facility is ballasted with either lead or cadmium. Ballast of approximately 0.0625 in.(0.001588 m) thick is placed between the canister and an inner liner. The liner layer of Cd orPb wraps twice around the internal aluminum tube, extends almost the full length of thecanister, about 46.75 inches, and includes an equivalent end disk at the bottom end. Each layerof the cadmium or lead liner folds over the bottom disks. The two vertical layers of thecadmium liner overlap-while the two vertical layers of lead do not overlap.With the exception of the ballast, the three element facility is manufactured from aluminum6061 alloy. Activation of the aluminum components are expected to be similar to the specificacidity described for the central thimble, except that (1) the three element facility does nothave the previous irradiation history from the earlier UT Austin Mark I reactor, and (2) the threeelement positions are at lower flux positions as compared to the central thimble. The lead andPage 10-15 CHAPTER 10, EXPERIMENTAL FACILTIES AND UTILIZATIONI12/2011cadmium used as ballast in the three element facility are at least 99.9% pure. Neither the leadnor the cadmium has potential for significant chemical activity in contact with aluminum.T w 711 r 0:V ý dVA Mor X~ Un 2 'I-. .. ...... .....4.-//Figure 10.4: Three Element Irradiator--A~C. REACTIVITYRemoval of three fuel elements for placement of the three element irradiation has a significanteffect on core reactivity. The reactivity change has been measured with a control rod bankconfiguration and with a configuration of one and two control rods full out. The average changein reactivity of the configurations to remove the three fuel elements was $2.30 with a minimumof $2.08 and maximum of $2.47. When three fuel elements are removed for placement of thethree element facility, recalibration of control rod worth curves is necessary.Experiments with the three element irradiator canister require that it remain in the core duringoperation. Insertion of the irradiator into the core or out of the core must be conducted whenthe reactor is in a shutdown state. However, the facility may be rotated in place duringirradiation. The reactivity effect of canister rotation has a non-measurable effect on reactivity.Only a redistribution of the liner absorbing material is capable of causing the rotationalreactivity to change. Unless accident conditions such as mechanical or thermal damageredistribute the neutron absorber materials, the rotation reactivity will remain effectively zeroto within a few cents. Estimates of the three element irradiation canister reactivity were madeprior to initial tests of the canister. Some of these reactivity estimates were made frommeasurements with similar equipment at another research reactor facility and includeextrapolation of measurements made on similar experiment components such as the twoirradiation terminals of the Pneumatic Transfer System.Page 10-16 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 10The reactivity limit for a single moveable experiment must be less than $1.00. As a moveableexperiment the three element irradiation canister must be less than $1.00 of reactivity to meetthis constraint of the present Technical Specifications. Classification of the three elementexperiment system as a moveable experiment would be a conservative condition since the totalreactivity of the canister will occur only during an insertion, removal, or an unknown type ofaccident that occurs with the reactor at critical conditions. The total available reactivity changeof the three element irradiation canister will not occur with the reactor at critical conditions.C (1) Reactivity CalculationEstimates of the experiment facility reactivity were insufficient to determine the operatingrequirements for the T3 canister with a neutron absorption liner made of Cd. Severalcalculations were done to develop the final design constraints for the neutron absorption liner.Measurements with the fi nal design were made prior to acceptance of the irradiation system.Calculations with MCNP(4a), a Monte Carlo particle transport computer code, were made todevelop a better evaluation of the canister component reactivity. Previous test measurementsand several test core configurations were useful to benchmark the calculation with themeasurements. Agreement of the benchmark measurements and MCNP( 4a) calculations wereadequate to pursue installation and test of the three element irradiation canister. Theirradiation canister analysis focused on the most significant reactivity conditions that occur withvarious configurations of the installation of the cadmium liner version of the system.Development of the MCNP( 4a) analysis proceeded in three steps. The first step was acalculation of several reactor core conditions for which measurements were available tocompare the experiment and calculation results. The second step was an analysis of theirradiation canister reactivity with a full-length liner of neutron absorber and a short versionwith a six-inch long neutron absorber. A final step was a calculation of the most plausibleaccident condition that is flooding of the irradiation canister volume with water.Calculations project the total three element irradiation canister reactivity worth will change by$1.08 as the absorbing liner changes from a zero-length liner to a full-length liner. Calculationerror is as much as 10 to 15%. Although this result exceeds the $1.00 constraint the calculationof the net reactivity available from insertion and removal of the system with the liner does notexceed the limit. Calculations indicate that the three element irradiation canister without anyneutron absorbing liner will create a positive reactivity of $0.16. This condition represents thecompetitive process of neutron leakage from the core and neutron moderation and absorptionby the water in the location of the canister.The goal of the MCNP canister calculation was to determine whether the full length Cd liner inthe canister would exceed the conservative constraint of $1.00 for the system worth as amoveable experiment. Initial test measurements in the core did not support the less than $1.00Page 10-17 CHAPTER 10, EXPERIMENTAL FACILTIES AND UTILIZATION 1 12/2011conclusion. The MCNP calculation predicts the reactivity worth of the irradiation canister with afull-length Cd liner will be less than one dollar. The canister reactivity with Cd liner reactivitywas -$0.89 +/- $0.12.A flooding accident with the canister in the core will decrease reactivity by increasing neutronabsorption. The MCNP result for the flooded canister condition calculates the negativereactivity change by $0.56 to -$1.45 +/- $0.12. Flooding of a canister with a neutron absorptionliner will exceed the $1.00.C (2) Reactivity MeasurementsTwo measurements of the reactivity of the three element irradiation canister with the full-length cadmium liner found the reactivity worth was -$0.92 with the control rods in a bankconfiguration. Measurements of the core reactivity were also made with two conditions of thecontrol rods full out. Control rod configuration measurements both decrease and increase thecanister worth in the range of $0.89 to $0.96. In the flooded condition the canister negativereactivity worth increases by $0.24 to -$1.16. Extreme positions of the control rods do notsignificantly change the flooded condition result. These measurement results are consistentwith MCNP calculations for the two canister non-flood and flood conditions.D. RADIOLOGICAL ASSESSMENTActivation of aluminum 6061 was discussed in the section describing the central thimble.An average neutron flux was calculated based on a nominal value of 2x1011 n cm2 s1 with anassumed irradiation at 2 MW, 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per day each week, 11 months each year. With anaverage neutron flux of 4.37x1010 n cm2 s-1 irradiation over 40 years followed by 1 week ofdecay, 61.6 pCi per gram of lead 205 is produced.A similar irradiation of cadmium produces the activities noted in Table 10.6.Table 10.6: Activity of Three Element Irradiator Cd LinerIsotope Activity Half Life 1 m Dose RateCd-107 1.274 pCi g-1 6.490 hCd-109 40.10 pCi g-1 464.0 d 7.5 VR h1 g-1Cd-113 10.00e-18 Ci g1 9.300e15 aCd-115 69.93 ItCi g-1 53.46 hIn-115m 76.34 pCi g' 4.486 hIn-115 3.189e-18 Ci g1 5.100e15 aSn-117m 40.97 nCi g' 13.61 dIf the canister is filled with air, 41Ar may be produced. Assuming Argon is 1.28% of the mass ofair, with the mass of air as 1.3 kg m-3.Irradiation and decay under the same scheme abovePage 10-18 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12;2011SAFETY ANALYSIS REPORT, CHAPTER 10followed by release to the reactor bay results in an atmospheric activity concentration of 128.3ItR h-1 m3 resulting from 41Ar.Based on 50.374 length of the air volume in the canister at 1.527 in. diameter, the canistervolume is 0.001512 M3.It should be noted that the neutron flux value utilized in this calculationis the maximum possible in the reactor (neutron flux is about a factor of five less at the threeelement irradiator position), is further reduced by the ballast (lead or cadmium), and notconstant across the container.To minimize the potential for the production of 41Ar the canister is flushed with dry nitrogenprior to insertion into the reactor.E. INSTRUMENTATIONThere is no instrumentation associated with the three element facility. Instrumentation thatmight be used as part of an experiment program will be evaluated as part of the experimentreview and approval process.F. PHYSICAL RESTRAINTS, SHIELDS, or BEAM CATCHERSNo special restraints or shields are in place for the three element facility. The facility is entirelyunder water during irradiation with no possible radiation streaming outside the reactor pool.The in-core facility utilizes the same shielding that is in place for the reactor core. Shieldedareas are available in the reactor bay area for sample deposition after irradiation.G. OPERATING CHARACTERISTICSThe three element irradiator is a widely utilized facility for in-core irradiations. The lead linedcanister is utilized for thermal neutron irradiation at powers up to the maximum licensedpower. The cadmium lined facility is utilized for epithermal neutron activation experiments atpower levels up to 500 kW. Irradiations are conducted by loading the three element irradiatorinto the core when the reactor is in a shutdown system. The facility may be rotated duringirradiation, but is not inserted or removed during reactor operation.H. SAFETY ASSESSMENTH (1) CoolingGrid holes beneath each fuel element are the coolant flow source for each fuel channel. Aprovision has been made to also provide coolant channel flow by water convection around thethree element canister assembly. The core grid frame contains two holes for each of the threePage 10-19 CHAPTER 10, EXPERIMENTAL FACILTIES AND UTILIZATION 1 12/2011fuel element positions that make up the experiment facility. Six holes in the grid frame bottomplate provide coolant flow to the three element canister assembly. The bottom fitting of thethree element canister contains fins to enhance the heat transfer to the coolant. Coolant flowspast the cooling fins along the length of the three element canister assembly. Six holes in thetop plate of the core grid frame provide an exit path for coolant flow around the assembly.Generation of heat by the three element canister is substantially less than that of the adjacentfuel element channels. Thermal neutron reaction rates in the neutron absorption liner are asubstantial source of heat. Cooling of the three element canister is an important designconsideration to protect canister components, specifically samples or materials, from thermaldamage. An estimate of the potential temperatures in the three element canister was found byexamination of the measurements made with the two PTS irradiation terminals.H (2) TemperatureThe physical design of the cylindrical irradiation canister with internal aluminum cylindricalinsert provides a 0.072-inch gap. The cylindrical gap prevents the mechanical rearrangement ofthe absorber material. Thermal redistribution of the materials depends on the melting point forthe three materials of the irradiation canister. The irradiation canister is made of 6061aluminum alloy. The aluminum has a melting temperature of 660 TC. By comparison the linermaterials of lead and cadmium have much lower melting temperatures of 327 and 321 'C,respectively. Reactor fuel elements at nominal conditions of full power operation producemaximum fuel temperatures of roughly 450 'C with a respective element cladding temperatureof about 140 'C. Heat from neutron activation reactions in the lead or cadmium liner materialwill produce higher temperatures in the irradiation canister than that of a canister without theliner. Experiments with the pneumatic transit system irradiation terminals found the aluminumterminal without a cadmium liner to have a 500-kilowatt temperature of 54 °C and a 950-kilowatt temperature of 72 *C. The aluminum terminal with a cadmium liner was found to havea 500-kilowatt temperature of 85 'C and a 950-kilowatt temperature of 120 0C. Experimentswith low-density polyethylene demonstrate that some material deformation begins at atemperature of 90-95 *C. The temperature limit recommendation for continuous use ofpolyethylene is a function of the polyethylene density and ranges from 60 to 200 °C. The testlocation for the pneumatic transit system irradiation terminals was in the reactor core G-ringand neutron fluxes are a factor of 1.8 less than those measured for the three elementirradiator. Thus, an equilibrium temperature adjustment by a factor of 1.8, assuming all otherheating and cooling factors remain about the same, can be made for the neutron flux differencebetween the three element irradiator location in rings D and E and the G-ring location of thepneumatic transit system. Estimates of the potential irradiation canister temperatures indicatethat the temperatures will not approach the melting temperatures of the lead or cadmiummaterial. The equilibrium temperatures that occur at one hour at full power could exceed 200°C. These temperatures may cause damage to polyethylene sample capsules and othermaterials that are irradiated in the canister.Page 10-20 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 10H (3) PressureAir pressure relief for excessive pressure buildup in the canister is a design feature to protectthe canister from rupture.Yield stress for the T6 6061-aluminum alloy of the irradiation canister is 30,000 psi. A limit forthe canister operating pressures has been set at 250 psi. This limit includes a safety factor oftwo and a strength reduction for the heat treatment from T6 to TO. Design of the top fittingcontrols the pressure with a double O-ring seal and two 1/8 inch valves, a pressure relief valveand a manual fill valve.Temperature changes on the three element canister during irradiation and the evolution ofgases from experiment materials in the canister will change the ambient pressure. A relief valvehas been chosen with a set-point of two to three psi. At pressures less than the setpoint thecanister gas inventory will remain constant. A double O-ring seal protects against leakage intothe canister. As a constant volume device the canister pressure is readily found from the gaslaw, PV=nRT. The number of moles of gas, n, the volume, V, and the gas constant R are allconstants. For the purpose of the analysis the canister to liner gap is 20 cm3 and the canistervolume is 2400 cm3 .At the operating depth of the canister the external pool water pressure is10 psi. The differential pressure at the relief valve must exceed the pressure due to water andthe pressure setting of the valve. During normal canister operation a change of the airtemperature from 300 K' to 350 'K will increase the pressure in the canister by 2.45 psi orabout 5 psi per 100 'C. This pressure increase will go to zero as the canister cools following anirradiation.A change in the number of moles of gas in the canister could also occur. Two source conditionscan occur that will increase the gas content in the canister. These potential sources of gasproduction are vaporization of the water in liquid samples and the evolution of gas by radiationof the polyethylene. Other sources may be present if volatile materials are part of theexperimentEvaporation of water by heating vials of liquid samples will create a total change of 1 cm3 ofliquid to gas. Conversion of 1 cm3 of water to gas produces 1000 cm3 of gas. The resultantcanister pressure change is +8.18 psi per cm3 of water vaporized assuming it is distributed overthe entire canister volume. The pressure increase should neutralize following cooling of thecanister. Irradiation of hydrocarbon materials has the potential to produce 0.1 cm3 of gas pergram per megarad. The release rate for polyethylene capsule materials is much less, 0.02cm3per gram per megarad. If the fast neutron and gamma ray dose in the canister is 1,500megarad/hour at 1 megawatt the potential gas release from the polyethylene capsules is 30cm3 per gram or about 750 cm3 for an irradiation of 25 sample capsules in a two hour 500kilowatt operation. This gas production represents a pressure increase of 4.6 psi. This is not aPage 10-21 CHAPTER 10, EXPERIMENTAL FACILTIES AND UTILIZATION12/2011significant pressure change in the canister although it may cause the canister to vent all thepressure that exceeds the relief valve setting. If sample materials in the capsules arehydrocarbon materials the pressure could be five times greater.Most of the gas release in the breakdown of polyethylene and other hydrocarbon materials ishydrogen. A purge of the canister atmosphere prior to irradiation with carbon dioxide ornitrogen gas will reduce the available oxygen and eliminate the air activation of argon.H (4) LOCA potentialThe canister is completely submerged during irradiation, and does not offer any leakage pathfor pool water.10.2.2.3 6/7 Element IrradiatorA. DESCRIPTIONThe 6/7 element irradiator is a large in-core facility to perform neutron irradiations. It islocated in the seven-element cutout in the top grid plate of the reactor as shown in Figure 10.2.The facility may be placed in the middle of the core removing 6 fuel elements and the centralthimble or it may be placed in the location that overlays part of the outer three fuel rings. Ithas largely been utilized for irradiation of circuit boards and irradiation of samples for neutronactivation analysis.B. DESIGN AND SPESIFICATIONSThe irradiation can is composed of 6061-T6 aluminum and contains a 0.08 in (2 mm) thickborated aluminum (B) liner. The inner diameter of the irradiation can is 2.25-in. The boronconcentration is 4.5% by weight in the 1100 series aluminum alloy. The boron is enriched togreater than 95% 10B, which is the boron isotope with a high thermal neutron cross-section. Thedesign of the irradiation can is very similar to that of the cadmium and lead lined three elementfacilities described above. The total height of the facility is approximately 52 inches. This heightis intended to elevate the stainless steel fittings, a purge valve, and a relief valve above thereactor top grid plate and thereby minimize activation of these components.The second component is a separate, hollow lead cylinder that is clad with 6061-T6 aluminum.This lead sleeve surrounds the main irradiation can. 6061-T6 aluminum is once again used forthis component. The sleeve resembles a thick, hollow cylinder. The outside diameter of theirradiation can is slightly smaller than the inside diameter of the sleeve. When inserted into themiddle of the sleeve, the can rests on a pin that is connected to the base of the sleeve. This pinhas been designed to accept the 3L facilities previously mentioned. The pin assembly alsoincludes six holes to allow pool coolant to pass through the center of the sleeve. A small gapPage 10-22 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 10exists for the coolant to pass between the can and the sleeve when the can is being used. Threepegs have been built into the top of the sleeve which center the irradiation can when it is inplace. The sleeve has been designed to be removable. An eye bolt attached to the top of whatresembles the handle of a bucket is used to raise and lower the sleeve.The connector box is a small, aluminum can which sits approximately 3-ft above the irradiationcan. The can and box are connected by an aluminum tube. The tube is for passing electricalwires from the box into the irradiation can. The box is designed to allow for electricalconnectors to mate on its inside which isolates electrical wire that is not activated by wire thathas been activated during irradiation. From the top of the box extends Tygon tubing to passthrough the remainder of the electrical wires to the top of the pool.C. REACTIVITY.A MCNPX model of the TRIGA was used to calculate the reactivity of the seven element facility.The base calculation had the seven element location empty with the three element locationfilled with fuel. The reactivity effect of the change from the fuel configuration with the threeelement irradiator to the seven element fuel configuration is -$1.28. The perturbation causedby the addition of the lead sleeve is +$0.08. When adding the irradiation can to the assembly,the total experiment worth is $0.25. Therefore, the reactivity of the experiment is far less than$1.00, Reactivity worth of individual experiments in the facility have to be evaluated on anindividual basis.D. RADIOLOGICAL ASSESSMENTFrom a radiological perspective, the seven element irradiator is similar to three elementirradiator. However, the seven element irradiator does not have a cadmium liner thatactivates. In its places is boron (95% '0B) liner. The primary absorption reaction is 1°1 + In 4 7Li+ 4ca. This reaction does not have a radioactive product, so activation hazards from the boronare minimal. Aluminum activation is similar to that of the other facilities analyzed.Experiments within the seven element irradiator require analysis on an individual basis.E. INSTRUMENTATIONThere is no instrumentation associated with the seven element irradiation facility.Instrumentation that might be used as part of an experiment program will be evaluated as partof the experiment review and approval process.Page 10-23 CHAPTER 10, EXPERIMENTAL FAClLTIES AND UTILIZATION F. PHYSICAL RESTRAINTS, SHIELDS OR BEAM CATCHERSNo special restraints or shields are in place for the seven element irradiation facility. Thefacility is entirely under water during irradiation with no possible radiation streaming outsidethe reactor pool. The in-core facility utilizes the same shielding that is in place for the reactorcore. Shielded areas are available in the reactor bay area for sample deposition afterirradiation.G. OPERATING CHARACTERISTICSOperation of the seven element facility for electronics damage facility is at 1 kW of power orless. The facility allows for electronics to be powered during irradiation through a curved watertight tube going to the pool surface. The facility allows for direct monitoring of electronics as itundergoes irradiation and reaches a point of failure.H. SAFETY ASSESSMENTH (1) Temperature (Fuel)Fuel temperature measurements at 1-kW show the fuel temperature to be +/- 1 °C of the poolwater temperature, which was recorded as 20.7 *C. As the reactor is operating at 1-kW, themaximum temperature in anyone fuel pin in the reactor is significantly below the maximumallowable temperatures for the outside clad temperature of greater or less than 500 °C.H (2) Temperature (Lead)Calculations were performed to ensure that the lead in the sleeve would not reach near meltingtemperatures even at a reactor power of 1 MW. The temperature was calculated to be lessthan 40 °C with a coolant inlet temperature of 25 °C and an inlet velocity of 0.15 m s-1. Themelting point of lead is 325 *C. A collision heating (+F6) tally was utilized with the MCNPXmodel to determine energy deposition in the lead sleeve. Since the temperature increase wasso minor, thermal expansion of the lead and aluminum clad are neglected. Additionally, a 1/16inch gap was added into the design as the distance between the outside edge of the sleeve andthe hole in the tap grid plate to prevent the sleeve from becoming stuck in the tap grid plate.H (3) Pressure (irradiation Can)Through the aluminum tube and Tygon tubing, the irradiation can is open to atmosphere.Therefore, no internal pressurization will occur..Page 10-24 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 10H (4) Pressure (Lead Sleeve)The lead sleeve consists of two aluminum tubes that are attached together by two end caps.The lead was not poured within the two tubes completely to the top the sleeve allowing for anair gap. Since the temperature does not rise within the fuel and the energy deposition in thelead is so small, the pressurization of the air within the lead sleeve is negligible and not a risk.H (5) MassThe lead sleeve weighs less than 60 pounds. The seven elements that the sleeve replaces weighapproximately 56 pounds. The weight of the lead sleeve is distributed as one single, circulararea of 3.874-inches in diameter whereas the weight of each of the fuel elements is distributedon a much smaller area of the grid plate. The irradiation can and connector box are slightlynegatively buoyant and do not contribute a significant amount to the total additional mass ofthe system. The mass of the lead sleeve, irradiation can, and connector box are not a risk.H (6) StructuralBoth lead sleeve and irradiation can are at risk for being dropped onto the top of the corewhich could cause structural damage to the reactor. To minimize the risk, both sleeve and canare lowered as closely to the side of the pool wall as possible before being maneuvered overthe reactor at the height of the top grid plate. The sleeve is stored on the underwater tablewhen not in use and is tied to the top of the pool to keep it from toppling over. Likewise, theirradiation can is stored on the inside of the reactor pool and tied to the top of the tank forstorage.The reactor power is no more than 1 kW for electronic component testing. There is nonoticeable increase in the fuel temperature at this power level above the bulk pool, watertemperature. With no increase in temperature and both the coolant pump and the diffusernozzle off, there is no flow through the core and no risk for flow blockage. At thesetemperatures, there is no risk for phase change of coolant either.All of the components of the lead sleeve, irradiation can, and connection box are fixed togetherby aluminum welds or tube fittings. The risk for any component of these parts to separate andbecome a hazard is negligible.All of the materials in this experiment are sealed water-tight either by welding, fasteners,gaskets, or a combination of these methods. Each of the components (lead sleeve, irradiationcan, and connector box) are leak tested prior to being utilized for any experiment requiring thereactor. Therefore, any part of the electronic components under test have no interaction withPage 10-25 CHAPTER 10, EXPERIMENTAL FACILTIES AND UTILIZATION 1 12/2011the reactor that would cause any material hazard. No hazardous chemicals are used in thisexperiment or materials that are flammable.10.2.3 Rotary Specimen RackA. DESCRIPTIONThe rotary specimen rack (RSR) is used to support neutron activation analysis and isotopeproduction. The rotary specimen rack consists of an air-filled water-tight canister enclosing asample rack and pinion drive assembly attached to a sample rack. The sample rack isassembled from an upper and lower ring attached to tubes. A ring-drive and an indexingmechanism allow samples to be placed in each position. The pinion gear drive shaft housing isa dry tube from the pool bridge to the rotary specimen rack housing.Sample vials are inserted and removed through curved dry tubes. Curvature minimizesradiation leakage through the dry tubes. Both a manual and an automatic dry tube areinstalled, but infrastructure supporting use of the automatic dry tube has not been developed.An electro-mechanical operator attached to a cable is available to support insertion andremoval of sample vials. The cable is coiled on a spool operated with a reel. The automatic drytube is designed to use pneumatic pressure to remove and insert samples.Rotation can be performed manually or with an installed drive motor, powered from the samesource as the pool lights. Rotating samples during a long irradiation evenly distributes theneutron fluence received by each sample.B. DESIGN SPECIFICICATIONSThe RSR housing is a cylindrical canister with an internal diameter of 22 in, and an outerdiameter of 27 2/7 in. Specimen positions are 1.23 in. (3.18 cm.) in diameter by 10.80 in. (27.4cm.) in depth. Figure 10.5 illustrates the RSR which basically forms a ring outside the reactorcore. There are ports for loading of samples as well a drive shaft for rotating the samples.t ..~~~~~ ... ..... .. ..... ... .. ... ..........(. "Lifting ... .. liftingI ie.1 .0. .... ......... .... ..S. .............m. ..... .a............................... ghrat md":. ~~~ ~ ~ .... ,<..<' ',.....* "" ... ./Figure 10.5 Rotary Specimen .Rack DiagramPage 10-26 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 10The RSR contains raceways that are supports for sample rotation as illustrated in Figure 10.6.These raceways are manufactured from titanium forgings. There are two concentric racewayswith a ball bearing assembly interface.The inner raceway has an inner diameter of 22 inches, with an outer diameter of 24 /2 inches.The inner raceway is manufactured by welding a 0.38 inch tall by 1 ' inch wide ring (ID 22, OD24 Y2) to a 1.12 inch tall by 0.56 inch wide ring (23.12 OD, 22 in ID). Ball bearings (0.045 in.radius) are spaced by four cylindrical titanium separators; separators in contact with thebearings are slightly smaller than the center separators.The outer raceway provides the second bearing surface and supports the specimen tubes. Therouter raceway is a ring 1.88 inches tall by 5 5/8 inch wide (21 Y2 inch ID by 27 7/8 inch OD).The bottom section of the ring, supporting the specimen tubes, is 0.50 inches tall. There are 40holes supporting specimen tubes 1.38 inches in diameter equally spaced on a 26.312 inchdiameter circle. The upper section is formed from a ring 2 5/8 inches thick (24 1/8 in. OD by 21Y2 in. ID) to accept a spur gear. A spur gear is secured to the top of the outer raceway.Table 10.7: Rotary Specimen Rack GearsItem Spur PinionTeeth 200 10Width 0.5 0.5Pitch 23.873 1.194Pressure angle 200 200Center Distance 12.5335 12.5335Gear OD 23.992 1.550Gears are used to drive the RSR rotation mechanism. These are fabricated from aluminum60601 T-6. Gear specifications are provided in Table 10.7.The overall length of all specimen tubes is 11.44 in., with the wall thicknesses of 0.058 in. Thetop of the tubes are flared 450 to 1.62 OD. Position 1 is in two sections. The top section is 5.5in. tall, with 1 3/8 in. OD. The bottom section OD is 1 in. Positions 2 through 40 have an OD of1 3/8 inches. The bottom of the cylinder is penetrated by 3 Y2 inch holes in the wall at 120'intervals. The bottom of tube is terminated with a ring 0.06 inches thick that has a / inchcentered hole.Figure 10.7 illustrates the RSR rotation control box. The RSR position for loading is indicated inthe index dial. Controls are available for manual RSR rotation of automated sample rotation.The direction of automated sample rotation may also be set.Page 10-27 CHAPTER 10, EXPERIMENTAL FACILTIES AND UTILIZATIONI12/2011r !,/ //7~/~ ~k~<~ d~ /j~I' -\ <Figure 10.6: Rotary Specimen Rack Raceway GeometryIndexManual RotationHandleMotorRotationPowerMotorRotationDirectionClutchFigure 10.7: Rotary Specimen Rack Rotation Control BoxC. REACTIVITYThe RSR is located outside the reactor core. Along with the graphite reflector and water, theRSR facility affects the reflection of neutrons back into the reactor. However, the facility doesnot largely affect reactivity due to its proximity to the reactor core. Reactivity worth ofindividual experiments need to be assessed on an individual basis.D. RADIOLOGICAL ASSESSMENTThe neutron flux at full reactor power within the RSR facility is c.a. 1 x 1012 n cm-2 s-1. As suchactivation rates are less than the three element and seven element facilities analyzed above.The facility does not have a cadmium liner like the three element irradiator, so there is nocadmium activation hazard to assess.Page 10-28 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 10 .E. INSTRUMENTATIONThere is no instrumentation associated with the RSR facility. Instrumentation that might beused as part of an experiment program will be evaluated as part of the experiment review andapproval process.F. PHYSICAL RESTRAINTS, SHIELDS OR BEAM CATCHERSNo special restraints or shields are in place for the RSR. The facility is entirely under waterduring irradiation. The sample loading tube has a bend to prevent streaming. The in-corefacility utilizes the same shielding that is in place for the reactor core. Shielded areas areavailable in the reactor bay area for sample deposition after irradiation.G. OPERATING CHARACTERISTICSThe RSR is commonly operated for neutron activation and isotope production experiments.Operation during irradiations is typically in the range of 100 kW to 1 MW. The facility has astrong thermal component to the neutron flux and is utilized for thermal activation. Multiplesamples are inserted for simultaneous irradiation. Sample removal is often hours after theirradiation is finished to allow for decay of short-lived radionuclides.H. SAFETY ASSESMENTThe RSR facility is external to the reactor core and physically isolated from the fuel. The sampleloading tube goes to the pool surface and would prevent over pressurization of the facility.Radiological effects and reactivity effects of samples need to be assessed on an individual basis.10.3 Beam PortsA. DESCRIPTIONAccess to horizontal neutron beams is created by five beam tubes penetrating the reactorshield structure. All beam tubes are 6 inch diameter tubes originating at or in the reactorreflector. One tangential beam tube is composed of a penetration in the reactor reflectorassembly with extensions through both sides of the reactor shield. A second tangential beamtube penetrates and terminates in the reactor reflector. The two remaining tubes are orientedradial to the reactor core.The beam ports, shown in Figure 10.8, provide tubular penetrations through the concreteshield and reactor tank water, making beams of neutrons (or gamma radiation) available forPage 10-29 CHAPTER 10, EXPERIMENTAL FACILTIES AND UTILIZATION12/2011experiments. The beam ports also provide an irradiation facility for large sample specimens in aregion close to the core. Beam port diameters near the core are 6 inches (15.2 cm). The fivebeam ports are divided into two categories: tangential beam ports and radial beam ports.B. DESIGN AND SPECIFICATIONSTwo tangential beam ports penetrate the graphite reflector, thecoolant water, and theconcrete shield. A hole is drilled in the graphite tangential to the outer edge of the core. Onebeam port terminates at the tangential point to the core. The other beam tubes extend bothdirections from the reflector and out opposite sides of the reactor shield.The two radial beam ports penetrate the concrete shield structure and the coolant water. Oneradial port terminates at the outer edge of the reflector. The second radial port also terminatesat the outer edge of the reflector. However, a hole drilled in the graphite reflector extends theeffective source of the radiations to the reactor core region.C. REACTIVITYPage 10-30 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 10The beam port facilities are external to the reactor core and pose minimal influence on corereactivity. Experiments utilizing beam port facilities require analysis on an individual basis. Inorder to limit reactivity effects, a collimator and/or filter assembly placed within a beam portmay have no portion closer than 2 feet from the outer edge of the core (fuel). Because of thisdistance, reactivity changes due to insertion of suc collimators and filters are negligible. If asample or other material is inserted closer than 2 feet from the outer edge of the core, itsreactivity worth shall be calculated and verified as part of reactor startup. Any single sampleestimated to be worth more than $0.20 shall be secured. Insertion or removal of samplesinside a beam port requires prior approval by a Senior Reactor Operator and notification of theReactor Operator at the time of the action.D. RADIOLOGICAL ASSESSMENTExperiments may be conducted within the beam ports tubes or external to shielding. In thecase of internal beam port experiments, neutron fluxes can reach up to 1012 n cm-2 s-1. Externalneutron beam fluxes range from 106 to 108 n cm-2 s-1 depending on the shielding and filtering inplace. Internal beam tube activations close to the core can reach levels similar to thoseassessed for the in-core facilities above. External neutron beam fluxes may reach hazardouslevels, but they need to be assessed on an individual experiment basis.E. INSTRUMENTATIONThere is no instrumentation associated with the beam port facilities. Instrumentation thatmight be used as part of an experiment program will be evaluated as part of the experimentreview and approval process. Radiation monitors around the reactor bay area are affected bybeam port operation and are indicative of experimental conditions. However, these monitorsare not directly associated with the beam ports.F. PHYSICAL RESTRAINTS, SHIELDS, OR BEAM CATCHERSA step is incorporated into each beam port to prevent radiation streaming through the gapbetween the beam tube and shielding plug. The inner section of each beam port is analuminum pipe 6 inches (15.2 cm) in diameter. The outer section of beam ports 1, 2 and 4consists of a steel pipe 8 inches (20.3 cm) in diameter.Beam ports 3 and 5 have three outer sections with 8 inch, 12 inch, and 15.25 inch diameters.A lead shield ring in the shield structure provides a "shadow" shield for the 15.25 inch beamport section. Special shielding reduces the radiation outside the concrete to a safe level whenthe beam port is not in use. The shielding is provided in four sections as follows:Page 10-31 CHAPTER 10, EXPERIMENTAL FACILTIES AND UTILIZATION 1 2/20111. inner shield plug,2. outer shield plug,3. lead-filled shutter,4. door.The inner shield plug consists of graphite cylinder, backed with a 0.125-inch (0.32-cm) sheet ofboral and 5 inches (12.7 cm) of lead, sandwiched between two 1.25 inch (3.2 cm) thick steelplates. Beam ports 1, 2, and 4 have a section of graphite 6 inch (15.2 cm) in diameter. Beamports 3 and 5 have the same configuration as the other beam ports, except that the graphiteportion is 6 inch (15.2 cm) in diameter, with a change to 8 inch (20.3 cm) in diameter to providegraphite shielding in the 6 inch and 8 inch portions of the tube. Two rollers are provided tofacilitate the insertion and removal of the inner shield plugs. To help guide the shield plug overthe steps in the beam tube during insertion, the inner end of the plug is cone-shaped. Athreaded hole is provided in the outer end of the plug for attaching the beam tube plug-handling tool. The graphite sections are encased in an aluminum canister.The outer shield plug is wooden and is 8 inch (20.3 cm) in diameter and 42 inch (1.07 m) longfor beam ports 1, 2, and 4. Beam ports 3 and 5 have a wooden shield plug for the outer portionof the tube that has a length of 48. inch (1.22 m) and diameter of 15 inch (38.1 cm) for theouter portion of the tube. A handle on the outer end of this plug is provided for manualhandling. The plug is equipped with an electrical circuit consisting of a position switch mountedin the front of the plug and an electrical connector at the rear of the plug. The switch can beactuated only by the inner plug when the inner plug is installed in the beam tube.A physical contact between the inner and outer shield plug, and an electrical connectionbetween the outer plug and the beam tube are part of an installation status circuit. The circuitmonitors the plug configuration or other experiment shield conditions. Information on theconsole for each beam tube indicates the plug or beam tube status.The lead-filled shutter and lead-lined door provide limited gamma shielding when the plugs areremoved. The shutter is contained in a rectangular steel housing recessed in the outer surfaceof the concrete shield. The shutter is -10 inch (25.4 cm) in diameter and 9.5 inch (24.1 cm) thickfor beam ports 1, 2, and 4. Beam ports 3 and 5 have a shutter that is 15.25 inch (38.7 cm) indiameter and 9.5 inch (24.1 cm) thick. The shutter is operated by a removable push rod on theface of the shield structure and can be moved even with the shutter housing door is closed. Inthe open position, a section of the shutter consisting of pipe of equal diameter to the outerportion of the beam tube is aligned with the beam port and the outer shield center plug tofacilitate insertion or removal of the beam plugs. The shutter housing is equipped with a steelcover plate lined with 1.25 inch (3.2 cm) of lead for additional shielding. A removable coverPage 10-32 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 10plate provides easy access to the beam port. The plate can be bolted shut so that the sealwould prevent loss of shielding water if the beam tube should develop a serious leak.While in use, each beam port has controlled access through concrete walls that serve asshielding and via locked gates. The gates have sensors that alert reactor operators to openingwhile the reactor is in operation. Beam stops are in place for each beam when the shutter is inthe open position.G. OPERATING CHARACTERISTICSNeutron beam experiments typically utilize radiation for nuclear analytical techniques. Facilityusage has included positron production through neutron irradiation of copper, neutron depthprofiling, prompt gamma activation analysis, and neutron radiography. Reactor operation forsuch experiments is nominally at full power, but can range to lower powers. For good countingstatistics, beam port experiments normally last on the order of hours and can take up an entireday of operation. Experiments on multiple beam port facilities may be run simultaneously.H. SAFETY ASSESSMENTThe main concern of the beam port facilities is that a puncture within the beam port walls intothe reactor pool area could cause drainage of the pool system. As a result placement of sharpobject, explosive material, or material with high chemical reactivity are limited within thefacility. Inflatable plugs may be placed in the beam ports to seal them and minimize loss ofcoolant.Experiments performed within the beam port facilities shall not change the cooling channelconfiguration of the reactor core and will produce negligible additional heating of the core.Thus, no thermal-hydraulic change will occur within the reactor core due to routine neutronbeam port usage.Heating loads to the beam ports due to collimators, neutron filters, or other materials insertedat a distance no closer than 2 feet from the outer edge of the core will be negligible. If asample or other material is inserted closer than 2 feet from the outer edge of the core, theheating rate shall be calculated and the capacity of the beam port to cool by normal flows of airor water shall be demonstrated to the satisfaction of a supervisory Senior Reactor Operator.Encapsulation of samples shall be sufficient to prevent encapsulation failure due to heating.Mechanical stresses resulting from the weight of collimator and/or filter pieces inserted nocloser than 2 feet from the outer edge of the reactor core will cause no deviations fromnominal design conditions because the beam ports are embedded into the concrete shield atdistances 2 feet and greater from the outer edge of the core. Any experiment inserted in aPage 10-33 CHAPTER 10, EXPERIMENTAL FACILTIES AND UTILIZATION12/2011beam port closer than 2 feet to the outer edge of the core must be designed such that weighton the 2 feet section is less than 100 pounds.10.4 Cold Neutron SourceA. DESCRIPTIONThe Texas Cold Neutron Source Facility is located at beam port 3. It consists of the Texas ColdNeutron Source (TCNS), a curved neutron guide system, a converging neutron guide system, aprompt gamma activation analysis system, and extensive shielding.B. DESIGN AND SPECIFICATIONSThe TCNS consists of a vacuum system, a cryorefrigerator, an aluminum thermosyphon (a.k.a.heat pipe), and a neon cooled moderator chamber. The purpose of the TCNS is to maintain thetemperature of the moderator chamber, filled with mesitylene (1, 3, 5-tri-methylbenzene,C9H112), at a temperature of approximately 45 'K when the reactor is operating at 950 kW andat 36 *K when the reactor is shutdown. The moderator chamber is made of aluminum and iscylindrical in shape (3.75 cm radius and a height of 2 cm). The mesitylene, that has a freezingtemperature of 228.3 OK, serves to moderate incoming thermal neutrons produced in thereactor core and effectively shift their energies to the subthermal region. The neutronsapproach the frozen mesitylene temperature as they travel through the moderator. It isexpected that a large fraction of the neutrons entering the moderating medium will exist at alower energy once they exit the chamber.The mesitylene temperature is maintained through the use of a gravity driven thermosyphonthat uses neon as its working fluid to transfer heat from the moderator to a copper heatexchanger. In turn the copper heat exchanger is coupled to a cold-head that is cryogenicallycooled by a helium cryorefrigerator that maintains a temperature of approximately 17 *K whenthe reactor is operating at 950 kW and 15 OK when the reactor is shutdown.The TCNS is currently equipped with a Cryomech model AL230 helium cryorefrigerator that iscapable of removing 25 W at 20 *K as shown in Figure 10.9. The cryorefrigerator keeps thecoldhead at its target temperature by way of its increased capability and range. Thecryorefrigerator consists of a compressor package and a cold-head. The cold-head (Figure 10.9),is vertically inserted into a Cryomech designed vacuum box shown in Figure 10.10. It is anexpansion device capable of reaching cryogenic temperatures. An extra silicon diode has beeninstalled in order to get more accurate cold-head temperature measurements.Page 10-34 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 10I12/2011Figure 10.9: A1230 Cryomech Cryorefrigerator and Cold HeadThe cold-head consists of two groups of parts; the motor assembly and the base tube assembly.A heat exchanger, made of oxygen free high conductivity (OFHC) copper, is attached to thebottom of the 304 stainless steel tube assembly. The volume of the newly installed OFHCcopper block is significantly larger than that of the former heat exchanger. The increasedvolume of the OFHC copper heat exchanger increased the contact area between itself and thethermosyphon condenser area. The increase in contact area aids in balancing the surface heatflux at the condenser and evaporator ends of the neon thermosyphon. Since the heat transportrate is approximately equal in each section one can transform the surface heat flux at the heatinput side to a lower or higher heat flux at the heat output side because the transformed heatflux varies inversely as the ratio of the surface areas [57]. This heat flux property is importantwhen the heat flux associated with the fixed heat source is either too high or too low to beaccommodated by the cold-head. The copper heat exchanger is in direct contact with the neonthermosyphon that acts to keep the mesitylene chamber at its target temperature. Themoderator, thermosyphon, mesitylene and neon transfer lines are encased within a stainlesssteel vacuum jacket as shown in Figure 10.11The neon contained within the thermosyphon, through use of a two phase transformation,transfers the heat generated by the moderator, due to gamma-ray heating (calculated to beless than 2 W), to the end where the cold-head is located. The two phase transformationperformed by the neon consists of condensation and subsequent vaporization.Page 10-35 CHAPTER 10, EXPERIMENTAL FACILTIES AND UTILIZATIONI12/2011Figure 10.10: Cryomech Cold-Head and Vacuum BoxPage 10-36 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 10C. REACTIVITYThe TCNS is external to the reactor core on Beam Port 3. Studies have shown that this facilityhas a minimal impact on core reactivity.D. RADIOLOGICALAt' the end of the TCNS beam line the thermal equivalent neutron flux was measures at 1 x Wn cm"2 s-1 when the reactor is operating at 950 kW. With the shielding in place, the dose ratesurrounding the facility is c.a. 1 mrem/hr. The neutron beam line can be turned on and offwhen via the remote controlled boral shutter.E. INSTRUMENTATIONThe TCNS is equipped with several sensors that are used to measure the various temperaturesand pressures associated with the TCNS. Five temperature sensors are used in conjunction withthe TCNS to monitor temperature changes and six other sensors are used to monitor pressurechanges. Three type "E" Chromel-Constantan thermocouples (TC1, TC2, and TC3) are attachedto the mesitylene moderator chamber and two silicon diodes (SD1 and SD2) are located in thevicinity of the cold-head. TC1 is located on the flat face of the moderator chamber closest tothe core while TC2 and TC3 are located on the flat face of the moderator furthest from thecore.TC1, TC2, and TC3 are all IOTech Model DBK81 -Built-in Cold Junction Compensationthermocouples. These temperature sensors support up to 7 thermistors per board. Theirmeasuring capabilities support 0.1 degree of precision and 0.5 degree of accuracy from 270 °Kto 650 °K. All three sensors connect to an IOTech Model DAQ2000 16-bit 200ksps ADC (64k 5I~sec conversion) that in turn plugs into the system computer's backplane.Figure 10.12: Silicone Diode and Heater Relative to Cold-HeadPage 10-37 CHAPTER 10, EXPERIMENTAL FACILTIES AND UTILIZATION 12/2011SD1 is located on the copper heat sink and SD2 is located on an aluminum yoke that is wrappedaround the thermosyphon effectively holding it in place mated to the heat exchanger as shownin Figure 10.12. SD1 is the digital temperature indicator and controller for the ScientificInstruments Model 9650 heater. The silicon diode temperature sensor is capable of measuringtemperatures from 1.5 °K to 450 °K with 0.1 "K accuracy of 0.1 degree or better from 1.5 °K to35 °K and 0.5 °K from 35 °K to 450 *K. The heater provides 60 W of heating (30 V @ 2 A) andconnects to the computer through a GPIB interface. SD2 is the temperature indicator andcontroller for the Scientific Instruments Model 9600 heater. The diode's operation range is 1.5°K to 450 °K and has a selected sensor excitation current of 100 pA that is can be switched to 10IVA. The heater provides 25 W of heating (25 V @ 1 A) and connects to the computer through aRS-232C serial port.Figure 10.13: Neon and Mesitylene Handling System with Pressure TransducersThe vacuum levels are monitored by an ion gauge (IG) model IGT 274 Bayard-Albert and threemodel CGT 275 convectron gauges (CG1, CG2, and CG3). Two diaphragm IOTech Model DBK16pressure transducers (PX302- 10OG V and PX302-50G V) are used to measure manometricpressures in psig. PX302-100G V is located on the neon handling system feed line while PX302-50G V is positioned on the mesitylene handling system feed line (Figure 10.13). Each transducerconnects to the DAQ2000. Up to 16 DBK16s can be connected to a single DAQ2000 channel. Itshould be noted that the pressure transducer located on the neon handling system can onlyrecord pressures of 100 psig (689 kpa) or less and the transducer on the mesitylene handlingsystem can only record pressures of 50 psig (345 kpa) or less.The IG and CG1 are located on the right face of the vacuum box. Both the IG and CG1 are usedto monitor the evacuated volume in the vacuum box. CG2 is located to the left of the vacuumbox between the Leybold manufactured Turbotronik/NT 50 turbo-molecular pump and theroughing pump that are used to obtain the required vacuum level (Figure 3.13). CG3 is placedPage 10-38 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 10with the vacuum pump used to evacuate the curved neutron guide. The convectron gauges arecapable of reading 10-4 torr to 990 torr. All of the vacuum sensors are connected to anextended capability vacuum gauge controller (307-VGC) that has an operating range of 5x10-12torr to 760 torr. The 307-VGC connects to the system computer through an RS-232C serial port.The TCNS vacuum system is also equipped with two remote control gate valves (GV1 and GV2)model DN 63 and DN 16 that are manufactured by the Swiss company VAT. The gate valves areused for isolating the vacuum system during TCNS startup and shutdown procedures. GV1 islocated between the vacuum box and the turbo-molecular pump and GV2 is located betweenthe turbomolecular pump and the mechanical pump as shown in Figure 3.13. Both valves arepneumatically actuated and have position indicator switches at each extent. The gate valves aremonitored and controlled by a Keithley PDISO-8 that contains 8 optically isolated inputs and 8electromechanical relay outputs with 3A ratings. The PDISO-8 plugs into the system computerbackplane.F. PHYSICAL RESTRAINTS, SHIELDS, OR BEAM CATCHERSThe TCNS system has an array of restraints, shields, and beam catchers. Figure 10.14 showsthese shielding structure surrounding the TCNS with materials including boral, polyethylene,borated polyethylene, Boroflex, Lithoflex, concrete, lead, and Li2CO3 powder. The boratedmaterials, Li based materials, and polyethylene are intended for neutron shielding. The lead isprimarily a gamma-ray shield. The concrete is in place for both neutron and gamma-rayshielding.G. OPERATING CHARACTERISTICSIf the TCNS has not been operated recently, the evacuated volume around the moderatorchamber and neutron channels should have a nitrogen atmosphere of less than 650 torr. Themoderator chamber and filling lines should be filled with low pressure (~1-2 psig or -7-14 kpa)helium. Mesitylene should be stored in its reservoir with all valves on the mesitylene handlingsystem shut. The thermosyphon valve should be in the off position from the neon-reservoirwhich should have a pressure neon atmosphere of about 145 psig (1 MPa). At this time, thevacuum system should be shut off and the instrumentation system may or may not be turnedoff.If the TCNS has been operated recently, the evacuated volume around the moderator chamberand neutron channels should be evacuated to less the 10-4torr.Page 10-39 CHAPTER 10, EXPERIMENTAL FACILTIES AND UTILIZATION 12/2011H. SAFETY ANALYSISIf during startup the heat transport rate is too high, the copper heat exchanger temperaturemay not significantly rise above that of the condenser. Therefore, if the neon in thethermosyphon is originally frozen more condensate will continue to freeze as melting andvaporization occurs in the evaporator end. Since the liquid in the evaporator will not bereplenished as long as the condensate in the condenser remains frozen, the evaporator andmesitylene chamber will begin to overheat which will cause an unwanted buildup in pressuretowards the bottom of the thermosyphon. In order to avoid this situation, care should be takento optimize the thermal resistance between the heat exchanger and the thermosyphon duringstartup. Freeze-out can be avoided by fully insulating the condenser against heat loss andallowing the thermosyphon condenser temperature to rise above neon's critical point of 24.5'K. This will allow the liquid neon to replenish the vaporized neon in the evaporator section andkeep the mesitylene from melting too fast.The vapor within the thermosyphon typically reaches sonic velocity during startup and thus thedrag force at the liquid-vapor interface may be relatively high. If the entrainment limit is notgreater than the sonic limit the neon liquid will be entrained by the neon vapor and willtherefore lead to evaporator dry out and overheating since the liquid return rate to theevaporator will be reduced. This type of failure will not cause any type of pressure buildupwithin the thermosyphon but will affect the ability of the TCNS to keep the moderator frozen.However, as long as the actual heat transport rate is equal to the sonic limit and theentrainment limit is greater than the sonic limit, entrainment can be avoided. Entrainment mayalso be avoided by adding a non-condensable gas to the vapor space. The non-condensable gas,during startup, will limit the effective condenser heat rejection area by occupying most of thevapor condenser area while the neon vapor is at a low pressure. By occupying the vapor space,the noncondensable gas also raises the thermal resistance between the condenser and heatexchanger and thus decreases the ability of freeze-out to occur.Page 10-40 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 10I12/2011None of the failure mechanisms presented here increases the probability of an accident,involving the use of the TCNS, to occur. Each of the above mentioned failures fall within thelimits and capabilities previously evaluated.10.5 Non-reactor experiment facilitiesThe NETL maintains a number of facilities related to nuclear radiation and detection. Thesefacilities are utilized for teaching, research, and service work.10.5.1 Neutron generator roomThe NETL houses a neutron generator room that has 3 foot thick concrete walls, floor andceiling. The room currently is utilized for operation of a Thermo Scientific MP 320 D-T neutrongenerator and other neutron based experiments. Figure 10.15 shows that this is a compactneutron generator designed for portability. The MP 320 has a flux of 1 x 108 n s-1 and has apulse rate of between 250 Hz to 20 kHz. The fast neutron source uses a deuterium-tritiumreaction to produce 14 MeV neutrons.The system is paired with an ORTEC GMX50P4-83 n-type HPGe detector. The detector isspecially equipped with an integrated heater for annealing the HPGe crystal after damage fromfast neutrons. The MP 320 provides an output to synchronize gamma-ray spectrum acquisitionwith the neutron pulses. For this setup, two MCAs are utilized so that spectra will be acquiredduring the neutron pulse (prompt) and between the pulses (delayed).Figure 10.15: Thermo MP 320 Neutron Generator at NETLPage 10-41 CHAPTER 10, EXPERIMENTAL FACILTIES AND UTILIZATIONI12/201110.5.2 Subcritical assemblyCylindrical subcritical assemblies of graphite and polyethylene are utilized for studentlaboratory experiments with neutron sources and a subcritical 235U assembly. The plutonium-beryllium neutron sources and uranium dioxide used in the polyethylene subcritical assemblymay be stored and used in the room containing the reactor, but are licensed separately fromthe reactor. The subcritical core and moderator assemblies are products of Lockheed NuclearProducts. Figure 10.16 Illustrates the subcritical facilities.C. ASWAA. GW.IAV11 FIFLECI0 ONFigure 10.16 Subcritical AssembliesThe subcritical polyethylene core is a cylinder 10 inches in diameter and 14 inches long.Reflector assemblies can be assembled with or without the fueled core. Dimensions of thecylindrical reflector assemblies are 30 inch diameter by 34 inch length for the graphitemoderator and 22 inch diameter by 25 inch length for the polyethylene moderator. Anadditional graphite moderator cylinder 30.5 inches high by 24 inch diameter is available forneutron source moderation.10.5.3 Laboratories10.5.3.1 Radiochemistry laboratoryThe radiochemistry laboratory focuses on work utilizing open nuclear sources. It contains afume hood along with laboratory equipment to support radiochemistry experiments. Wetchemistry experiments and radioactive gas experiments are often conducted in this facility.Nuclear detection equipment including alpha spectroscopy, beta-gamma coincidencespectroscopy, and standard Nal(TI) detectors are currently utilized in the laboratory.Page 10-42 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 1010.5.3.2 Neuron Activation Analysis LaboratoryA neutron activation analysis laboratory contains a terminal for the pneumatic transit system.The laboratory includes a glove box utilized for sample handling and houses the terminal for themanual pneumatic transit system. The laboratory contains shielded areas for neutronactivation analysis samples and HPGe detectors for gamma-ray spectral acquisitions.10.5.3.3 Radiation detection laboratoryThe radiation detection laboratory is utilized for gamma-ray spectroscopy as well as laboratoryclasses. It is one of the larger laboratories with benches that may be utilized for a wide varietyof radiation detection experiments. Multiple HPGe detectors are in the facility that are utilizedfor measurement of long-lived radionuclides. This laboratory is primarily utilized forexperiments with sealed nuclear sources.10.5.3.4 Sample preparation laboratoryThe sample preparation laboratory is utilized for sample packaging and recording. It has a fumehood for experiments. It has a clean bench, high precision scale, and ovens for sample drying.Radioactive materials are not utilized in this laboratory to prevent contamination of samplesbeing prepared for experiments.10.5.3.5 General purpose laboratoryThe general purpose laboratory is utilized for radioactive sample based experiments along withnon-radioactive material experiments. The laboratory includes work benches and storagecabinets.10.6 Experiment ReviewThe Reactor Oversight Committee (ROC) oversees the nuclear reactor and approval ofexperiments. The ROC ensures that the experiment follows ALARA protocols and does notviolate any Technical Specifications. In addition a general safety analysis is performed.Experimenters are required to submit a document describing their experiment and address theitems identified in Table 10.8.The ROC reviews the safety analysis report with respect to facility Technical Specifications,public safety, experimenter safety, protection of the facility, and ALARA principles.Experimental proposals may be accepted, rejected, or have suggested modifications. The ROCmay also require additional analysis to support the safety assessment of the experiment. OncePage 10-43 CHAPTER 10, EXPERIMENTAL FACILTIES AND UTILIZATION12/20211ian experiment is approved, experimenters may schedule experiments through an OperationsRequest. An Operations Request requires the approval of a Senior Reactor Operator prior tobeing conducted.Table 10.8: Items to be Addressed in Safety Analysis for ExperimentsTopic DescriptionDescription and Purpose of Experiment This section shall include a general review of theexperiment. A purpose and goals should beidentified.Experimental RequirementsExperiment Facility and LocationMaximum Reactor PowerMaximum Operation TimePhysical Experiment EffectsReactivityThermal Hydraulic and ExperimentTemperatureMechanical StressThis section identifies the facilities andoperational requirements for the facility.Identify the specific facility and location withinthe reactor.Describe the maximum power at which theexperiment will be conducted (for pulseexperiments the reactivity insertion should beidentified as well).Provide a conservative estimate of the time atpower required for the experiment.This section describes the reactor effects.Conservatively based reactivity calculationsshould be performed. Identify worst casescenarios for the experiment and calculate thereactivity effect of these cases.Identify heat transfer concerns that will occur inexperiment. If there appears to be any heattransfer concerns, conservative calculationsshould be made to calculate maximumtemperatures in the fuel and in the experiments.Mechanical stress issues should be identified.Calculations should support conclusions based onpossible pressure increases or other mechanicalstresses.The materials in the experiment should beidentified and classified.Activation calculations should be performed.Based on these calculations, health physicsMaterial EvaluationRadioactivityPage 10-44 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 10I12/2011Table 10.8: Items to be Addressed in Safety Analysis for ExperimentsTopic Descriptionconcerns should be addressed. If radioiodine orradiostrontium are produced, calculations shouldbe compared to maximum values stated in theTechnical Specifications.Material HazardsTrace Element Impurities Which MayRepresent a Significant RadiologicalHazardHigh Cross-Section ElementsFlammable, Volatile, or Liquid MaterialsExplosive ChemicalsRadiation Sensitive Materials WhichWhen Exposed to Radiation ExhibitDegradation of Mechanical Properties,Decomposition, Chemical Changes, orGas EvolutionToxic CompoundsCryogenic LiquidsUnknown MaterialsExperiment ClassificationThis relates to specific material hazards.Identify elements which may activate to produceradiation hazards.Identify high cross-section elements and addressreactivity and radioactivity concerns.Identify flammable, volatile, or liquid materials.If such materials are in the experiment, addresscontainment issues and estimate consequences ofworst case accident scenario.Identify explosive chemicals within theexperiments. Address safety concerns and makesure quantities are less than those stated in theTechnical Specifications.Identify materials that suffer from radiationeffects. Special concern should be placed onmaterials that emit hydrogen or othercombustible gasses upon being irradiated. Alsoaddress possible degradation of samplecontainment during irradiation.Identify toxic compounds and chemicals withinthe experiment. Address safety concerns.Identify cryogenic liquids within each experimentand address safety concerns.Sometimes samples are analyzed via variousnuclear techniques. In such cases the makeup ofsamples may not be entirely known. Try toestimate the bounds of experimental samplecompositions and address safety concerns.Experiments are identified as being Class A, B, orC.1) Class A experiments require a senior operator(Class A, SRO) to direct an activity ofPage 10-45 CHAPTER 10, EXPERIMENTAL FACILTIES AND UTILIZATION 12/2011Table 10.8: Items to be Addressed in Safety Analysis for ExperimentsTopic Descriptionexperiment.2) Class B experiments require only an operatorand if necessary an experimenter (Class B,RO) to perform the experiment, with an SROavailable.3) Class C experiments are all non-reactorexperiments.Page 10-46 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 1111.0 RADIATION PROTECTION AND WASTE MANAGEMENTThis chapter deals with the overall NETL radiation protection program and the correspondingprogram for management of radioactive waste. The chapter is focused on identifying theradiation sources which will be present during normal operation of the reactor and upon themany different types of facility radiation protection programs carried out to monitor andcontrol these sources. This chapter also identifies expected radiation exposures due to normaloperation and use of the reactor.11.1 Radiation ProtectionThe purpose of the NETL radiation protection program is to allow the maximum beneficial useof radiation sources with minimum radiation exposure to personnel and the general public.Requirements and procedures set forth in this program are designed to meet the fundamentalprinciple of maintaining radiation exposures As Low As Reasonably Achievable (ALARA).11.1.1 Radiation SourcesThe radiation sources present at the NETL can be categorized as airborne, liquid, or solid.Airborne sources consist mainly of argon-41 due largely to neutron activation of air dissolved inthe reactor's primary coolant. Liquid sources include mainly the reactor primary coolant. Solidsources are more diverse, but are typical of a research reactor facility. Such sources include thefuel in use in the core, irradiated fuel in storage, and fresh unirradiated fuel. In addition, othersolid sources are present such as the neutron startup source, irradiated experiment materials,items irradiated as part of normal reactor use, various check, reference, and calibration sourcesand a limited amount of solid waste.11.1.1.1 Airborne Radiation SourcesDuring normal operation of the NETL reactor, airborne radioactivity is almost exclusively Ar-41..11.1.1.1.1 Production of Ar-41 in the Reactor RoomProduction of Ar-41 in the pool water can be found by determining the concentration of Ar-40in the water and multiplying by the volume of water irradiated, the Ar-41 production crosssection, and the thermal neutron flux. From information obtained from Dorsey', one sees thatthe Ar-40 concentration in water at typical core inlet temperature is approximately 7.1xlO15atoms cm-3. Given the volume of water in the core is 18500 cm3, the effective cross section forproduction of Ar-41 is 0.661x10-24 cm2, and thermal neutron flux of 2.4x1013 n cm-2 s-1 at thecentral thimble at 1.1 MW is assumed to be the uniform flux across the entire core, aconservative Ar-41 production rate is approximately 2.1x109 atom s-1. Assuming continuousoperation at 1.1 MW, the equilibrium activity of Ar-41 in the pool water is 2.1x109Bq.Page 11-1 CHAPTER 11, RADIATION PROTECTION AND WASTE MANAGEMENT PROGRAM 12/2011Likewise, the production of Ar-41 in experimental facilities can be found by multiplying theconcentration of Ar-40 in air by the volume of air irradiated, the Ar-41 production cross section,and the thermal neutron flux. The natural concentration of argon in air is 0.93% which equates(at STP) to 2.5x1017 argon-40 atoms cm-3. The effective air volume of the beam tubes is 5.9x105cm-3 and the average thermal neutron flux in the beam tubes is 1x1011 n cm-2 s-1. This results inan argon-41 production rate in the beam tubes of 9.7x109 atom s-1. The effective air volume ofthe rotary specimen rack (RSR) is 3.3x104 cm-3 and the average thermal neutron flux in the RSRis 6x1012 n cm-2 s-1. This results in an argon-41 production rate in the RSR of 3.3x101° atom s1.Assuming continuous operation at 1.1 MW, the equilibrium activity of Ar-41 in theexperimental facilities is 4.3x1010 Bq.At equilibrium, the production of Ar-41 in the poo waterl and experimental facilities is equal tothe removal of Ar-41 from the pool water and experimental facilities. Assuming this removal isexclusively diffusion of Ar-41 into the air of the reactor room and assuming all this activitydiffuses uniformly into the volume of the reactor room (4.12x109 cm3), the Ar-41 activityconcentration would be 3.0x10-4 VCi cm-3 which is 100 times the DAC value of 3x10-6 VCi cm-3.As Ar-41 is a noble gas, assuming a semi-infinite cloud model, the dose rate in the reactor roomwould be approximately 320 mrem hr-1 during extended 1.1 MW operations due to airborne Ar-41. While this would be a high radiation area, exposures to this airborne radiation source caneasily be controlled by personnel monitoring and procedural control over access to the reactorroom. However, in reality, all the experimental facilities are not utilized simultaneously(resulting in less volume of air for Ar-41 production) and a facility ventilation system exchangesthe room air mitigating this potential exposure. Additionally, due to the utilization trends at theNETL, extended 1.1 MW operations are not the norm. Operational experience has shown thatairborne argon-41 is not a significant contribution to occupational dose at the NETL.11.1.1.1.2 Radiological Impact of Ar-41 Outside the Operations BoundaryArgon-41 is the only routine effluent from the NETL. A conservative estimate of effluentconcentration outside the facility is to calculate the ground level concentration at the buildingusing:X(0,0,0)= Q/(0.5)(A)(0)whereX(0,0,0) = Ground level concentration at the building in jiCi m-3Activity release rate in IVCi s-1A = Cross sectional area of the reactor building (256 M2)0 = Mean wind speed (assumed as 1 m s-1)Page 11-2 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 11Q is determined by multiplying the activity concentration in the reactor room (3.0x104 I.Ci cm-3)by the volume release rate of the stack (3.9x106 cm3 s-1). Thus Q = 1170 IpCi s1 and X(0,0,0) =9.1 ICi m3 = 9.1x10-6 lCi cm-3.While this concentration is about 900 times the effluentconcentration limit of Ix10i8 I[Ci cm3, this is based on a very conservative calculation based oncontinuous operation at 1.1MW. In reality, operations are not continuous and are not alwaysat full power. Measured Ar-41 releases over the past several years shows an average annualAr-41 release of less than 6 Ci per year (0.2 VICi s1). Using a 6 Ci per year release rate in theabove equation gives a ground level concentration at the building of 1.6x103 I.Ci m-3 = 1.6x10-9IVCi cm-3 which is well below the effluent concentration limit.Determination of radiation dose to the general public from airborne effluents may also becarried out using several computer codes recognized by regulatory authorities. One suchmethod involves use of the Clean Air Assessment Package -1988 (CAP88-PC). Application of thiscode to the very conservatively projected Ar-41 releases from continuous 1.1MW operation atthe NETL predicts a dose to the maximally exposed individual of approximately 66 mrem peryear. Applying the code to the more reasonable release rate of 6 Ci per year predicts a dose tothe maximally exposed individual of less than 0.02 mrem per year.11.1.1.2 Liquid Radioactive SourcesLiquid radioactive material routinely produced as part of the normal operation of the NETLincludes miscellaneous neutron activation products in the primary coolant. Many of theseactivation products are deposited in the mechanical filter and the demineralizer resins.Therefore, these materials are dealt with as solid sources. Non-routine liquid radioactive wastecould result from decontamination or maintenance activities (i.e., filter or resin changes). Theamount of this type of liquid waste is expected to remain small, especially based on pastexperience. There are also various liquid radioactive materials used as reference or calibrationstandards for instruments. However, these materials tend to be low volume and low activity. Aliquid analytical samples produces liquid radioactive sources. However, these materials too aretypically low volume and low activity. Thus, the primary liquid radioactive source at the NETL isthe primary coolant.11.1.1.2.1 Radioactivity in the Primary CoolantNitrogen-16 is produced by fast neutron activation of oxygen-16 in the water of the primarycoolant. The oxygen density in water is approximately 3.3x1022 atoms cm-3. Given the volumeof water in the core is 18500 cm3, the effective cross section for production of N-16 is 2.1X10-29cm 2, and neutron flux of lxl013 n cm-2 s-1 in the energy range of interest at 1.1 MW is assumedto be the uniform flux across the entire core, a conservative N-16 production rate isapproximately 1.3x1011 atom s-1. Assuming continuous operation at 1.1 MW, the equilibriumPage 11-3 CHAPTER 11, RADIATION PROTECTION AND WASTE MANAGEMENT PROGRAM 12/2011activity of nitrogen-16 in the core region is 1.3x1011 Bq. At equilibrium, the production of N-16in the core region is equal to the removal of N-16 from the pool. As the N-16 tends to stay insolution and the half-life of N-16 is 7.1s, the primary removal mechanism from the pool isdecay.The N-16 from the core region moves through the reactor tank by natural convection.Assuming the water containing the N-16 continues upward to the surface of the pool at thecoolant flow velocity through the core (17 cm s-1), itwill traverse the distance to the surface(640 cm) in about 38 seconds. In that time period, substantial radioactive decay will haveoccurred resulting in 3.2x109 Bq actually reaching the surface. Assuming the N-16 that makes itto the surface of the pool spreads out into a uniform disk of 2 meter diameter, the calculateddose rate at I meter above the surface of the water would be about 90 mrem hr-1. Exposuresto this liquid radiation source can easily be controlled by personnel monitoring and proceduralcontrol over access to the area of the surface of the reactor pool. However, in reality, due tothe utilization trends at the NETL, extended 1.1 MW operations are not the norm. Operationalexperience has shown that nitrogen-16 is not a significant contribution to occupational dose atthe NETL.11.1.1.2.2 N-16 Radiation Dose Rates from Primary CoolantNitrogen-16 is produced by fast neutron activation of oxygen-16 in the water of the primarycoolant. The oxygen density in water is approximately 3.3x1022 atoms cm-3. Given the volumeof water in the core is 18500 cm3, the effective cross section for production of N-16 is 2.1X10-29cm 2, and neutron flux of lx103 n cm2 s1 in the energy range of interest at 1.1 MW is assumedto be the uniform flux across the entire core, a conservative N-16 production rate isapproximately 1.3x1011 atom s-1.Assuming continuous operation at 1.1 MW, the equilibriumactivity of nitrogen-16 in the core region is 1.3x1011 Bq. At equilibrium, the production of N-16in the core region is equal to the removal of N-16 from the pool. As the N-16 tends to stay insolution and the half-life of N-16 is 7.1s, the primary removal mechanism from the pool isdecay. The N-16 from the core region moves through the reactor tank by natural convection.The time it takes for the N-16 to move to the surface of the tank, T, is given by the ratio of thevolume above the core region (4x107 cm3) to the rate at which the activated coolant is flowinginto that volume (8x103 cm3 s-1). Thus, T is equal to 5000 s. By the time the N-16 would reachthe surface of the tank, it has decayed to background. Therefore, an equilibrium concentrationof N-16 in the primary coolant will never be reached. Thus, the N-16 becomes a radiationsource below the surface of the reactor tank. As it takes 5000s for the coolant exiting the coreto reach the surface 6.4m above, the vertical velocity of the coolant is approximately 1.3 cm s-.After ten half-lives (71s), the activity would be reduced by approximately three orders ofmagnitude. In 71 seconds, the N-16 would move upward approximately 92 cm. Additional timespent moving upward results in additional decay. Thus it is assumed any significantcontribution to dose at the surface of the tank results from N-16 activity approximately 5.5mbelow the surface of the tank. As a conservative case, the dose rate from a disk source of 2meter diameter with total activity equal to the equilibrium N-16 activity located 5.5m below thesurface of the tank is calculated to be approximately 170 mrem hr-' at the surface of the tankPage 11-4 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 11I12/2011without taking into account the shielding provided by the 5.5m of water. The tenth valuethickness of water for N-16 photons is approximately Im. Thus, even taking into account abuildup factor of approximately an order of magnitude for this thickness of water, the dose ratewould be attenuated by approximately four orders of magnitude due to the shielding providedby the water resulting in actual dose rates from N-16 near background at the surface of thetank.11.1.1.3 Solid Radioactive SourcesThe solid radioactive sources associated with the NETL program are summarized in thefollowing table. Because the actual inventory of reactor fuel and other radioactive sourcescontinuously changes as part of the normal operation, the information in the table is to beconsidered representative rather than an exact inventory.Page 11-5 CHAPTER 11, RADIATION PROTECTION AND WASTE MANAGEMENT PROGRAMI12/2011r~rrnr~r ~na~nAlthough solid waste is included in the preceding table, more information on wasteclassification, storage, packaging and shipment is included in Section 11.2.11.1.1.3.1 Shielding LogicAlthough not a solid source of radioactivity itself, shielding is involved in reducing radiationlevels from many solid sources and therefore the basic logic used for the reactor shielding isincluded here. The logic and bases used for the NETL shielding design originated from GeneralAtomic developed source terms for 1.5MW operation. Shielding was designed for a surfacedose rate of no more than 1 mrem hr-1.Operational experience has shown the shield performs as designed. As the irradiated fuel is themost significant solid radioactive source at the NETL, as long as it remains within the reactorshield structure, no significant occupational radiation exposure is expected.11.1.2 Radiation Protection ProgramThe radiation protection program for the NETL is executed with the goal of limiting radiationexposures and radioactivity releases to levels that are as low as reasonably achievable withoutseriously restricting operation of the facility for purposes of education, research, and service.The program is executed in coordination with The University of Texas at Austin, Office ofEnvironmental Health and Safety, Radiation Safety Office. The program has been reviewed andapproved by the Reactor Oversight Committee for the facility. The program was developedfollowing the guidance of ANSI 15.11 Radiation Protection at Research Reactor Facilities anddesigned to meet the requirements of 10CFR20. Some aspects of the program deal withradioactive materials regulated by the Texas Department of State Health Services (TDSHS)under license L00485 and the program has been reviewed by the Radiation Safety CommitteePage 11-6 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 11which has responsibility for administering the radiation protection program under the TDSHSlicense.11.1.2.1 Management and Administration11.1.2.1.1 Level 1 PersonnelLevel 1 represents the central administrative functions of the university and the Cockrell Schoolof Engineering. The University of Texas at Austin is composed of 16 separate colleges andschools; the Cockrell School of Engineering manages eight departments with individual degreeprograms. The Nuclear Engineering Teaching Laboratory (NETL) is one of several education andresearch functions within the School.President, The University of Texas at AustinThe President is the individual vested by the University of Texas System with responsibility forthe University of Texas at Austin.Executive Vice President and ProvostResearch and educational programs are administered through the Office of the Executive VicePresident and Provost. Separate officers assist with the administration of research activities andacademic affairs with specific management functions delegated to the Dean of the CockrellSchool of Engineering and the Chairman of the Mechanical Engineering Department.Dean of the Cockrell School of EngineeringThe Dean of the Cockrell School of Engineering reports to the Provost. The School consists of 8departments and undergraduate degree programs and 12 graduate degree programs.11.1.2.1.2 Level 2 PersonnelThe Nuclear Engineering Teaching Laboratory operates as a unit of the Department ofMechanical Engineering at The University of Texas at Austin. Level 2 personnel are those withdirect responsibilities for administration and management of resources for the facility, includingthe Chair of the Mechanical Engineering Department, the NETL Director and Associate Director.Oversight roles are provided at Level 2 by the Radiation Safety Committee, the Radiation SafetyOfficer and the Reactor Oversight Committee.Chair, Department of Mechanical EngineeringThe Chairman reports to the Dean of the Cockrell School of Engineering. The Departmentmanages 8 areas of study, including Nuclear and Radiation Engineering.Page 11-7 CHAPTER 11, RADIATION PROTECTION AND WASTE MANAGEMENT PROGRAM12/2011Director, Nuclear Engineering Teaching Laboratory (NETL Director)Nuclear Engineering Teaching Laboratory programs are directed by an engineering facultymember with academic responsibilities in nuclear engineering and research related to nuclearapplications. The Director is a member of the Cockrell School of Engineering, and theDepartment of Mechanical Engineering.Associate DirectorThe Associate Director is responsible for safe and effective conduct of operations andmaintenance of the TRIGA nuclear reactor. Other activities performed by the Associate Directorand staff include neutron and gamma irradiation service, operator/engineering trainingcourses, and teaching reactor short courses. In addition to Level 3 staff, an AdministrativeAssistant and an Electronics Technician report to the Associate Director. Many staff functionsoverlap, with significant cooperation required.Safety OversightSafety oversight is provided for radiation protection and facility safety functions. A Universityof Texas Radiation Safety Committee is responsible programmatically for coordination, trainingand oversight of the University radiation protection program, with management of the programthrough a Radiation Safety Officer. Nuclear reactor facility safety oversight is the responsibilityof a Reactor Oversight Committee.Radiation Safety CommitteeThe Radiation Safety Committee reports to the President and has the broad responsibility forpolicies and practices regarding the license, purchase, shipment, use, monitoring, disposal andtransfer of radioisotopes or sources of ionizing radiation at The University of Texas at Austin.The Committee meets at least three times each calendar year. The Committee is consulted bythe Office of Environmental Health and Safety concerning any unusual or exceptional actionthat affects the administration of the Radiation Safety Program.Radiation Safety OfficerA Radiation Safety Officer holds delegated authority of the Radiation Safety Committee in thedaily implementation of policies and practices regarding the safe use of radioisotopes andsources of radiation as determined by the Radiation Safety Committee. Radiation Safety Officerresponsibilities are outlined in Radioactive Materials License Commitments for The University ofTexas at Austin. The Radiation Safety Officer has an ancillary function reporting to the NETLDirector as required on matters of radiological protection. The Radiation Safety Program isadministered through the University Office of Environmental Health and Safety. A NETL HealthPhysicist (Level 3) manages daily radiological protection functions at the NETL, and reports toPage 11-8 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 11the Radiation Safety Officer as well as the Associate Director. This arrangement assuresindependence of the Health Physicist through the Radiation Safety Officer while maintainingclose interaction with NETL line management.Reactor Oversight Committee (ROC)The Reactor Oversight Committee evaluates, reviews, and approves facility standards for safeoperation of the nuclear reactor and associated facilities. The ROC meets at least semiannually.The ROC provides reports to the Dean on matters as necessary throughout the year andsubmits a final report of activities no later than the end of the spring semester. The ROC makesrecommendations to the NETL Director for enhancing the safety of nuclear reactor operations.Specific requirements in the Technical Specifications are incorporated in the committee charter,including an audit of present and planned operations. The ROC is chaired by a professor in theCockrell School of Engineering. ROC membership varies, consisting of ex-officio and appointedpositions. The Dean appoints at least three members to the Committee that represent a broadspectrum of expertise appropriate to reactor technology, including personnel external to theSchool.11.1.2.1.3 Level 3 PersonnelLevel 3 personnel are responsible for managing daily activities at the NETL. The ReactorSupervisor and Health Physicist are Level 3.Reactor SupervisorThe Reactor Supervisor function is incorporated in a Reactor Manager position, responsible fordaily operations, maintenance, scheduling, and training. The Reactor Manager is responsible forthe maintenance and daily operations of the reactor, including coordination and performanceof activities to meet the Technical Specifications of the reactor license. The Reactor Managerplans and coordinates emergency exercises with first responders and other local support(Austin Fire Department, Austin/Travis County EMS, area hospitals, etc.). The Reactor Manager,assisted by Level 4 personnel and other NETL staff, implements modifications to reactorsystems and furnishes design assistance for new experiment systems. The Reactor Managerassists with initial experiment design, fabrication, and setup. The Reactor Manager providesmaintenance, repair support, and inventory control of computer, electronic, and mechanicalequipment. The Administrative Assistant and Reactor Manager schedule and coordinate facilitytours, and support coordination of building maintenance.Health PhysicistThe Health Physicist function is incorporated into a Laboratory Manager position, responsiblefor radiological protection (Health Physics), safe and effective utilization of the facility (LabManagement), and research support. Each of these three functions is described below. ThePage 11-9 CHAPTER 11, RADIATION PROTECTION AND WASTE MANAGEMENT PROGRAM 1 12/2011Laboratory Manager is functionally responsible to the NETL Associate Director, but maintains astrong reporting relationship to the University Radiation Safety Officer and is a member of theRadiation Safety Committee. This arrangement allows the Health Physicist to operateindependently of NETL operational constraints in consideration of radiation safety.-Health Physics: NETL is a radiological facility operating in the State of Texas under afacility operating license issued by the Nuclear Regulatory Commission (NRC).Radioactive material and activities associated with operation of the reactor areregulated by the NRC, and the uses of radioactive materials at the NETL not associatedwith the reactor are regulated by the Texas Department of State Health Services(TDSHS). The NETL Health Physicist ensures operations comply with these requirements,and that personnel exposures are maintained ALARA. One or more part-timeUndergraduate Research Assistants (URA) may assist as Health Physics Technicians.-Lab Management: The lab management function is responsible for implementation ofoccupational safety and health programs at the NETL. The Laboratory Manager supportsUniversity educational activities through assistance to student experimenters in theirprojects by demonstration of the proper radiation work techniques and controls. TheLaboratory Manager participates in emergency planning for NETL and the City of Austinto provide basic response requirements and conducts off-site radiation safety training toemergency response personnel such as the Hazardous Materials Division of the FireDepartment, and Emergency Medical Services crews.-Research Support: The mission of The University of Texas at Austin is to achieveexcellence in the interrelated areas of undergraduate education, graduate education,research and public service. The Laboratory Manager and research staff supports theresearch and educational missions of the university at large, as well as development orsupport of other initiatives. The Laboratory Manager is responsible for coordinating allphases of a project, including proposal and design, fabrication and testing, operation,evaluation, and removal/dismantlement. Researchers are generally focused onaccomplishing very specific goals, and the research support function ensures the NETLfacilities are utilized in a safe efficient manner to produce quality data. The LaboratoryManager obtains new, funded research programs to promote the capabilities of theneutron beam projects division for academic, government and industrial organizationsand/or groups.11.1.2.1.4 Level 4 PersonnelReactor Operators and Senior Reactor Operators (RO/SRO) operate and maintain the reactorand associated facilities. An RO/SRO may operate standard reactor experiment facilities asdirected by the Reactor Supervisor.11.1.2.1.5 Other Facility StaffPage 11-10 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 11In addition to the line management positions defined above, NETL staff includes anAdministrative Assistant, an Electronics Technician, and variously one or more UndergraduateResearch Assistants assigned either non-licensed maintenance support (generally but notnecessarily in training for Reactor Operator licensure) or to support the Laboratory Manager asHealth Physics Technicians and/or research support.11.1.2.2 Health Physics Procedures and Document ControlOperation of the radiation protection program is carried out under the direction of the HealthPhysicist using formal NETL health physics procedures. These procedures are reviewed foradequacy by the Health Physicist and others as appropriate, and are approved by the FacilityDirector for submission to the Reactor Oversight Committee for review and approval. Theoriginal copy of the procedures is maintained by and the distribution of the procedures ismanaged by the Reactor Supervisor. A current copy is maintained in the reactor control room.The procedures are reviewed periodically and changes are made as necessary. While notintended to be all inclusive, the following list provides an indication of typical radiationprotection procedures used in the NETL program:-Radiation Monitoring -Personnel-Radiation Monitoring -Facility-NETL ALARA Program-Radiation Protection Training-Radiation Monitoring Equipment-Radioactive Material Control-Radiation Work Permits11.1.2.3 Radiation Protection TrainingIndividuals who do not have formal training in radiation safety must attend the University'sradiation worker training course. The course is approximately eight hours in length.Alternatively, the course may be conducted via computer or over the Internet, or by using videoinstruction. If these methods of training are used the course will include the same topics asthose included in a live course. The Radiation Safety Officer may waive the course if thePage 11-11 CHAPTER 11, RADIATION PROTECTION AND WASTE MANAGEMENT PROGRAM12/2011individual can provide evidence of equivalent training and/or experience. If the Radiation SafetyOfficer waives the course, the individual must take the radiation worker refresher course.The radiation worker refresher course is approximately one hour in length and addresses topicsspecific to the University such as dosimetry, waste disposal, purchasing, emergency procedures,operating procedures, record keeping, as well as a basic review of radiation safety techniques.Alternatively this course may be conducted via computer or over the Internet, or by using videoinstruction. If these methods of training are used the course will include the same topics asthose included in a live course.Upon successful completion of either course, credit is posted to the individual's electronictraining history in the campus-wide training database. If requested, the successful graduate isissued a certificate of completion.Radiation safety courses are taught by senior staff of the Radiation Safety Office. At the NuclearEngineering Teaching Laboratory (NETL), comparable, site-specific radiation worker training istaught by the NETL health physicist. If necessary or desired, outside training specialists may beutilized to present the courses. Subjects covered in the radiation worker training include, butare not limited to the following:-Atomic Structure and Radioactivity-Interactions of Radiation with Matter-Quantities and Units of Radiation-Basic Principles of Radiation Protection-Safe Handling of Radioactive Materials and Sources-Radiation Detection Instruments and Surveys-Dosimetry-Waste Disposal-Purchasing and Receiving Radioactive Materials-Regulations-Emergency Procedures-Record KeepingPage 11-12 THE UNIVERSITY OF TEXAS TRIGA 11 RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 11 .. ...... 1_/The Radiation Safety Officer may also require radiation workers to be trained in other areas,such as general hazard communication (Texas Hazard Communication Act) and laboratorysafety. The Radiation Safety Office shall maintain records of course attendance and coursecredit.11.1.2.4 Audits of the Radiation Protection ProgramReview and audit of the radiation protection program is conducted at least annually by atechnically competent person appointed by the Reactor Oversight Committee. The annualradiation protection program audit normally covers areas such as health physics training forNETL staff and users, health physics procedures, personnel monitoring, environmentalmonitoring, effluent monitoring, operational radiological surveys, instrument calibration,radioactive waste management and disposal, radioactive material transportation, and a reviewof unusual occurrences. The audit reports are sent to the ROC for review and follow-up action.11.1.2.5 Health Physics Records and Record KeepingRadiation protection program records such as radiological survey data sheets, personnelexposure reports, training records, inventories of radioactive materials, environmentalmonitoring results, waste disposal records, instrument calibration records and many more, aremaintained by the Health Physicist. The records will typically be retained for the life of thefacility either in hard copy, or on photographic or electronic storage media. Records for thecurrent and previous year are typically retained in the health physicist's office. Other recordsmay be retained in long-term storage. Radiation protection records are reviewed by the healthphysicist prior to filing. Radiation protection records are used for developing trend analysis,particularly in the personnel dosimetry area, for keeping management informed regardingradiation protection matters, and for reporting to regulatory agencies. In addition, they areused for planning radiation protection related actions, e.g., radiological surveys to preplan workor to evaluate the effectiveness of decontamination or temporary shielding efforts.11.1.3 ALARA ProgramThe objectives of the ALARA program are to maintain exposures to ionizing radiation andreleases of radioactive effluents at levels that are as low as reasonably achievable (ALARA)within the established dose equivalent and effluent release limits of the appropriate regulatoryauthority. The management of the NETL does not desire to limit the ability of researchers toperform experiments and participate in reactor operations. However, the management isfirmly and unequivocally committed to keeping exposures to personnel and the general publicALARA. The NETL Health Physicist is the individual given explicit responsibility and authority forimplementation of the radiation protection and ALARA programs.In support of ALARA, local occupational dose limits (whole body) have been established asfollows:Page 11-13 CHAPTER 11, RADIATION PROTECTION AND WASTE MANAGEMENT PROGRAM 12/20111. An annual limit, which is the more limiting of:a. the total effective dose equivalent being equal to 1 rem (10 mSv); orb. the sum of the deep dose equivalent and the committed dose equivalent to anyindividual organ or tissue other than the lens of the eye being equal to 1 rem (10mSv).2. The annual dose limits to the lens of the eye, to the skin, and to the extremities, whichare:a. an eye dose equivalent of 1.5 rem (15 mSv), andb. a shallow dose equivalent of 5 rem (50 mSv) to the skin or any extremity.These dose limits may only be exceeded by written permission of the NETL director who willassign a new individual local dose limit for the person.Procedures provide for a review of all experiments and reactor operations and maintenanceactivities for radiological considerations by the Health Physicist and Reactor Supervisor.11.1.4 Radiation Monitoring and SurveyingThe radiation monitoring program for the NETL is structured to ensure that all three categoriesof radiation sources (airborne, liquid and solid) are detected and assessed in a timely manner.To achieve this, the monitoring program is organized such that two major types of radiationsurveys are carried out: namely, routine radiation level and contamination level surveys ofspecific areas and activities within the facility, and special radiation surveys necessary tosupport non-routine facility operations.11.1.4.1 Monitoring for Radiation Levels and ContaminationThe routine monitoring program is structured to make sure that adequate radiationmeasurements of both radiation fields and contamination are made on a regular basis. Thisprogram includes but is not limited to the following:Typical surveys for radiation fields:-Weekly surveys in restricted areas-Monthly surveys of exterior walls and roof-Quarterly surveys of non-restricted areasPage 11-14 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 11I12/2011-Surveys required for certain incoming radioactive materials packages-Surveys to determine radiological impact of non-routine operationsTypical surveys for contamination:-Weekly surveys in restricted areas-Monthly surveys of reactor room roof-Quarterly surveys of exterior of facility-Quarterly surveys in non-restricted areas-Surveys required for certain incoming radioactive materials packages-Surveys to determine radiological impact of non-routine operations11.1.4.2 Radiation Monitoring EquipmentRadiation monitoring equipment used in the NETL is summarized below. Because equipment isupdated and replaced as technology and performance requires, the equipment listed should beconsidered representative rather than an exact listing.Table 11.2, Representative Radiation Detection InstrumentationVendorBicronBicronEberlineLudlumEberlineVarious PICsCanberraVictoreenEberlineLudlumBertholdProteanWallacP.R.M.LudlumEberlineModelFrisk-TechMicro-RemRO-2A12-4RM-14SDosicard450BE600375 DualLB-1043WPC 95501409AR-1000333-2RMSIIRange0-500,000 cpm0-20 mrem/hr0-50 R/hr0-10 rem/hr0-5,000,000 cpm0-200 mremN/A0-5 R/hr0-1000 R/hr0.1-1,000 mrem/hrN/AN/AN/AN/AN/A0.1-10000 mR/hrPurpose/FunctionPortable Contamination Survey InstrumentPortable Radiation Survey InstrumentPortable Radiation Survey InstrumentPortable Neutron Survey InstrumentPortable Contamination Survey InstrumentPersonnel dosimetryPersonnel dosimetryPortable Radiation Survey InstrumentExtendable Radiation Survey InstrumentCriticality MonitorHand/Foot MonitorGas Flow Proportional CounterLiquid Scintillation CounterAr-41 CAMParticulate CAMArea Radiation Monitor11.1.4.3 Instrument CalibrationPage 11-15 CHAPTER 11, RADIATION PROTECTION AND WASTE MANAGEMENT PROGRAM 12/2011Radiation monitoring instrumentation is calibrated according to written procedures developedfrom the guidance of industry standards such as ANSI N323A Radiation ProtectionInstrumentation Test and Calibration, Portable Survey Instruments. A calibration sticker shall beattached to all calibrated instruments showing the last calibration date, the initials of theperson who performed the calibration, and the next calibration due date. The NETL HealthPhysicist shall maintain all instrument calibration records.11.1.5 Radiation Exposure Control and DosimetryRadiation exposure control depends on many different factors including facility design features,operating procedures, training, proper equipment, etc. Training and procedures have beendiscussed previously under the section dealing with the NETL's radiation protection program.Therefore, this section will focus on design features such as shielding, ventilation, containmentand entry control devices for high radiation areas, and will also include protective equipment,personnel dosimetry, and estimates of annual radiation exposure. A description of thedosimetry records used to document facility exposures and a summary of exposure trends atthe NETL will also be presented.11.1.5.1 ShieldingThe biological shielding around the NETL reactor is the single biggest design feature incontrolling radiation exposure during operation of the facility. The shielding is based on TRIGAshield designs used successfully at many other similar reactors. The shield has been designedwith beam ports to allow extraction of radiation from the core for use in research, education,and service work. When beam port shielding is removed, additional control measures areneeded to control radiation exposure. Restricting access to the areas of elevated radiationlevels and/or additional shielding are typically used to control radiation exposure. Radiationsurvey data and the ALARA principle are used determine the appropriate control measures fornew configurations as necessary.11.1.5.2 ContainmentContainment of radioactivity within the NETL is primarily a concern with respect to experimentsbeing irradiated in the various irradiation facilities and with the reactor fuel. Containment offission products within the fuel elements is achieved by maintaining the integrity of the fuel'scladding, which is accomplished by maintaining the fuel and cladding temperatures belowspecified levels. Containment of other radionuclides generated during use of the irradiationfacilities is achieved through strict encapsulation procedures for samples and strict limits onwhat materials will be irradiated. To further improve containment and minimize the potentialrelease of radioactivity from experiments irradiated in the in-core pneumatic transfer system,the terminal where samples are manually loaded and unloaded is located inside a fume hood.The hood maintains an in-flow of air to prevent the release of radioactivity to the surroundingarea.Page 11-16 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 1111.1.5.3 Entry ControlFor security purposes, the entire NETL facility perimeter is access controlled. In addition,restricted areas within the NETL are access controlled with unescorted access granted only totrained radiation workers. Most of the restricted areas within the NETL are not high radiation areas. However, in areaswhich are known high radiation areas, additional. measures are in place to control access. Thebeam port enclosures are the areas typically controlled due to high radiation areas. Entrywaysto the beam port enclosures are normally locked. When the beam port shutter is open(creating the high radiation area), a conspicuous visible signal is activated at the entryway. If abeam port enclosure entryway is opened, a signal is sent to the control console immediatelynotifying the reactor operator.11.1.5.4 Personal Protective EquipmentTypical personal protective equipment used in the NETL radiation protection program consistsof anti-contamination items (gloves, lab coats, coveralls, etc.) used when working with unsealedsources of radiation. Other than Ar-41, no airborne radioactive material is expected duringnormal operation. Thus, no respiratory protection program has been implemented.11.1.5.5 Representative Annual Radiation DosesRegulation 10CFR.20.1502 requires monitoring of workers likely to receive, in one year fromsources external to the body, a dose in excess of 10 percent of the limits prescribed in1OCFR20.1201. The regulation also requires monitoring of any individuals entering a high orvery high radiation area within which an individual could receive a dose equivalent of 0.1 rem inone hour. According to Regulatory Guide 8.7, if a prospective evaluation of likely dosesindicates that an individual is not likely to exceed 10 percent of any applicable limit, then thereare no requirements for recordkeeping or reporting. Likewise, Regulatory Guide 8.34 indicatesthat, if individual monitoring results serve as confirmatory measures, but monitoring is notrequired by 10CFR20.1502, then such results are not subject to the individual doserecordkeeping requirements of IOCFR20.2106(a) even though they may be used to satisfy1OCFR20.1501 requirements.The following table lists recent occupational exposures at the NETL. There have been noinstances of any exposures in excess of 10 percent of the above limits. Thus, retrospectively,only confirmatory monitoring is required and 1OCFR20.2106(a) recordkeeping requirements donot apply, so long as there are no significant changes in the facility, operating procedures, oroccupational expectations.Table 11.3, Representative Occupational ExposuresNumbers of persons in annual-dose categoriesPage 11-17 CHAPTER 11, RADIATION PROTECTION AND WASTE MANAGEMENT PROGRAM12/2011Year Immeasurable < 0.1 0.1-0.5 > 0.5 remrem rem2010 13 5 0 02009 6 7 0 02008 4 9 0 02007 8 5 3 02006 4 10 2 02005 15 22 0 0Although it appears monitoring of workers is not required, it is the policy of the NETL tomonitor workers and members of the public for radiation exposure. Anyone entering arestricted area within the NETL is monitored for radiation exposure with a dosimeter and/orradiation survey and occupancy time data. Although the NETL is likely exempt from recordkeeping requirements of 10CFR20.2106(a), records of this monitoring are maintained.11.1.5.5.2 Personnel Dosimetry DevicesPersonnel dosimetry devices are available to provide monitoring of all radiation categorieslikely to be encountered. Direct reading dosimeters (pocket ion chambers or electronicdosimeters) are used by personnel and visitors when in restricted areas. OSL dosimeters withneutron capabilities are assigned to personnel who regularly work in restricted areas. TLDextremity dosimeters are assigned to personnel where extremity exposure may be thedominant issue. The OSL and TLD dosimeters are provided and processed by a NVLAPaccredited vendor. Uptakes of radioactive material are not expected during normal operations.Thus, no internal dosimetry program has been implemented.11.1.6 Contamination ControlRadioactive contamination is controlled at the NETL by using written procedures for radioactivematerial handling, by using trained personnel, and by operating a monitoring program designedto detect contamination in a timely manner. While there are no accessible areas of the NETLthat are routinely grossly contaminated, personnel are trained in contamination detection andcontrol, methods for avoiding contamination, and procedures for handling, storing, anddisposing of identified contaminated material. After working in contaminated areas, personnelare required to perform surveys to ensure that no contamination is present on clothing, shoes,etc., before leaving the work location. Activities that are likely to create significantcontamination may have special work procedures applied such as a Radiation Work Permit.Contamination events are documented in a special survey report.11.1.7 Environmental MonitoringThe NETL has routinely performed environmental radiation monitoring throughout itsoperational history. While many different types of samples have been collected and analyzed,to date there has been no indication that NETL operations have significantly impacted theenvironment and there are no trends in environmental data which indicate that future impactsPage 11-18 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 11will occur. This result is consistent with expectations for a facility of this type. With theexception of Ar-41, there are virtually no pathways for radioactive materials from the NETL toenter the unrestricted environment during normal facility operations. However, the NETLenvironmental monitoring program has been structured to provide surveillance over a broadrange of environmental media even though there is no credible way the facility could beimpacting these portions of the environment. The current environmental monitoring programconsists of the following basic components which may change from time to time to meetprogram objectives:-Direct gamma radiation measurements performed monthly around the perimeter of thefacility.-Integrated gamma dose measurements using dosimeters located at the perimeter andin the general area of the facility which are exchanged quarterly.-Ground water sample obtained quarterly from under the reactor structure.-Monthly contamination monitoring on the roof of the reactor building.-Quarterly contamination monitoring at the perimeter and in the general area of thefacility.Results of this monitoring are reviewed and records are maintained as part of the radiationmonitoring program. In addition, the Texas Department of State Health Services conductsenvironmental monitoring independently of the NETL program. The TDSHS monitoringprogram includes quarterly integrated gamma dose using dosimeters at locations around thefacility and ground water samples from near the facility. Reports from the TDSHS monitoringare made available to the NETL for comparison with in-house results.11.2 Radioactive Waste ManagementThe NETL routinely generates very modest quantities of radioactive waste due to the type ofprogram carried out at the facility and to the fact that a conscious effort is made to keep wastevolumes to a minimum. Much of the waste that is generated consists of radioactive materialswith a relatively short half-life. Thus, much of the radioactive waste generated at the NETL isheld in a restricted area and allowed to decay to background levels and then disposed as non-radioactive waste. Radioactive waste that is not decayed in storage is typically transferred tothe university Radiation Safety Office for appropriate disposal.11.2.1 Radioactive Waste Management ProgramThe objective of the radioactive waste management program is to ensure that radioactivewaste is minimized, and that it is properly handled, stored and disposed of. The NETL healthPage 11-19 CHAPTER 11, RADIATION PROTECTION AND WASTE MANAGEMENT PROGRAM 12/2011physicist is responsible for administering the radioactive waste management program. Writtenprocedures address handling, storing and disposing of radioactive waste. The radioactive wastemanagement program is audited as part of the oversight function of the Reactor OversightCommittee. Waste management training is part of both the initial radiation protection trainingand operator requalification training. Radioactive waste management records are maintainedby the health physicist. As stated previously, minimization of radioactive waste is a policy ofthe NETL. Although there are no numerical volume goals set due to the small volume of wastegenerated, the health physicist and the reactor supervisor periodically assess operations for thepurpose of identifying opportunities or new technologies that will reduce or eliminate thegeneration of radioactive waste.11.2.2 Radioactive Waste ControlsAt the NETL, radioactive waste is generally considered to be any item or substance which is nolonger of use to the facility and which contains radioactivity above the established naturalbackground radioactivity. Because NETL waste volumes are small and the nature of the wasteitems is limited and reasonably repetitive, there is usually little question about what is or is notradioactive waste. Equipment and components are categorized as waste by the reactoroperations staff or health physics staff, while standard consumable supplies like plastic bags,gloves, absorbent material, disposable lab coats, etc., automatically become radioactive wasteif detectable radioactivity above background is found to be present. When possible, radioactivewaste is initially segregated at the point of origin from items that will not be considered waste.Screening is based on the presence of detectable radioactivity using appropriate monitoringand detection techniques and on the projected future need for the items and materialsinvolved. All items and materials initially categorized as radioactive waste are monitored asecond time before packaging for disposal to confirm data needed for waste records, and toprovide a final opportunity for decontamination/reclamation of an item. This helps reduce thevolume of radioactive waste by eliminating disposal of items that can still be used.11.2.2.1 Gaseous WasteGaseous waste is not created at the NETL under normal operations. Although Ar-41 is releasedfrom the NETL stack, this release is not considered to be waste in the same sense as the solidwaste which is collected and disposed of by the facility. The Ar-41 is usually classified as aneffluent which is a routine part of the normal operation of the NETL reactor.11.2.2.2 Liquid WasteBecause normal operations create only small volumes of liquid which contain radioactivity, it istypically possible to convert the liquids to a solid waste form. In limited cases, larger volumesof radioactive liquid waste could be generated. In these cases, decay in storage or disposal bythe sanitary sewer in accordance with 1OCFR20 may be required.11.2.2.3 Solid WastePage 11-20 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 11 1As with most research reactors, solid waste is routinely generated from reactor maintenanceoperations and irradiations of various experiments. Average annual solid radioactive wastevolume produced at the NETL is approximately However, as mentionedpreviously, much of this waste contains radioactive material with a relatively short half-life.Thus, much of this solid waste is held in a restricted area until it has decayed to backgroundlevels of radioactivity. Once decayed and surveyed to confirm background levels ofradioactivity, the waste is disposed as non-radioactive. The remaining solid waste whichcontains radioactive materials with a relatively long half-life typically amounts to approximatelytwo cubic feet per year. Appropriate radiation monitoring instrumentation will be used foridentifying and segregating solid radioactive waste. Solid radioactive waste to be held for decayis typically packaged in plastic bags, labeled appropriately, and moved to a designate storagearea within a restricted area. Solid radioactive waste to be transferred for disposal is packagedaccording to USDOT, waste processor, and disposal site requirements as applicable and istemporarily stored in a restricted area until transfer for disposal. No solid radioactive waste isintended to be retained or permanently stored on site.11.2.2.4 Mixed WasteAs mixed waste has in addition to being radioactive, the characteristic of being chemicallyhazardous and falling under RCRA regulations, great care is taken at the NETL to avoidgenerating mixed waste whenever possible. However, generation of mixed waste cannot becompletely avoided. The University of Texas at Austin is considered a RCRA "Large QuantityGenerator." Thus, any mixed waste generated at the NETL must be disposed within 90 days.Processes that may generate mixed waste are reviewed with the intent of modifying theprocess or substituting materials were appropriate to minimize the mixed waste generated. Inmany cases, the mixed waste contains radioactive materials with a half-life such that decay tobackground levels within the 90-day disposal requirement is possible. Where decay is not anoption, the mixed waste is packaged appropriately and transferred to the university RadiationSafety Office for disposal.11.2.2.5 Decommissioning WasteThere is no intention of decommissioning the NETL in the near future. Thus, there is noexpectation of decommissioning waste being generated.11.2.3 Release of Radioactive WasteControlled releases of radioactive waste to the environment are not a routine occurrence at theNETL. However, there is the possibility of infrequent releases of liquid waste to the sanitarysewer in compliance with applicable regulations. The typical release of radioactive waste fromthe NETL is via transfer of solid waste to the university Radiation Safety Office for appropriatedisposal.Page 11-21 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 1212 CONDUCT OF OPERATIONS12.1 ORGANIZATONThis chapter describes and discusses the Conduct of Operations at the University of TexasTRIGA. The Conduct of Operations involves the administrative aspects of facility operations,the facility emergency plan, the security plan, the Reactor Operator selection andrequalification plan, and environmental reports. License is used in Chapter 12 in reference toreactor operators and senior reactors subject to 10CFR50.55 requirements.12.1.1 Structure12.1.1.1 University AdministrationFig. 12.1 illustrates the organizational structure that is applied to the management andoperation of the University of Texas and the reactor facility. Responsibility for the safeoperation of the reactor facility is a function of the management structure of Fig. 12.11. Theseresponsibilities include safeguarding the public and staff from undue radiation exposures andadherence to license or other operation constraints. Functional organization separates theresponsibilities of academic functions and business functions. The office of the Presidentadministers these activities and other activities through several vice presidents.12.1.1.2 NETL Facility AdministrationThe facility administrative structure is shown in Fig. 12.2. Facility operation staff is anorganization of a director and at least four full time equivalent persons. This staff of fourprovides for basic operation requirements. Four typical staff positions consist of an associatedirector, a reactor supervisor, a reactor operator, and a health physicist. One or more of thelisted positions may also include duties typical of a research scientist. The reactor supervisor,health physicist, and one other position are to be full time. One full time equivalent positionmay consist of several part-time persons such as assistants, technicians and secretaries. Faculty,students, and researchers supplement the organization. Titles for staff positions are descriptiveand may vary from actual designations. Descriptions of key components of the organizationfollow."Standard for Administrative Controls" ANSI/ANS -15.18 1979Page 12-1 CHAPTER 12, CONDUCT OF OPERATIONSI12/2011CHAPTER 12, CONDUCT OF OPERATIONS 12/2011office of the PresidentThe University of Texas at AustinExecutive Vice President Vice President for IIand Provost Uniersity OperationsAssociate Vice President*Safety and SecurityDirector Environmental ]University Police !iat ndSftHealth and SafetyRadiation Safety Officer Radiation Safety CommitteeDean of the CockrellSchool of EngineeringChairman of the Dept.of Mechanical EngineeringReactor Oversight Director of NETLCommitteeAssociate Directorof NETLIReactor Supervis Health PhysicistFigure 12.1, University AdministrationNElI- DirectorNE iL.Associate DirectorReactor Supervisor NEI L Lab Manager Health ,PhhysicistReactor Operators Frecinical Support F Lab AsisstantsRa-oecnanr _ IFigure 12.2, NETL Facility AdministrationPage 12-2 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 1212.1.2 Responsibility12.1.2.1 Executive Vice President and ProvostResearch and academic educational programs are administered through the Office of theExecutive Vice President and Provost. Separate officers assist with the administration ofresearch activities and academic affairs with functions delegated to the Dean of the CockrellSchool of Engineering and Chairman of the Mechanical Engineering Department.12.1.2.2. Vice President for University OperationsUniversity operations activities are administered through the Office of the Vice President forOperations. This office is responsible for multiple operational functions of the Universityincluding university support programs, human resources, campus safety and security, campusreal estate, and campus planning and facilities management.12.1.2.3 Associate Vice President Campus Safety and SecurityThe associate vice president for campus safety and security oversees multiple aspects of safetyand security on campus including environmental health and safety, campus police, parking andtransportation, fire prevention, and emergency preparedness.12.1.2.4 Director of Nuclear Engineering Teaching LaboratoryNuclear Engineering Teaching Laboratory programs are directed by a senior classified staffmember or faculty member. The director oversees strategic guidance of the NuclearEngineering Teaching Laboratory including aspects of facility operations, research, and servicework. The director must interact with senior University of Texas at Austin managementregarding issues related to the Nuclear Engineering Teaching Laboratory.12.1.2.5 Associate Director of Nuclear Engineering LaboratoryThe Associate Director performs the day to day duties of directing the activities of the facility.The Associate Director is knowledgeable of regulatory requirements, license conditions, andstandard operating practices. The associate director will also be involved in soliciting andcarrying out research utilizing the reactor and other specialized equipment at the NuclearEngineering Teaching Laboratory.Page 12-3 CHAPTER 12, CONDUCT OF OPERATIONS [12/201112.1.2.6 Reactor Oversight CommitteeThe Reactor Oversight Committee is established through the Office of the Dean of the CockrellSchool of Engineering of The University of Texas at Austin. Broad responsibilities of thecommittee include the evaluation, review, and approval of facility standards for safe operation.The Dean shall appoint at least three members to the Committee that represent a broadspectrum of expertise appropriate to reactor technology. The committee will meet at leasttwice each calendar year or more frequently as circumstances warrant. The Reactor OversightCommittee shall be consulted by the Nuclear Engineering Teaching Laboratory concerningunusual or exceptional actions that affect administration of the reactor program.12.1.2.7 Radiation Safety OfficerA Radiation Safety Officer acts as the delegated authority of the Radiation Safety Committee inthe daily implementation of policies and practices regarding the safe use of radioisotopes andsources of radiation as determined by the Radiation Safety Committee. The Radiation SafetyProgram is administered through the University Environmental Health and Safety division. Theresponsibilities of the Radiation Safety Officer are outlined in The University of Texas at AustinManual of Radiation Safety.12.1.2.8 Radiation Safety CommitteeThe Radiation Safety Committee is established through the Office of the President of TheUniversity of Texas at Austin. Responsibilities of the committee are broad and include allpolicies and practices regarding the license, purchase, shipment, use, monitoring, disposal, andtransfer of radioisotopes or sources of ionizing radiation at The University of Texas at Austin.The President shall appoint at least three members to the Committee and appoint one asChairperson. The Committee will meet at least once each year on a called basis or as requiredto approve formally applications to use radioactive materials. The Radiation Safety Committeeshall be consulted by the University Safety Office concerning any unusual or exceptional actionthat affects the administration of the Radiation Safety Program.12.1.2.9 Reactor SupervisorReactor operation at the Nuclear Engineering Teaching Laboratory is directed by a ReactorSupervisor. Responsibilities of the Reactor Supervisor include control of license documentation,reactor operation, equipment maintenance, experiment operation, and instruction of personswith access to laboratory areas.Activities of reactor operators with USNRC licenses will be subject to the direction of a personwith a USNRC senior operator license. The Reactor Supervisor shall be qualified as a seniorPage 12-4 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 12operator. This person is to be knowledgeable of regulatory requirements, license conditions,and standard operating practices.12.1.2.10 Health PhysicistRadiological safety of the Nuclear Engineering Teaching Laboratory is monitored by a healthphysicist, who will be knowledgeable of the facility radiological hazards. Responsibilities of thehealth physicist will include calibration of radiation detection instruments, measurements ofradiation levels, control of radioactive contamination, maintenance of radiation records, andassistance with other facility monitoring activities.Activities of the health physicist will depend on two conditions. One condition will be thenormal operation responsibilities determined by the director of the facility. A second conditionwill be communications specified by the radiation safety officer. This combination ofresponsibility and communication provides for safety program implementation by the director,but establishes independent review. The health physicist's activities will meet therequirements of the director and the policies of an independent university safety organization.12.1.2.11 Laboratory ManagerLaboratory operations and research support is provide by a designated Laboratory Manager.The function is typically combined with the Health Physicist position.12.1.2.12 Reactor OperatorsReactor operators (and senior reactor operators) are licensed by the USNRC to operate the UTTREIGA II nuclear research reactor. University staff and/or students may be employed asreactor operators.12.1.2.13 Technical SupportStaff positions supporting various aspects of facility operations are assigned as required.12.1.2.14 Radiological Controls TechniciansRadiological Controls Technicians are supervised by the Health Physicist to perform radiologicalcontrols and monitoring functions. Radiological Controls Technicians are generally supportedas Undergraduate Research Assistant positions.Page 12-5 CHAPTER 12, CONDUCT OF OPERATIONS 12/201112.1.2.15 Laboratory AssistantsLaboratory Assistants are supervised by the Laboratory Manager to perform laboratoryoperations and analysis. Laboratory Assistants are generally supported as UndergraduateResearch Assistant positions.12.1.3 StaffingOperation of the reactor and activities associated with the reactor, control system, instrumentsystem, radiation monitoring system, and engineered safety features will be the function ofstaff personnel with the appropriate training and certification2.Whenever the reactor is not secured, the reactor shall be under the direction of a (USNRClicensed) Senior Operator who is designated as Reactor Supervisor. The Supervisor may be oncall if capable of arriving at the facility within thirty minutes and cognizant of reactoroperations. The Reactor Supervisor shall directly supervise:a. All fuel element or control rod relocations or installations within the reactor core region,and subsequent initial startup and approach to power.b. Relocation or installation of any experiment in the core region with a reactivity worth ofgreater than one dollar, and subsequent initial startup and approach to power.c. Recovery from an unscheduled shutdown or significant power reductions,d. All initial startup and approach to power following modifications to reactor safety orcontrol rod drive systems.Whenever the reactor is not secured, a (USNRC licensed) Reactor Operator (or Senior ReactorOperator) who meets requirements of the Operator Requalification Program shall be at thereactor control console, and directly responsible for control manipulations. All activities thatrequire the presence of licensed operators will also require the presence in the facility complexof a second person capable of performing prescribed written instructions.Only the Reactor Operator at the controls or personnel authorized by, and under directsupervision of, the Reactor Operator at the controls shall manipulate the controls. Wheneverthe reactor is not secured, operation of equipment that has the potential to affect reactivity orpower level shall be manipulated only with the knowledge and consent of the Reactor Operatorat the controls. The Reactor Operator at the controls may authorize persons to manipulatereactivity controls who are training either as (1) a student enrolled in academic or industry2 Selection and Training of Personnel for Research Reactors", ANSI/ANS -15.4 -1970 (N380)Page 12-6 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 12course making use of the reactor, (2) to qualify for an operator license, or (3) in accordance theapproved Reactor Operator requalification program.Whenever the reactor is not secured, a second person (i.e., in addition to the reactor operatorat the control console) capable of initiating the Reactor Emergency Plan will be present in theNETL building. Unexpected absence of this second person for greater than two hours will beacceptable if immediate action is taken to obtain a replacement.Staffing required for performing experiments with the reactor will be determined by aclassification system specified for the experiments. Requirements will range from the presenceof a certified operator for some routine experiments to the presence of a senior operator andthe experimenter for other less routine experiments.12.1.4 Selection and Training of Personnel12.1.4.1 QualificationsPersonnel associated with the research reactor facility3 shall have a combination of academictraining, experience, skills, and health commensurate with the responsibility to providereasonable assurance that decisions and actions during all normal and abnormal conditions willbe such that the facility and reactor are operated in a safe manner.12.1.4.2 Job DescriptionsQualifications for University positions are incorporated in job descriptions, summarizingfunction and scope. The typical description includes title, duties, supervision, education,experience, equipment, working conditions, and other special requirements for the jobposition. Student employment is typically under the general description of Undergraduate orGraduate Research Assistant, with minimal specification to accommodate a wide range of jobs.12.1.4.2.1 Facility DirectorA combination of academic training and nuclear experience will fulfill the qualifications for theindividual identified as the facility director. A total of six years' experience will be required.Academic training in engineering or science, with completion of a baccalaureate degree, mayaccount for up to four of the six years' experience. The director is generally a faculty memberwith a Ph.D. in nuclear engineering or a related field.3 ANS/ANSI-15.4, op. cit.Page 12-7 CHAPTER 12, CONDUCT OF OPERATIONS 12/201112.1.4.2.2 Associate DirectorA combination of academic training and nuclear experience will fulfill the qualifications for theindividual identified as the facility director. Academic training in engineering or science, withoperating and management experience at a research reactor is required. The AssociateDirector will be qualified by certification as a senior operator and is typically a person with atleast one graduate degree in nuclear engineering or a related field.12.1.4.2.3 Reactor SupervisorA person with special training to supervise reactor operation and related functions will bedesignated as the reactor supervisor. The reactor supervisor will be qualified by certification asa senior operator as determined by the licensing agency. Additional academic or nuclearexperience will be required as necessary for the supervisor to perform adequately the dutiesassociated with facility activities. The supervisor is typically a person with at least one graduatedegree in nuclear engineering or a related field.12.1.4.2.4 Health PhysicistA person with a degree related to health, safety, or engineering, or sufficient experience that isappropriate to the job requirements will be assigned the position of health physicist. A degreein health physics or similar field of study and some experience is preferred. Certification isnot a qualification, but work towards certification should be considered a requirement.12.1.4.3.4 Laboratory ManagerLaboratory operations and research support id provide by a designated Laboratory Manager.The function is typically combined with the Health Physicist position.12.1.2.12 Reactor OperatorsReactor operators (and senior reactor operators) are licensed by the USNRC to operate the UTTREIGA II nuclear research reactor. Training and requalification requirements are indicatedbelow.12.1.2.13 Technical SupportStaff positions supporting various aspects of facility operations are assigned as required.Selection, qualification and training are on a case by case basis.12.1.2.14 Radiological Controls TechniciansRadiological Controls Technicians training is provided in the Radiation Protection Program.Page 12-8 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 1 12/2011SAFETY ANALYSIS REPORT, CHAPTER 1212.1.2.15 Laboratory AssistantsLaboratory Assistants are supervised by the Laboratory Manager to perform laboratoryoperations and analysis, with specific training requirements related to job responsibilities..12.1.5 Radiation SafetyProtection of personnel and the general public against hazards of radioactivity and fire isestablished through the safety programs of the University Safety Office. Safety programs at the-reactor facility supplement the university programs so that appropriate safety measures areestablished for the special characteristics of the facility4 5.Safety programs are operated as a function of the Vice President for University Operations andinclude a radiation safety organization as presented in Fig. 12.1. Radiation protection at thereactor facility is the responsibility of the Reactor Supervisor, Health Physicist, or a designatedsenior operator in charge of operation activities. The person responsible for radiationprotection at the reactor facility will have access to other individuals or groups responsible forRadiological safety at the University. Contact with the Radiation Safety Officer will occur on anas needed basis and contact with the Reactor Oversight Committee will occur on a periodicbasis. Responsibility includes the authority to act on questions of radiation protection, theAcquisition of appropriate training for radiation protection and the reporting to management ofproblems associated with radiation protection. Radiological management policies andprograms are described in Chapter 11.12.2 REVIEW AND AUDIT ACTIVITESThe review and audit process is the responsibility of the Reactor Oversight Committee (ROC).12.2.1 Composition and QualificationsThe ROC shall consist of at least three (3) members appointed by the Dean of the CockrellSchool of Engineering that are knowledgeable in fields which relate to nuclear safety. Theuniversity radiological safety officer shall be a member or an ex-officio member. The committeewill perform the functions of review and audit or designate a knowledgeable person for auditfunctions.4 "Radiological Control at Research Reactor Facilities", ANSI/ANS-15.11 1977(N628)5 "Design Objectives for and Monitoring of Systems Controlling Research Reactor Effluents", ANSI/ANS -15.121977(N647)Page 12-9 CHAPTER 12, CONDUCT OF OPERATIONS12/201112.2.2 Charter and RulesThe operations of the ROC shall be in accordance with an established charter, includingprovisions for:a. Meeting frequency (at least twice each year, with approximately 4-8 month frequency).b. Quorums (not less than one-half the membership where the operating staff does notcontribute a majority).c. Dissemination, review, and approval of minutes.d. Use of subgroups.12.2.3 Review FunctionThe responsibilities of the Reactor Safeguards Committee to shall include but are not limited toreview of the following:a) All new procedures (and major revisions of procedures) with safety significanceb) Proposed changes or modifications to reactor facility equipment, or systemshaving safety significancec) Proposed new (or revised) experiments, or classes of experiments, that couldaffect reactivity or result in the release of radioactivityd) Determination of whether items a) through c) involve unreviewed safetyquestions, changes in the facility as designed, or changes in TechnicalSpecifications.e) Violations of Technical Specifications or the facility operating licenseef) Violations of internal procedures or instruction having safety significanceg) Reportable occurrencesh) Audit reportsMinor changes to procedures and experiments that do not change the intent and do notsignificantly increase the potential consequences may be accomplished following review andapproval by a senior reactor operator and independently by one of the Reactor Supervisor,Associate Director or Director. These changes should be reviewed at the next scheduledmeeting of the Reactor Oversight Committee.Page 12-10 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 1212.2.4 Audit FunctionThe audit function shall be a selected examination of operating records, logs, or otherdocuments. Audits will be by a Reactor Oversight Committee member or by an individualappointed by the committee to perform the audit. The audit should be by any individual notdirectly responsible for the records and may include discussions with cognizant personnel orobservation of operations. The following items shall be audited and a report made within 3months to the Director and Reactor Committee:a. Conformance of facility operations with license and technical specifications at least onceeach calendar year.b. Results of actions to correct deficiencies that may occur in reactor facility equipment,structures, systems, or methods of operation that affect safety at least once percalendar year.c. Function of the retraining and requalification program for reactor operators at leastonce every other calendar year.d. The reactor facility emergency plan and physical security plan, and implementingprocedures at least once every other year.12.3 PROCEDURESWritten procedures shall govern many of the activities associated with reactor operation.Activities subject to written procedures will include:a) Startup, operation, and shutdown of the reactorb) Fuel loading, unloading, and movement within the reactor.c) Control rod removal or replacement.d) Routine maintenance, testing, and calibration of control rod drives and other systemsthat could have an effect on reactor safety.e) Administrative controls for operations, maintenance, conduct of experiments, andconduct of tours of the Reactor Facility.f) Implementing procedures for the Emergency Plan or Physical Security Plan.Written procedures shall also govern:Page 12-11 CHAPTER 12, CONDUCT OF OPERATIONS12/2011a) Personnel radiation protection, in accordance with the Radiation Protection Program asindicated in Chapter 11b) Administrative controls for operations and maintenancec) Administrative controls for the conduct of irradiations and experiments that could affectcore safety or reactivityA master Procedure Control procedure specifies the process for creating, changing, editing, anddistributing procedures. Preparation of the procedures and minor modifications of theprocedures will be by certified operators. Substantive changes or major modifications toprocedures, and new prepared procedures will be submitted to the Reactor OversightCommittee for review and approval. Temporary deviations from the procedures may be madeby the reactor supervisor or designated senior operator provided changes of substance arereported for review and approval.Proposed experiments will be submitted to the reactor oversight committee for review andapproval of the experiment and its safety analysis6, as indicated in Chapter 10. Substantivechanges to approved experiments will require re-approval while insignificant changes that donot alter experiment safety may be approved by a senior operator and independently one ofthe following, Reactor Supervisor, Associate Director, or Director. Experiments will be approvedfirst as proposed experiments for one time application, and subsequently, as approvedexperiments for repeated applications following a review of the results and experience of theinitial experiment implementation.12.4 REQUIRED ACTIONSThis section lists the actions required in the event of certain occurrences.12.4.1 Safety Limit ViolationIn the event that a Safety Limit is not met,a. The reactor shall be shutdown, and reactor operations secured.b. The Reactor Supervisor, Associate Director, and Director shall be notifiedc. The safety limit violation shall be reported to the Nuclear Regulatory Commission within24 hours by telephone, confirmed via written statement by email, fax or telegraphd. A safety limit violation report shall be prepared within 14 days of the event to describe:6 ANSI/ANS 15.6, op. cit.Page 12-12 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 121. Applicable circumstances leading to the violation including (where known) cause andcontributing factors2. Effect of the violation on reactor facility components, systems, and structures3. Effect of the violation on the health and safety of the personnel and the public4. Corrective action taken to prevent recurrencee. The Reactor Oversight Committee shall review the report and any followup reportsf. The report and any followup reports shall be submitted to the Nuclear RegulatoryCommission.g. Operations shall not resume until the USNRC approves resumption.12.4.2 Release of Radioactivity Above Allowable LimitsActions to be taken in the case of release of radioactivity from the site above allowable limitsshall include a return to normal operation or reactor shutdown until authorized bymanagement if necessary to correct the occurrence. A prompt report to management andlicense authority shall be made. A review of the event by the Reactor Oversight Committeeshould occur at the next scheduled meeting. Prompt reporting of the event shall be bytelephone and confirmed by written correspondence within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. A written follow upreport is to be submitted within 14 days.12.4.3 Other Reportable OccurrencesIn the event of a reportable occurrence, as defined in the Technical Specifications, and inaddition to the reporting requirements,a. The Reactor Supervisor, the Associate Director and the Director shall be notifiedb. If a reactor shutdown is required, resumption of normal operations shall beauthorized by the Associate Director or Directorc. The event shall be reviewed by the Reactor Oversight Committee during a normallyscheduled meeting12.5 REPORTSThis section describes the reports required to NRC, including report content, timing of reports,and report format. Refer to section 12.4 above for the reporting requirements for safety limitviolations, radioactivity releases above allowable limits, and reportable occurrences. All writtenreports shall be sent within prescribed intervals to the United States Nuclear RegulatoryCommission, Washington, D.C., 20555, Attn: Document Control Desk.Page 12-13 CHAPTER 12, CONDUCT OF OPERATIONS 12/201112.5.1 Operating ReportsRoutine annual reports covering the activities of the reactor facility during the previouscalendar year shall be submitted to licensing authorities within three months following the endof each prescribed year. Each annual operating report shall include the following information:a. A narrative summary of reactor operating experience including the energy produced bythe reactor or the hours the reactor was critical, or both.b. The unscheduled shutdowns including, where applicable, corrective action taken topreclude recurrence.C. Tabulation of major preventive and corrective maintenance operations having safetysignificance.d. Tabulation of major changes in the reactor facility and procedures, and tabulation ofnew tests or experiments, or both, that are significantly different from those performedpreviously, including conclusions that no new or unanalyzed safety questions wereidentified.e. A summary of the nature and amount of radioactive effluents released or discharged tothe environs beyond the effective control of the owner-operator as determined at orbefore the point of such release or discharge. The summary shall include, to the extentpracticable, an estimate of individual radionuclides present in the effluent. If theestimated average release after dilution or diffusion is less than 25% of theconcentration allowed or recommended, a statement to this effect is sufficient.f. A summarized result of environmental surveys performed outside the facility.g. A summary of exposures received by facility personnel and visitors where suchexposures are greater than 25% of that allowed or recommended.12.5.2 Other or Special ReportsA written report within 30 days to the chartering or licensing authorities of:a. Permanent changes in the facility organization involving Director or Supervisor.b Significant changes in the transient or accident analysis as described in the SafetyAnalysis Report.Page 12-14 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR J 12/2011SAFETY ANALYSIS REPORT, CHAPTER 1212.6. RECORDSRecords of the following activities shall be maintained and retained for the periods specifiedbelow7.The records may be in the form of logs, data sheets, electronic files, or other suitableforms. The required information may be contained in single or multiple records, or acombination thereof.12.6.1. Lifetime RecordsLifetime records are records to be retained for the lifetime of the reactor facility. (Note:Applicable annual reports, if they contain all of the required information, may be used asrecords in this section.)a. Gaseous and liquid radioactive effluents released to the environs.b. Offsite environmental monitoring surveys required by Technical Specifications.c. Events that impact or effect decommissioning of the facility.d. Radiation exposure for all personnel monitored.e. Updated drawings of the reactor facility.12.6.2 Five Year PeriodRecords to be retained for a period of at least five years or for the life of the componentinvolved whichever is shorter.a. Normal reactor facility operation (supporting documents such as checklists, log sheets,etc. shall be maintained for a period of at least one year).b. Principal maintenance operations.c. Reportable occurrences.d. Surveillance activities required by technical specifications.e. Reactor facility radiation and contamination surveys where required by applicable7 Records and Reports for Research Reactors", ANSI/ANS -15.3-1974 (N399).Page 12-15 CHAPTER 12, CONDUCT OF OPERATIONS 12/2011regulations.f. Experiments performed with the reactor.g. Fuel inventories, receipts, and shipments.h .Approved changes in operating procedures.n. Records of meeting and audit reports of the review and audit group.12.6.3 One Training CycleTraining records to be retained for at least one license cycle are the requalification records oflicensed operations personnel. Records of the most recent complete cycle shall be maintainedat all times the individual is employed.12.7 EMERGENCY PLANNINGEmergency planning is guided by an NRC approved Emergency Plan following the generalguidance set forth in ANSI/ ANS15.16, Emergency Planning for Research Reactors. The planspecifies two action levels, the first level being a locally defined Non-Reactor Specific Event,and the second level being the lowest level FEMA classification, a Notification of UnusualEvent. Procedures reviewed and approved by the reactor Oversight Committee are establishedto manage implementation of emergency response.12.8 SECURITY PLANNINGSecurity planning is guided by an NRC approved Security Plan. The plan incorporatescompensatory measures implemented following security posture changes initiated post 9/11.The Plan and portions of the procedures are classified as Safeguards Information. Securityprocedures implementing the plan, approved by the Reactor Oversight Committee, areestablished.12.9 QUALITY ASSURANCEObjectives of quality assurance (QA) may be divided into two major goals. First is the goal ofsafe operation of equipment and activities to prevent or mitigate an impact on public healthand safety. Second is the reliable operation of equipment and activities associated witheducation and research functions of the University. The risk or potential release of radioactivematerials is the primary impact on public health and safety, and may be divided into direct risksand indirect risks. Direct risks are activities such as waste disposal, fuel transport anddecommissioning that introduce radioactive materials into the public domain. Indirect risks areaccident conditions created by normal or abnormal operating conditions that generate thepotential or actual release of radioactive materials from the controlled areas of a facility.Page 12-16 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 12Quality assurance program procedures have been developed that apply to items or activitiesdetermined to be safety-related follows the guidelines of Reg. Guide 2.58 9. Specific proceduresapply to fuel shipment and receipt, a general procedure guides unspecified safety relatedactivities.12.10. OPERATOR REQUALIFICATIONRegulatory requirements and standards provide guidance for requalification training. Specificregulatory requirements are found in 10CFR55 for the licensing of operators and senioroperators with regulations for requalification set forth in section 55.59. Standards for theselection and training of facility personnel and reactor operators are available. Specificregulations in the form of two sets of license conditions also apply to the facility personnel andreactor operators. One set of conditions for the facility license, 10CFR 50.54, applies to facilitypersonnel. The other set of conditions for individual licenses, 10CFR 55.53 applies to operatorsand senior operators.An NRC approved UT TRIGA Requalification Plan is used to maintain training and qualification ofreactor operators and senior reactor operators. License qualification by written and operatingtest, and license issuance or removal, are the responsibility of the U.S. Nuclear RegulatoryCommission. No rights of the license may be assigned or otherwise transferred and thelicensee is subject to and shall observe all rules, regulations and orders of the Commission.Requalification training maintains the skills and knowledge of operators and senior operatorsduring the period of the license. Training also provides for the initial license qualification.Active status of any licensee requires successful participation in the UT Operator Requalificationprogram. A process is in place to manage re-establishment of active status where conditions ofan active license status are not met.The program addresses training by lectures, instruction, discussion and self-study. The programaddresses training topics. The program establishes requirement for a biennial schedule ofactivities. The program addresses on the job training. The program requires:a. Observation at least once each year of a satisfactory understanding of the reactivitycontrol system and knowledge of operating procedures.b. Each operator or senior operator will review facility design changes, procedure changesand license changes as they occur or once each 6 to 8 months.8 "Quality Assurance Requirements for Research Reactors", Nuclear Regulatory Guide 2.5 (77/05).9 "Quality Assurance Program Requirements for Research Reactors," ANSI/ANS -15.8 -1976 (N402).Page 12-17 CHAPTER 12, CONDUCT OF OPERATIONS12/2011c. A review of the contents of abnormal and emergency procedures will be done by eachoperator or senior operator at 6 to 8 month intervals so that at least 3 reviews occurduring the two year training cycle.The program addresses performance evaluation of on annual examination and periodicobservations, including methods to address deficiencies identified in evaluation. The programaddresses records to be generated, including required information and retention schedule.12.11 STARTUP PROGRAMStartup and testing of the Balcones Research Center TRIGA facility was completed in 1992,therefore a startup plan is not applicable.12.12 ENVIRONMENTAL REPORTThe Environmental Report is provided as a separate document.Page 12-18 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 13M12/201113.0 ACCIDENT ANALYSISThis chapter provides information and analysis to demonstrate that the health and safety of thepublic and workers are not challenged by equipment malfunctions or other abnormalities inreactor performance. The analysis demonstrates that facility design features, limiting safetysystem settings, and limiting conditions for operation ensure that unacceptable radiologicalconsequences to the general public, facility personnel or the environment will not occur as aresult of credible accidents. Reference values for physical properties and values used in analysisare provided in 13.1. An overview of accident scenarios is provided in 13.2, followed by detailedanalyses.13.1 Notation and Fuel PropertiesTables 13.1-13.3 identify physical characteristics of the TRIGA Mark II fuel. Table 13.4 identifiesthe assumptions and design basis values used in the accident analyses.Table 13.1. Neutronic Properties of TRIGA Mkll ZrH1.6 Fuel Elements.Property Symbol ValueEffective delayed neutron fractions f3 0.007Effective neutron lifetime £ 43 jisecTemperature coefficient of reactivity , -0.000115 K-1Source: West et al. (1967).Table 13.2, Dimensions of TRIGA Mkll ZrH1.6 Fuel Elements.Property of Individual Element Symbol ValueLength of fuel zone Lf Fuel radius ri Clad outside radius r. Fuel volume Vf Clad volume V, Fuel mass Mf Clad mass M, Wt. Fraction U in fuel XU Wt. Fraction ZrH1.6in fuel Xm Page 13-1 CHAPTER 13, ACCIDENT ANALYISI12/2011Table 13.3, Thermal and Mechanical Properties of TRIGA Mkll ZrH1.6 Fuel Elements and Type 304Stainless Steel Cladding.Property Symbol .Value Temp.FuelDensityThermal conductivityPf kf Cpf All0 OCHeat capacity, Cpf = 340.1 + 0.6952T(°C)CladdingDensity PC 300 KThermal conductivity kc 300 K400 K600 KHeat capacity C 300 K400 KYield strength 400 °CTensile strength 400 °CSource: fuel properties from Simnad (1980); cladding properties from Incropera and DeWitt(1990) and from Metals Handbook (1961).Table 13.4, UT TRIGA Core-Conditions Basis for Calculations.ParameterSteady state maximum power, P0Fuel mass per elementHeat capacity per element at T (°C)Minimum number of fuel elements, NCore radial peaking factorAxial peaking factorExcess reactivityMaximum pulsing reactivity insertionExcess reactivity at 500 kW maximum poweraFuel average temperature at 500 kW maximumaSource: Data from GA Torrey Pines TRIGA reactorValue1,100 kW2.367 kg805.0 + 1.646T (J K')832$0/2$4.00 (2.8% Ak/k)$3.00 (2.1% AN/k)$1.16 (0.81% Ak/k)285 °C13.2 Accident Initiating Events and ScenariosThree accident scenarios were identified in the initial licensing of the University of Texas TRIGAreactor in 1992: maximum hypothetical accident (fuel element failure in air), insertion of excessof reactivity, and loss of coolant. The current accident analysis substantially reprises the original,with updates to the methodology based on current standards.NUREG/CR-2387 (Credible Accident Analyses for TRGIA and TRIGA Fueled Reactors, Hawley &Kathren, 1982) was the definitive work in identifying and evaluating the spectrum of accidents tobe addressed for TRIGA reactors, addressing seven scenarios:Page 13-2 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 13* Excess reactivity insertion0 Metal-water reactions* Lost/misplaced or inadvertent experiments0 Mechanical rearrangement of the core0 Loss of coolant accident* Changes in fuel morphology and ZrHx composition0 Fuel handling accidentNUREG-1537 (Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors) provides guidance for format and contents as well as a standard review plan.The spectrum of accidents specified in NUREG-1537 includes:Maximum Hypothetical AccidentA fuel handling accident is considered to lead to the maximum hypothetical accident,with consequences analyzed in section 13.3.Insertion of Excess ReactivityExcess reactivity insertion accidents are analyzed for the UT TRIGA in section 13.4. Rapidinsertion of reactivity into a TRIGA reactor is a designed feature of the fuelperformance. Thus, most plausible reactivity accidents do not subject the fuel toconditions more severe than normal operating situations. Insertion of excess reactivity isconsidered for two sets of initial conditions. First, the maximum reactivity addition of2.8% Ak/k/k ($4.00) from operations below feedback range is considered with respect tomaximum fuel temperature. A second initial condition is considered where the 2.8%Ak/k/k ($4.00) reactivity addition occurs with the core operating at a power levelequivalent to the balance of the core excess reactivity Analysis demonstrates thatmaximum fuel temperature does not exceed acceptable limits. An administrative limitfor experiments of $3.00 assures that experiment removal cannot exceed the analysis bya large margin.Loss of CoolantLoss of coolant accident is analyzed in 13.5. A loss of coolant accident is analyzed todemonstrate that maximum fuel temperature does not exceed acceptable limits. Coolingin air is considered and the results are compared to limiting fuel temperatures. Dosesfrom scattered radiation from the uncovered core are analyzed.Page 13-3 CHAPTER 13, ACCIDENT ANALYIS 12/2011* Loss of Coolant FlowLoss of coolant flow is analyzed in 13.6.Mishandling or Malfunction of FuelTransport of fission products released in the pool is significantly affected by pool water.Failure during operation would occur under water, leading ultimately to atmosphericrelease of fission products. A failure of an element immersed in pool water would bepartially retain and/or retard migration of gaseous material and substantially retainparticulate material. Consequently, the failure in water is bounded by the failure in air.Fuel handling accident (the maximum hypothetical accident) is considered in section13.3.NUREG/CR-2387 identifies two categories of specific fuel failures, changes in fuelmorphology and metal-water reactions. Changes in fuel' morphology and ZrH)composition for fuel used by the UT TRIGA reactor is not credible for accident scenarios,as described in Chapter 4. As noted in the NUREG and in Chapter 4, significant metalwater reactions are not possible at TRIGA operating temperatures.Loss of Normal Electric PowerA loss of normal electric power is analyzed in section 13.7.External EventsExternal events are analyzed in section 13.8. External events are considered with respectto potential mechanical rearrangement of the core (specified in NUREG/CR-2387). Asdescribed in Chapter 4, the core support structure is secured to the floor, the core issurrounded by a canister of graphite, and fuel elements are positioned in the core by theupper and lower grid plate approximately 10 in. thick. There is no credible scenario thatwould disturb the core lattice or structure while simultaneously retaining fuel elementsin a critical geometry.* Mishandling or malfunction of experimentExperiment mishandling or malfunction is described in section 13.9. Lost/misplaced orinadvertent experiments; administrative controls on experiments as described in Chapter10 require an assessment of personnel and facility hazard, with specific limits onpotential hazard to personnel and the facility13.3 Maximum Hypothetical Accidents, Single Element Failure in AirPage 13-4 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 13The maximum hypothetical accident for a TRIGA reactor is the failure of the encapsulation ofone fuel element, in air, resulting in the release of airborne fission products to the reactor bayand the environment. Failure in air could result from a fuel-handling accident or from a failureduring operation in the core following a loss of coolant accident. This section addressespotential consequences, should failure occur in air.13.3.1 AssumptionsFuel mass of 1 metric ton is used in burnup calculations; results are scaled to actual fuelmass in determining fission product inventory.Continuous operation at specified power levels is assumed until end of useful fuel life(burnup of 10 grams 235U). At the end of useful life, one week of regular operations (8hours per day) is assumed. Radioisotope inventory decay calculations begin 5 minutesafter termination of power operations based on the loss of pool water scenario; fuelhandling after shutdown requires a substantially longer decay time for practical reasons.The fraction of noble gases and iodine contained within the fuel that is actually releasedis assumed to be 1.0 x 10-4.This is a very conservative value prescribed in NUREG 2387[Hawley and Kathren, 1982] and may be compared to the value of 1.5 x 10-5 measured atGeneral Atomics [Simnad et al., 1976] which is used in SARs for other reactor facilities[NUREG-1390, 1990].The fractional release of particulates (radionuclides other than noble gases and iodine) isassumed to be 1.0 x 10- , a very conservative estimate used by Hawley and Kathren(1982).The reactor bay free air volume is 4120 M3; 10% of this volume is not credited in dilutioncalculations. ,Radioisotopes specified in NURGE/CR-2387 with limits specified in IOCFR20 Appendix Bare used in consequence analysis, including iodine, noble fission product gases, andcesium and strontium. Halogen (bromine) was analyzed in the 1992 UT SAR, and istherefore included in this analysis. The relevant information from 10CFR Appendix B isprovided in Table 13.5.Table 13.5, Relevant IOCFR20 Appendix B ValuesNoble Gas & Iodine Radioisotopes Particulate RadioisotopesALl DAC EL ALl DAC ELIsotope 10 PCi/ml IpCi/mI Isotope Pci pCi/ml Br80 2.aE+05 8.OE-05 3.OE-07 Csl31 3.OE+04 1.OE-05 4.OE-08Br80m 2.OE+04 6.OE-06 2.OE-08 Cs132 4.OE+03 2.OE-06 6.OE-09Br82 4.0E03 2.OE-06 5.OE-09 Cs234m 1.0E+05 6.OE-05 2.OE-07Page 13-5 CHAPTER 13, ACCIDENT ANALYIS12/2011CHAPTER 13, ACCIDENT ANALYIS 12/2011Table 13.5, Relevant 1OCFR20 Appendix B ValuesNoble Gas & Iodine Radioisotopes Particulate RadioisotopesALl DAC EL ALl DAC ELIsotope Ci PCi/ml pCi/ml Isotope pCi tCi/ml pCi/mIBr83Br84112511281129113011311132113311341135Kr79Kr8lKr83mKr85Kr85mKr87Kr88Xe125Xe127Xe129mXe131mXe133Xe133mXe135Xe135mXe1386.OE+046.OE+046.OE+O11.OE+059.OE+007.OE+025.OE+018.OE+033.0E+025.OE+042.OE+033.OE-052.0E-053.OE-085.OE-054.OE-093.0E-072.OE-083.OE-061.OE-072.OE-057.OE-072.0E-057.OE-041.OE-021.OE-042.OE-055.OE-062.OE-062.OE-051.OE-052.OE-044.OE-041.OE-041.OE-041.OE-059.OE-064.OE-069.OE-088.OE-083.OE-108.OE-044.OE-112.OE-1O2.OE-102.OE-081.OE-096.OE-086.OE-097.OE-083.OE-065.OE-057.OE-071.OE-072.OE-089.OE-097.OE-086.OE-089.OE-072.OE-065.OE-076.OE-077.OE-084.OE-082.OE-08Cs135Cs135mCs136Cs137Cs138Sr85Sr85mSr87mSr89Sr90Sr9lSr921.OE+032.OE+057.OE+022.OE+026.0E+042.OE+036.OE+051.OE+051.0E+024.OE+004.OE+037.OE+035.OE-078.OE-053.OE-076.OE-082.OE-056.OE-073.OE-045.OE-056.OE-082.0E-091.OE-063.OE-062.OE-093.OE-079.OE-102.OE-108.OE-082.OE-099.OE-072.OE-072.OE-106.OE-125.OE-099.OE-09The 1OFR20 appendix B AnnualLimit on Intake (ALl) and Derived Air Concentration (DAC)values include the effects of the ingrowth of daughter radionuclides produced in thebody by the decay of the parent nuclide (10CFR20, Appendix B, Notation, Table 1), andtherefore daughters are not calculated or considered separately.13.3.2 AnalysisAnalysis of the maximum hypothetical accident begins with (A) a discussion of calculation's forfission product generated in the reactor, (B) methods and strategy for calculating the UT TRIGAfission product inventory, (C) fraction of fission products released from a single fuel element.The calculation of fission product inventory is used to evaluate the impact with respect to the10CFR20 (D) Annual Limit on Intake, (E) Derived Air Concentration, and (F) Effluent Limits. Theresults are reviewed (F), concluding that measures prescribed by the Radiation ProtectionPage 13-6 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 13 1Program would be required for worker protection in the worst case scenario, and that theeffluent limits are met.A. Radionuclide Inventory Buildup and Decay, TheoryConsider a mass of 235U yielding thermal power P (kW) due to thermal-neutron induced fission.The fission rate is related to the thermal power by the factor k = 3.12 x 1013 fissions per secondper kW. Consider also a fission product radionuclide, which is produced with yield Y, and whichdecays with rate constant k. It is easily shown that the equilibrium activity Ao, (Bq) of the fissionproduct, which exists when the rate of creation by fission is equal to the rate of loss by decay, isgiven by A=A.N. Here it should be noted that the power must be small enough or the uraniummass large enough that the depletion of the 235U is negligible. Starting at time t = 0, the buildupof activity is given by:A(t)= A. *(I e(I )For times much greater than the half-life of the radionuclide, and for times much less than thehalf-life, A(t) = A-,* ?A* t. If the fission process ceases at time tj, the specific activity at later timet is given byA(t)= Aý, * (I -e-" *e-'(1-1')Consider the fission product 1311, which has a half-life of 8.04 days (X = 0.00359 h1) and a chain(cumulative) fission product yield of about 0.031. At a thermal power of 1 kW, the equilibriumactivity is about A,,, = 9.67 x 1011 Bq (26.1 Ci). After only four hours of operation, though, theactivity is only about 0.37 Ci. For equilibrium operation at 3.5 kW, distributed over 81 fuelelements, the average activity per element would be 26.1 x 3.5 + 81 = 1.13 Ci per fuel element.The worst case element would contain twice this activity. With a release fraction of 1.0 x 10-4,the activity available for release would be about 1.13 x 2 x 1.0 x 10-4 = 2.26 x 10.4 Ci. This typeof calculation is performed by the ORIGEN ARP code for hundreds of fission products and forarbitrary times and power levels of operation as well as arbitrary times of decay after conclusionof reactor operation. The code accounts for branched decay chains. It also may account fordepletion of 235U and ingrowth of 239Pu, although those features were not invoked in thecalculations reported here because of minimal depletion in TRIGA fuel elements.B. Fission Product Inventory CalculationsWhen burnup for TRIGA fuel containing 8.5% uranium enriched to reaches about 6grams 235U, the element does not have enough net positive reactivity to contribute to criticality,and is removed from service. Since end of fuel element life is about 6 grams burnup, a 10 gramburnup is used in calculations to maximize potential fission product inventory.Page 13-7 CHAPTER 13, ACCIDENT ANALYIS 12/2011SCALE is a comprehensive modeling and simulation suite for nuclear safety analysis and designdeveloped and maintained by Oak Ridge National Laboratory under contract with NRC and DOEto perform reactor physics, criticality safety, radiation shielding, and spent fuel characterizationfor nuclear facilities and transportation/storage package designs. A SCALE depletion sequence(code input, Appendix 13.1) was used to generate inventories of radioactive fission products foroperation at steady state power until the target burnup was achieved. The sequence uses KENOVI to develop a reactor specific (geometry and materials) flux, with SCALE integratingcalculations of flux averaged cross sections by several modules in sequence accounting forvarious factors that influence interaction rate, such as resonance self-shielding. Core andreflector geometry used to model the core is described in Chapter 4. Flux average cross sectionsare then used by ORIGEN S to calculate fission product generation and depletion. ORIGENdefaults to 1 metric ton of heavy metal (i.e., uranium) for calculations; the default value wasused to simulate fission product inventory buildup with negligible burnup.ORIGEN ARP (code input, Appendix 13.2) was used to determine the fission product inventoryfollowing specified decay intervals based on the depletion code output data. Burnupcalculations were performed to evaluate (1) the core inventory for nominal 1 MW operations,(2) core inventory for nominal 2 MW power level, which may alternately be considered as thevalue for a single fuel element operating at twice the 1 MW core average power level, and (3)core inventory for 3.5 MW power level, or the a peak value for a 2 MW core average with apeaking factor of 1.5. The number of days to achieve each burnup was determined manual byiterations of the code to find an end point 235U mass for 1 MTU that is 75% of the original value.SCALE burnup calculations are limited to a specified number of days to limit errors in calculationfor large burnup values; the TRIGA burnup is not large, and the maximum number of days percalculation step was changed in the input for the TRIGA core, but still resulted in a differentnumber of steps for each burnup value. Parameters of the calculations are provided in Table13.6.Table 13.6, SCALE T-6 Sequence Continuous Burnup ParametersSTART ENDMW DAYS STEPS MASS MAS RATIOMASS MASS1 41000 104 1.97E+05 1.47E+05 7.46E-012 20475 54 1.97E+05 1.47E+00 7.46E-063.5 11625 20 1.97E+05 1.47E+05 7.48E-01While long lived radionuclides should reasonably be represented by continuous operations atthese intervals, the irradiation schedule is not representative of NETL operations. The facility isnot staffed for continuous operations, and radioisotopes that have half-lives on the order of afew hours to days are not well represented. Therefore, a schedule for 1 working week (5 days, 8hour operations at the specified power level followed by 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> of decay) was added to eachirradiation following the continuous burnup interval. Fission product inventories calculated byORIGEN are provided in Tables 13.7A and 13.7B for the specified gaseous and particulate fissionproducts.Page 13-8 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFET ANALYSIS REPORT, CHAPTER 131Table 13.7A, 1 MTU Gaseous Fission Product Inventory for 3.5 kW Case (Ci)s m h D1 30 1 8 1 7 30 90 180 365br80 br80m br82 br83 br84 i125 i128 i129 i130 i131 i132 i133 i134 i135 kr79 kr8l kr83m kr85 kr85m kr87 kr88 xe125 xe127 xe129m xe131m xe133 xe133m xe135 xe135m xe138 0Table 13.7B, 1 MTU Particulate Fission Product Inventory (Ci)s m h d1 30 1 8 1 7 30 90 180 365cs131 cs132 cs134m cs135 cs135m cs136 Page 13-9 CHAPTER 13, ACCIDENT ANALYISI12/2011Table 13.7B, 1 MTU Particulate Fission Product Inventory (Ci)s m h d1 30 1 8 1 7 30 90 180 365cs137 cs138 sr85 sr85m sr87m sr89 sr90 sr9l sr92 C. Fission Product releaseThe calculated fission product inventory for 1 MTU is scaled by the ratio of the mass of a fuelelement to a metric ton to determine the fission product inventory for a single fuel element.Most of the fission products generated during operation are trapped in the fuel matrix, only afraction of the inventory has enough mobility to escape. The fraction that escapes is calculatedusing release fractions provided by NUREG/CR-2387 applied to each radionuclide identified. TheNUREG considers noble gas, iodine, cesium, and strontium as the isotopes significant toconsequence analysis; other refractories are neglected as they do not contribute significantly topotential exposure. Release inventories are provided in Tables 13.8A and 13.8B for gaseous andparticulate fission products.Table 13.8A. Gaseous Fission product Release from Single Element (tICi)S M H D1 30 1 8 1 7 30 90 180 365br80br80mbr82br83br84i125 i128i129i130i131i132i133i134i135kr791.5E-49.OE-51.7E-13.1E25.5E21.1E-102.7E-11.4E-41.51.7E32.5E33.9E34.5E3 3.6E31.9E-l15.4E-34.3E-38.91.5E41.6E41.1E-101.2E-1 1.4E-41.51.7E32.5E3 3.9E34.1E33.5E31.9E-108.7E-57.7E-51.7E-12.7E21.7E211E-105.OE-21.4E-41.41.7E32.5E33.8E33.4E33.3E31.9E-101.2E-31.1E-37.51.4E32.4E-11.1E-108.3E-81.4E-49.1E-11.6E32.4E33.0E31.4E11.4E31.6E-101.9E-61.8E-61.1E-12.7E-13.9E-121.1E-102.3E-191.4E-43.7E-11.6E32.1E31.7E35.OE-52.6E21.2E-103.5E-163.3E-162.OE-11.2E-200.09.7E-11.0.01.4E-43.OE-58.6E24.5E26.40.05.2E-64.2E-120.00.07.7E-80.0 0.07.4E-110.01.4E-41.1E-181.2E23.16.6E-80.02.7E-317.7E-170.00.02.OE-180.0 0.03.7E-110.01.4E-40.06.6E-17.2E-69.5E-290.00.03.3E-290.00.0 0.00.0 0.01.3E-110.01.4E-40.02.8E-42.5E-140.00.00.00.00.00.00.00.00.01.5E-120.01.4E-40.03.lE-110.00.00.00.00.0kr8l 5.7E-11 5.7E-11 5.7E-11 5.7E-11 5.7E-11 5.7E-11 5.7E-11 5.7E-11 5.7E-11 5.7E-11Page 13-10 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 1 12/2011SAFETY ANALYSIS REPORT, CHAPTER 13kr83m 3.1E2 3.1E2 3.0E2 7.2E1 1.0 4.2E-7 3.5E-7 2.1E-7 1.OE-7 2.4E-8kr85 1.0E2 1.0E2 1.0E2 1.0E2 1.0E2 1.0E2 1.0E2 9.9E1 9.7E1 9.4E1kr85m 7.7E2 7.2E2 6.7E2 1.9E2 1.6E1 8.4E-11 0.0 0.0 0.0 0.0kr87 1.5E3 1.1E3 8.7E2 1.1E1 1.8E-3 0.0 0.0 0.0 0.0 0.0kr88 2.0E3 1.8E3 1.6E3 2.2E2 4.5 7.OE-18 0.0 0.0 0.0 0.0xe125 9.4E-13 9.2E-13 9.OE-13 6.5E-13 3.4E-13 3.4E-16 5.OE-26 0.0 0.0 0.0xe127 3.7E-8 3.7E-8 3.7E-8 3.7E-8 3.6E-8 3.2E-8 2.OE-8 6.5E-9 1.2E-9 3.5E-11xe129m 1.1E-5 1.1E-5 1.1E-5 1.1E-5 1.OE-5 5.9E-6 9.8E-7 9.OE-9 8.OE-12 4.3E-18xe131m 2.OE1 2.OE1 2.OE1 2.OE1 2.OE1 1.8E1 7.2 2.9E-1 1.5E-3 3.1E-8xe133 3.9E3 3.9E3 3.9E3 3.9E3 3.7E3 1.6E3 7.8E1 2.8E-2 1.9E-7 4.5E-18xe133m 4.2E1 4.2E1 4.2E1 4.1E1 3.8E1 5.4 3.7E-3 2.1E-11 9.OE-24 0.0xe135 3.6E3 3.6E3 3.6E3 2.9E3 1.3E3 5.7E-3 3.8E-21 0.0 0.0 0.0xe135m 4.7E2 3.8E2 3.4E2 1.4E2 2.7E1 5.4E-7 0.0 0.0 0.0 0.0xe138 3.6E3 8.3E2 1.9E2 1.0E-8 3.1E-29 0.0 0.0 0.0 0.0 0.0Table 13.8B. Particulate Fission Product Release from Single ElementS M H D1.0 3.OE1 1.0 8.0 1.0 7.0 3.0E1 9.OE1 1.8E2 3.7E2cs131 2.7E-10 2.7E-10 2.7E-10 2.7E-10 2.5E-10 1.5E-10 3.OE-11 4.1E-13 6.5E-16 1.2E-21cs132 1.0E-5 9.9E-6 9.9E-6 9.6E-6 8.9E-6 4.2E-6 3.6E-7 5.9E-10 3.9E-14 9.8E-23cs134m 3.3E-2 3.OE-2 2.6E-2 3.9E-3 8.7E-5 3.7E-22 0.0 0.0 0.0 0.0cs135 1.1E-4 1.1E-4 1.1E-4 1.1E-4 1.1E-4 1.1E-4 1.1E-4 1.1E-4 1.1E-4 1.1E-4cs135m 2.OE-3 1.4E-3 9.2E-4 1.7E-6 6.1E-12 0.0 0.0 0.0 0.0 0.0cs136 9.3E-2 9.3E-2 9.3E-2 9.1E-2 8.8E-2 6.1E-2 1.8E-2 7.7E-4 6.7E-6 3.9E-10cs137 7.3 7.3 7.3 7.3 7.3 7.3 7.3 7.3 7.2 7.2cs138 3.9E1 2.9E1 1.7E1 8.9E-4 2.OE-12 0.0 0.0 0.0 0.0 0.0sr85 1.4E-9 1.4E-9 1.4E-9 1.4E-9 -1.4E-9 1.3E-9 1.0E-9 5.4E-10 2.1E-10 2.9E-11sr85m 7.8E-10 5.7E-10 4.2E-10 3.1E-12 1.6E-16 0.0 0.0 0.0 0.0 0.0sr87m 2.2E-6 2.OE-6 1.7E-6 2.4E-7 4.7E-9 5.1E-27 0.0 0.0 0.0 0.0sr89 2.7EI 2.7EI 2.7E1 2.7E1 2.7E1 2.4E1 1.8EI 7.8 2.3 1.8E-1sr90 7.1 7.1 7.1 7.1 7.1 7.1 7.1 7.0 7.0 6.9sr9l 3.3EI 3.2E1 3.1E1 1.7EI 5.5 3.1E-5 1.7E-22 0.0 0.0 0.0sr92 3.4E1 3.OE1 2.6EI 3.3 5.1E-2 4.9E-21 0.0 0.0 0.0 0.0D. ALl Consequence AnalysisRegulatory Guideline 8.34, Monitoring Criteria and Methods To Calculate Occupational RadiationDoses, provides methodology to determine potential dose rates from ingestion of, or immersionin, radionuclides using data in 1OCFR20 Appendix B. The ALl is used to determine potentialconsequences from an ingestion of a radionuclide. If the radionuclide inventory is less than one1OCFR20 Appendix B "Annual Limit on Intake" (ALl), then it is not physically possible to exceedthe annual limits for worker exposure. If the available radionuclide release exceeds an ALl, thenit is necessary to examine the fraction of the inventory to which individuals will be exposed. Theratio of a radionuclide inventory to the ALl value determines the fraction of the limit subsumedPage 13-11 CHAPTER 13, ACCIDENT ANALYIS L 12/2011by a single radionuclide. The sum of the ratios for all radionuclides bounds the consequences,with a sum-value less than 1 indicating a total value less than the ALl value and a total valuegreater than 1 exceeding the ALl value.Table 13.9A, Fraction of Gaseous Fission Product Inventory to 10CFR20 ALls M h d1 30 1 8 1 7 30 90 180 365br8O 1.2E-5 2.1E-2 4.2E-2 3.3E-1 1.0 7.0 3.OE1 9.0E1 1.8E2 3.7E2br80m 7.5E-10 2.7E-8 4.4E-10 6.OE-9 9.5E-12 1.8E-21 0.0 0.0 0.0 0.0br82 4.5E-9 2.1E-7 3.8E-9 5.6E-8 8.9E-11 i.7E-20 0.0 0.0 0.0 0.0br83 4.4E-5 2.2E-3 4.3E-5 1.9E-3 2.7E-5 5.OE-5 1.9E-11 5.1E-22 0.0 0.0br84 5.2E-3 2.5E-1 4.4E-3 2.3E-2 4.5E-6 1.9E-25 0.0 0.0 0.0 0.0i125 9.2E-3 2.7E-1 2.8E-3 4.OE-6 6.4E-17 0.0 0.0 0.0 0.0 0.0i128 1.8E-12 1.8E-12 1.8E-12 1.8E-12 1.8E-12 1.6E-12 1.2E-12 6.1E-13 2.1E-13 2.5E-14i129 2.7E-6 1.2E-6 5.OE-7 8.3E-13 2.3E-24 0.0 0.0 0.0 0.0 0.0i130 1.6E-5 1.6E-5 1.6E-5 1.6E-5 1.6E-5 1.6E-5 1.6E-5 1.6E-5 1.6E-5 1.6E-5i131 2.1E-3 2.1E-3 2.OE-3 1.3E-3 5.3E-4 4.3E-8 1.5E-21 0.0 0.0 0.0i132 3.4E1 3.4E1 3.4E1 3.3E1 3.1E1 1.7E1 2.3 1.3E-2 5.5E-6 6.3E-13i133 3.1E-1 3.1E-1 3.1E-1 3.OE-1 2.6E-1 5.6E-2 3.9E-4 9.OE-10 3.1E-18 0.0i134 1.3E1 1.3E1 1.3E1 9.8 5.8 2.1E-2 2.2E-10 3.2E-31 0.0 0.0i135 9.OE-2 8.1E-2 6.7E-2 2.8E-4 1.OE-9 0.0 0.0 0.0 0.0 0.0br80 1.8 1.7 1.6 7.OE-1 1.3E-1 2.6E-9 1.3E-34 0.0 0.0 0.0SUMS: 48.8 48.6 48.4 43.6 37.2 17.2 2.4 0.13 2.1E-5 1.5E-5Table 13.91, Fraction of Particulate Fission Product Inventory to IOCFR20 ALls M h d1 30 1 8 1 7 30 90 180 365csl31 9.1E-15 9.1E-15 9.1E-15 8.9E-15 8.5E-15 5.1E-15 9.9E-16 1.4E-17 2.2E-20 3.9E-26cs132 2.5E-9 2.5E-9 2.5E-9 2.4E-9 2.2E-9 1.1E-9 9.OE-11 1.5E-13 9.7E-18 2.5E-26cs134m 3.3E-7 3.OE-7 2.6E-7 3.9E-8 8.7E-10 3.7E-27 0 0 0 0cs135 1.1E-7 1.1E-7 1.1E-7 1.1E-7 1.1E-7 1.1E-7 1.1E-7 1.1E-7 1.1E-7 1.1E-7cs135m I.OE-8 6.8E-9 4.6E-9 8.6E-12 3.OE-17 0 0 0 0 0cs136 1.3E-4 1.3E-4 1.3E-4 1.3E-4 1.3E-4 8.7E-5 2.6E-5 l.1E-6 9.6E-9 5.6E-13cs137 3.7E-2 3.7E-2 3.7E-2 3.7E-2 3.7E-2 3.7E-2 3.7E-2 3.6E-2 3.6E-2 3.6E-2cs138 6.5E-4 4.8E-4 2.9E-4 1.5E-8 3.3E-17 0 0 0 0 0sr85 7.2E-13 7.2E-13 7.2E-13 7.2E-13 7.1E-13 6.6E-13 5.2E-13 2.7E-13 1.OE-13 1.4E-14sr85m 1.3E-15 9.5E-16 7.OE-16 5.1E-18 2.7E-22 0 0 0 0 0sr87m 2.2E-11 2.OE-11 1.7E-11 2.4E-12 4.7E-14 5.1E-32 0 0 0 0sr89 2.7E-1 2.7E-1 2.7E-1 2.7E-1 2.7E-1 2.4E-1 1.8E-1 7.8E-2 2.3E-2 1.8E-3sr90 1.8 1.8 1.8 1.8 1.8 1.8 1.8 1.8 1.7 1.7sr9l 8.3E-3 8.OE-3 7.8E-3 4.4E-3 1.4E-3 7.7E-9 4.3E-26 0 0 0sr92 4.9E-3 4.3E-3 3.8E-3 4.7E-4 7.2E-6 7.OE-25 0 0 0 0SUMS: 2.1 2.1 2.1 2.1 2.1 2.0 2.0 1.9 1.8 1.8Page 13-12 THE UNIVERSITY OF TEXAS TRIGA 11 RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 13ALl values are exceeded for the 3.5 MW case; data from all cases is provided graphically in Fig.13.1. The gaseous radionuclide inventory is shown to be greater than the ALl for approximately25-40 days following the release, while the particulate radionuclide inventory remains above theALl for all cases, principally driven by 90Sr.Ratio, Radionuclide Inventory to ALl10' ... ... ............ .. .... .1 1 -0 4 ..... ........ .................. .... .... .. .. ... ... ... ..... .. ..:1 1.0. , .... ....... .................. ....... .. .............. .................... ...... ........... ..0 50 100 150 200 250 300 350-1 MW Bromine, Iodine -1] MIN Particulates -2 MW Bromnie, Iodine-2 MW Particulate -3.5 MW Bromnine and Iodine -3.5 MW ParticualtesDays Following MHAFigure 13.1, Ratio of Radionuclide Inventory to ALIThis analysis is extremely conservative in neglecting transport to personnel. There is noconceivable scenario where all r~adionuclide inventories are delivered to a single individual, andany reduction in the amount of uptake to the individual reduces the uptake.This analysis is conservative in assuming a burnup of 10 gra MS 235 U and a continuous operatinghistory. A slightly more realistic assumption of 6 grams burnup reduces the inventory of thelong lived 9°Sr by approximately 60%, and a less aggressive operating schedule reduces shorterlived radionuclides considerably.This analysis is conservative in assuming that the radionuclide inventory is maintained in onelocation for the duration of the analysis, and does not consider any removal of the inventoryfrom the receptor through normal atmospheric transport such as simple settling of particulatematter or removal from the reactor bay by the HVAC system, wind driven exchange of buildingair, or active cleanup processes. A reduction in inventory reduces the ALI ratio.Finally, this analysis does not consider any compensatory or mitigating actions in response to therelease. The reactor Radiation Protection Program requires monitoring and control of exposure,and with a maximum hypothetical ALI ratio of 1.8 for 9°Sr, measures to reduce and controlexposure to an individual by a factor of approximately 2 for particulate radionuclides are easilyPage 13-13 CHAPTER 13, ACCIDENT ANALYIS 12/2011achievable by passive measures or active processes such as dilution in the reactor bay air orfiltering.Therefore although the ALl values in the reactor bay are exceeded for the maximumhypothetical accident, control of personnel exposure under the Radiation Protection Program tothe radionuclides released into the reactor bay is adequate to manage personnel dose withinlimits of 1OCFR20.E. DAC Consequence AnalysisThe escaping fission product inventory is assumed to mix with reactor bay atmosphere. Nominalfree volume of the reactor bay is 4120 M3; 10% of the nominal volume is assumed occupied byequipment or materials. The radionuclide inventory is therefore assumed to be distributed in3719 M3.The 1OCFR20 "Derived Air Concentration" (DAC) is used to limit potentialconsequences for workers based on the radionuclide inventory released into a volume of air. Ina manner similar to the ALl analysis described above, consequences of exposure to a mixture ofradionuclides are evaluated based on the derived air concentration in 10CFR20 Appendix B withthe results in Table 13.10A and 13.10B.Table 13.10A, Fraction of Instantaneous Gaseous Fission Product Inventory to 10CFR20 DAC"1s m h d1 30 1 8 1 7 30 90 180 365br80 6.1E-10 2.2E-8 3.5E-1o 4.9E-9 7.7E-12 1.4E-21 0 0 0 0br80rn 4.8E-9 2.3E-7 4.1E-9 6.1E-8 9.6E-11 1.8E-20 0 0 0 0br82 2.8E-5 1.4E-3 2.8E-5 1.2E-3 1.7E-5 3.3E-5 1.2E-11 3.3E-22 0 0br83 3.4E-3 1.6E-1 2.9E-3 1.5E-2 2.9E-6 1.3E-25 0 0 0 0br84 8.9E-3 2.6E-1 2.7E-3 3.9E-6 6.2E-17 0 0 0 0 0i125 1.1E-12 1.1E-12 1.1E-12 1.1E-12 1.1E-12 1.OE-12 8.OE-13 4.OE-13 1.4E-13 1.6E-14i128 1.7E-6 7.5E-7 3.3E-7 5.4E-13 1.5E-24 0 0 0 0 0i129 1.1E-5 1.1E-5 1.1E-5 1.1E-5 1.1E-5 1.1E-5 1.1E-5 1.1E-5 1.1E-5 1.1E-5i130 1.6E-3 1.6E-3 1.5E73 9.8E-4 4.OE-4 3.2E-8 1.2E-21 0 0 0i131 2.7E1 2.7E1 2.7E1 2.7E1 2.5E1 1.4E1 1.90 1.1E-2 4.5E-6 5.1E-13i132 2.7E-1 2.7E-1 2.7E-1 2.6E-1 2.2E-1 4.9E-2 3.4E-4 7.8E-10 2.7E-18 0i133 1.3E1 1.2E1 1.2E1 9.60 5.60 2.1E-2 2.1E-10 3.1E-31 0 0i134 7.3E-2 6.6E-2 5.4E-2 2.3E-4 8.2E-10 0 0 0 0 0i135 1.70 1.60 1.50 6.5E-1 1.2E-1 2.4E-9 1.2E-34 0 0 0kr79 3.1E-15 3.1E-15 3.1E-15 2.6E-15 1.9E-15 6.9E-17 1.2E-21 5.3E-34 0 0kr8l 2.6E-17 2.6E-17 2.6E-17 2.6E-17 2.6E-17 2.6E-17 2.6E-17 2.6E-17 2.6E-17 2.6E-17kr83m 1.OE-5 1.OE-5 9.8E-6 2.3E-6 3.3E-8 1.4E-14 1.1E-14 7.OE-15 3.4E-15 7.6E-16kr85 3.2E-4 3.2E-4 3.2E-4 3.2E-4 3.2E-4 3.2E-4 3.2E-4 3.2E-4 3.1E-4 3.OE-4kr85m 1.2E-2 1.2E-2 1.1E-2 3.1E-3 2.6E-4 1.4E-15 0 0 0 0kr87 9.6E-2 7.4E-2 5.6E-2 7.2E-4 1.2E-7 0 0 0 0 0kr88 3.3E-1 2.9E-1 2.6E-1 3.6E-2 7.3E-4 1.1E-21 0 0 0 0xe125 1.5E-17 1.5E-17 1.5E-17 l.E-17 5.5E-18 5.6E-21 8.2E-31 0 0 0xe127 1.2E-12 1.2E-12 1.2E-12 1.2E-12 1.2E-12 1.OE-12 6.6E-13 2.1E-13 3.8E-14 1.1E-15Page 13-14 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 13I12/2011Table 13.10A, Fraction of Instantaneous Gaseous Fission Product Inventory to 10CFR20 DAC[11s m h d1 30 1 8 1 7 30 90 180 365xe129m 1.8E-11 1.8E-11 1.8E-11 1.7E-11 1.6E-11 9.5E-12 1.6E-12 1.5E-14 1.3E-17 7.OE-24xe131m 1.6E-5 1.6E-5 1.6E-5 1.6E-5 1.6E-5 1.4E-5 5.8E-6 2.3E-7 1.3E-9 2.5E-14xe133 1.3E-2 1.3E-2 1.3E-2 1.2E-2 1.2E-2 5.2E-3 2.5E-4 9.OE-8 6.1E-13 1.5E-23xe133m 14E-4 1.4E-4 1.4E-4 1.3E-4 1.2E-4 1.7E-5 1.2E-8 6.8E-17 2.9E-29 0xe135 1.2E-1 1.2E-1 1.2E-1 9.5E-2 4.1E-2 1.8E-7 1.2E-25 0 0 0xe135m 1.7E-2 1.4E-2 1.2E-2 5.2E-3 9.6E-4 1.9E-11 0 0 0 0xe138 2.9E-1 6.7E-2 1.5E-2 8.4E-13 2.5E-33 0 0 0 0 0SUMS: 3.6 3.5 3.5 3.1 2.6 1.2 1.6 9.1E-3 2.8E-4 2.6E-4Table 13.10B, Fraction of Instantaneous Particulate Fission Product Inventory to 10CFR20 DACs M h D1 30 1 8 1 7 30 90 180 365csl31 7.4E-15 7.4E-15 7.4E-15 7.2E-15 6.8E-15 4.1E-15 8.OE-16 1.1E-17 1.8E-20 3.2E-26cs132 1.3E-9 1.3E-9 1.3E-9 1.3E-9 1.2E-9 5.7E-10 4.9E-11 7.9E-14 5.2E-18 1.3E-26cs134m 1.5E-7 1.3E-7 1.2E-7 1-8E-8 3.9E-10 1.7E-27 0 0 0 0cs135 5.8E-8 5.8E-8 5.8E-8 5.8E-8 5.8E-8 5.8E-8 5.8E-8 5.8E-8 5.8E-8 5.8E-8cs135m 6.8E-9 4.6E-9 3.1E-9 5.8E-12 2.OE-17 0 0 0 0 0cs136 8.3E-5 8.3E-5 8.3E-5 8.2E-5 7.9E-5 5.5E-5 1.6E-5 6.9E-7 6.OE-9 3.5E-13cs137 3.3E-2 3.3E-2 3.3E-2 3.3E-2 3.3E-2 3.3E-2 3.3E-2 3.3E-2 3.3E-2 3.2E-2cs138 5.2E-4 3.9E-4 2.4E-4 1.2E-8 2.7E-17 0 0 0 0 0sr85 6.5E-13 6.5E-13 6.5E-13 6.4E-13 6.4E-13 5.9E-13 4.6E-13 2.4E-13 9.3E-14 1.3E-14sr85m 7.OE-16 5.1E-16 3.8E-16 2.8E-18 1.5E-22 0 0 0 0 0sr87m 1.2E-11 1.iE-1i 9.4E-12 1.3E-12 2.6E-14 2.8E-32 0 0 0 0sr89 1.2E-1 1.2E-1 1.2E-1 1.2E-1 1.2E-1 1.1E-1 8.OE-2 3.5E-2 1.OE-2 8.OE-4sr90 9.5E-1 9.5E-1 9.5E-1 9.5E-1 9.5E-1 9.5E-1 9.5E-1 9.5E-1 9.4E-1 9.3E-1sr9l 9.OE-3 8.7E-3 8.4E-3 4.7E-3 1.5E-3 8.3E-9 4.6E-26 0 0 0sr92 3.1E-3 2.7E-3 2.4E-3 2.9E-4 4.5E-6 4.4E-25 0 0 0 0SUMS: 1.12 1.12 1.12 1.11 1.11 1.09 1.06 1.02 0.985 0.963NOTE[1]: DAC limits are based on 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> of exposure over a year; these tables compare the instantaneous valueof airborne radionuclides, and not the 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> exposure period. Integration of the instantaneous valuesover a year evaluates compliance with DAC limits.The DAC values are exceeded for gaseous fission product concentration for about 40 daysfollowing the accident in the 3.5 MW case. Atmospheric particulate activity remains elevatedabove the DAC value because of the long-lived strontium (as with the ALl) for about Y2 of theyear in the 3.5 MW case; however, the particulate values for the 2 and the 1 MW cases do notexceed the DAC values at any time, as indicated in Fig. 13.2.Page 13-15 CHAPTER 13, ACCIDENT ANALYIS112/2011CHAPTER....13 ....ACCIDENT .. ANA..YIS.........Ratio. Radionuclide Inventory to DAC100...10ai.00.010.0010.0001... ...... ... .... .............. .......... ....... ...... ... .....0.00001 -_-_i :' _:.. .. ...... .. .... ........ ..---------"0 50 100 150 200 250 300 350-1 MWDAC NGI -1 MWDAC PART-2 MWDAC NGI-2 MW DAC PART-3.5 MW NGI -3.5 MW PARTDays Following MHAFigure 13.2, Ratio of Radionuclide Concentration to 10CFR 20 DAC ValuesDAC values apply to continuous exposure over a year. Concentrations averaged using timeinterval weighting over a year following the event for all three cases are provided in Table 13.11.Table 13.11, DAC Ratios for All CasesSeconds 1Minutes 30Hours 18Days 173090180365Weighted averageGaseous1010108.97.43.30.452.7E-031.4E-041.4E-040.111 MWParticulate0.690.690.690.690.690.690.680.660.650.640.672 MWGaseous Particulate20 0.8920 0.8920 0.8918 0.8915 0.896.6 0.880.91 0.865.3E-03 0.832.1E-04 0.812.OE-04 0.80.25 0.863.5 MWGaseous Particulate36 1.135 1.135 1.131 1.126 1.112 1.11.6 1.19.1E-03 1.02.8E-04 0.982.6E-04 0.960.46 1.10As in the ALl analysis, this analysis is conservative in assuming a burnup of 10 grams 235U and acontinuous operating history. A slightly more realistic assumption of 6 grams burnup reducesthe inventory of the long lived 90Sr by approximately 60%, interpolating between the particulateratios indicates that the DAC ratio is 1 at or below 3 MW, and a less aggressive operatingschedule reduces shorter lived radionuclides considerably.Page 13-16 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 13This analysis is conservative in assuming 10% of the volume is occupied by equipment.Increasing the volume decreases nuclide concentration.This analysis is conservative in assuming that the radionuclide inventory is maintained in onelocation for the duration of the analysis, and does not consider any removal of the inventoryfrom the reactor bay through normal atmospheric transport, either simple settling of particulatematter or removal from the reactor bay by natural or active cleanup processes. The reactor bayHVAC control system is designed to automatically secure ventilation on detecting a preset levelof airborne contamination, and there is some delay before the radionuclides buildup to the triplevel. During this interval the reactor bay continues to exhaust by design 34.3 m3 s-1, and anactual 57.2 m3 s-. A reduction in fission product inventory reduces the DAC ratio.Finally, this analysis does not consider any compensatory or mitigating actions in response to therelease. The reactor Radiation Protection Program requires monitoring and control of exposure,and with a maximum hypothetical DAC ratio of 1.14 for 90Sr that dominates the particulateanalysis, measures to reduce and control exposure to an individual by a factor of approximately2 for particulate radionuclides are easily achievable by passive measures or active processessuch as dilution in the reactor bay air or filtering.Therefore although the DAC values in the reactor bay are exceeded for the 2 and 3.5 MW case ofthe maximum hypothetical accident under extremely conservative assumptions, access controlto manage personnel exposure under the Radiation Protection Program is adequate to maintainpersonnel dose within limits of 1OCFR20.F. Effluent release Consequence AnalysisThe radionuclide concentration in the reactor bay atmosphere following the maximumhypothetical accident is compared to the effluent limit, assuming the radionuclide inventory isnot transported from confinement and is only removed through decay.Table 13.12, Reactor Bay Atmosphere Following MHA Compared to Effluent Limit1IMW 21MW 3.5MWTIME NGI PART NGI PART NGI PART1s 2.29E+02 2.29E+02 2.94E+02 2.94E+02 3.67E+02 3.67E+0230 m 2.29E+02 2.29E+02 2.94E+02 2.94E+02 3.67E+02 3.67E+02lh 2.29E+02 2.29E+02 2.93E+02 2.93E+02 3.67E+02 3.67E+028 h 2.29E+02 2.29E+02 2.92E+02 2.92E+02 3.65E+02 3.65E+02Id 2.28E+02 2.28E+02 2.92E+02 2.92E+02 3.64E+02 3.64E+027 d 2.27E+02 2.27E+02 2.90E+02 2.90E+02 3.60E+02 3.60E+0230d 2.24E+02 2.24E+02 2.84E+02 2.84E+02 3.51E+02 3.51E+0290 d 2.20E+02 2.20E+02 2.75E+02 2.75E+02 3.36E+02 3.36E+02180d 2.16E+02 2.16E+02 2.70E+02 2.70E+02 3.27E+02 3.27E+02365 d 2.13E+02 2.13E+02 2.65E+02 2.65E+02 3.20E+02 3.20E+02Page 13-17 CHAPTER 13, ACCIDENT ANALYIS 12/2011The results demonstrate that the reactor bay atmosphere cannot be discharged to theenvironment in the absence of mitigating factors. However, individuals are not directly exposedto effluent releases. The environment dilutes the radionuclide concentration in atmosphericdispersion.F (1) Atmospheric DispersionStandard plume modeling is used to assess dilution of contaminants at the exit of the reactorbay. A standard approach assumes a Gaussian distribution for the dispersion of contaminantsperpendicular to wind-driven motion of material in a plume. The Gaussian dispersionparameters are a function of atmospheric stability and the distance of plume travel. TheWorkbook of Atmospheric Dispersion Estimates (D. B. Turner, 2nd Ed., 1994) reports dispersionparameters determined experimentally for urban areas, tabulated below.Table 13.13: BRIGGS URBAN DISPERSION PARAMETERSo, (meters) ac (meters)x,km A-B C D E-F A-B C D E-F0.01 3.19 2.20 1.60 1.10 2.41 2.00 1.40 ' 0.790.02 6.37 4.38 3.19 2.19 4.85 4.00 2.79 1.580.03 9.54 6.56 4.77 3.28 7.31 6.00 4.18 2.350.04 12.70 8.73 6.35 4.37 9.79 8.00 5.57 3.110.05 15.80 10.90 7.92 5.45 12.30 10.00 6.95 3.860.06 19.00 13.00 9.49 6.52 14.80 12.00 8.33 4.600.07 22.10 15.20 11.00 7.59 17.40 14.00 9.70 5.330.08 25.20 17.30 12.60 8.66 20.00 16.00 11.10 6.050.09 28.30 19.50 14.10 9.73 22.60 18.00 12.40 6.760.10 31.40 21.60 15.70 10.80 25.20 20.00 13.80 7.46Dispersion parameters are used to develop a conversion factor (X/Q) at each distance from thecenter of the release relating a contamination release rate (Qo, contaminant released persecond) to a concentration (Nj) at the specified location.QxThe release of the radioactive inventory (Qo) can be characterized as the product of the nuclideconcentration being released (NO) and the volumetric flow rate (1i volume per second), and theproduct of the nuclide and the decay constant (A) is the activity of the radionuclide. Therefore,where *i, is the volumetric flow rate (in units consistent the nuclide concentration) the equationcan be written as:Ao .i,=A,Page 13-18 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 13 aTwo cases are considered. The reactor bay ventilation is automatically secured on detection ofairborne radioactive material in the reactor bay, but the auxiliary purge system override is usedto re-initiate purge flow. Therefore the first case considers that the auxiliary purge systemdischarges the reactor bay effluent continuously through a HEPA filter and the building stack atthe normal flow rate (0.52 m3 per second). In the second case, the auxiliary purge systemremains secured trough the event. In the second case the discharge is not through the stack,but through normal building aspiration processes as a result of environment (wind) drivendifferential pressures developed across the building acting on vents and other buildingpenetrations.F (2) CASE I:Regulatory Guide 1.145, Atmospheric Dispersion Models for Potential Accident ConsequenceAssessments at Nuclear Power Plants, addresses releases from stacks. The equation for ground-level relative concentration at the plume centerline for stack releases is given as:1 e- 2.-heX 2-a _ _ _ _ _ _Where:U,, is the wind speed applicable to the release heighthe is the effective stack heightay and oy are Gaussian dispersion coefficients for distance and height of the releaseFor the case of auxiliary purge system operation,0 Effective stack height is calculated in Chapter 9 as 1.71/{wind velocity} m above the topof the stack at 63 feet (19.202 m);0 A high efficiency particulate filter is required in operation of the auxiliary purge system,and is credited in analysis;0 The auxiliary purge system operates at a nominal 1100 cfm (0.52 m3/s, 5.2E5 cm3/s); thisflow rate is used in the dilution calculation;0 Removal rate for air from the reactor bay in days is calculated:cm3 3600s 24h0.52S 1h Id4120m.3 1E6cm3m3For a limiting case, and the wind speed is assumed to be 1, and the X/Q was calculated for eachassociated set of dispersion parameters (cr, a,), with the results provided in Table 13.14; themaximum X/OQ value that provides the least dispersion is 0.001416 (Class C, 0.02 km).Page 13-19 CHAPTER 13, ACCIDENT ANALYISI12/2011CHAPTER 13, ACCIDENT ANALYIS 12/2011Table 13.14, Calculated x/Q Valueskm0.010.020.030.040.050.060.070.080.090.1A-B0.0005410.0011930.0010920.000880.00070.0005580.0004540.0003740.0003130.000266C0.0003880.0013310.0014160.0012330.0010260.0008540.0007090.0005980.0005070.000437DE-F8.11E-050.0008430.0013090.0013770.0012850.0011480.0010150.0008870.0007830.0006896.54E-070.0001230.0004830.0008120.0010080.0010930.0011060.0010790.001030.000973Effluent limit values are exceeded by gaseous fission products for about 1 day. However, theeffluent limit values are bases on a continuous discharge over a year, and the total annualaverage is well within limits.Table 13.15, Reactor Bay Atmosphere Following MHA Compared to Effluent Limit1MW 2MW 3.5MWTIME NGI PART NGI PART NGI PARTIs 1.19 5.07E-05 1.53 6.49E-05 2.71 8.11E-0530 m 1.16 5.06E-05 1.53 6.48E-05 2.66 8.10E-051 h 1.13 5.06E-05 1.52 6.48E-05 2.63 8.10E-058 h 0.859 5.05E-05 1.33 6.46E-05 2.30 8.06E-051 d 0.633 5.05E-05 1.11 6.45E-05 1.92 8.04E-057 d 0.258 5.02E-05 0.493 6.40E-05 0.856 7.96E-0530 d 3.34E-02 4.95E-05 6.68E-02 6.28E-05 0.117 7.75E-0590 d 2.01E-04 4.85E-05 3.95E-04 6.08E-05 6.80E-04 7.42E-05180 d 1.46E-05 4.77E-05 2.16E-05 5.94E-05 2.85E-05 7.20E-05365 d 1.41E-05 4.68E-05 2.07E-05 5.82E-05 2.73E-05 7.04E-05In all cases where the auxiliary purge system is operating and the confinement ventilationsystem is secured, 1OCFR20 effluent limits are met in the maximum hypothetical accident.F (2) CASE II:Regulatory Guide 1.145, Atmospheric Dispersion Models for Potential Accident ConsequenceAssessments at Nuclear Power Plants, addresses releases through vents or other buildingpenetrations. REGGUIDE 1.145, section 1.3.1, considers three different effects for decreasingconcentrations of an effluent release from vents or building openings: fundamental atmosphericdispersion, effects of the building itself on atmospheric mixing characteristics, and the effects ofPage 13-20 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 13the building and plume meandering. The REGGUIDE provides three different formulae todetermine relative concentration values, and directs the use of the highest calculated value ofthe first two formulae (building effects and basic atmospheric dispersion); the third formulaaddresses mitigation (i.e., reduced concentration of contaminants) caused by turbulent mixingfrom building wake effects and plume meandering. Equation (3) applies to neutral and stableatmospheric conditions (Class D, E, F, G), where the 10 meter wind speeds are slow enough thatthe effects are significant (less than 6 m s-1).1U10 * (IT a-, o'.v + '7 =U,10 -.2 .o'- , CPlume meander and building wake effects (mixing effects) from the building at distances lessthan 800 meters from the release use a coefficient, ly in equation (3), which is the product of acorrection factor (M) and the dispersion coefficient, o,. The correction factor is presentedgraphically in the REGGUIDE for stability classes D, E, F, and G; each class has a constant valuefrom the minimum wind speed to 2 m s1, and decreases linearly from the maximum value at 2 ms-1 to 0 at 6 m sl. REGGUIDE guidance effectively allows the use of calculated plume meanderfactors (M) greater than 3, where winds are less than 6 m s-1.Or (for winds < 6 m s-1, where M >3):lM1Table 13.16, Calculated Plume Meander Factor (M) for < 6 m s1 WindsClass 0.77 m s1 2.57 m s-1 4.37 m slD 2 1.8575 1.4075E 3 2.715 1.815F 4 3.5725 2.2225G 6 5.2875 3.0375The minimum 10 meter dispersion parameters in Table 13.17 and the lowest correction factor(M) for the applicable category are provided in Table 13.16. The X/Q for each stability classcalculated for each equation in REGUIDE 1.45, using the minimum values for oy, Oz, and M, arereported in Table 13.18.Page 13-21 CHAPTER 13, ACCIDENT ANALYIS 12/2011Table 13.17, Minimum Dispersion Parameters byStability ClassA-B C D E-FGy 3.19 2.2 1.6 1.1oz 2.41 2 1.4 0.79M 2 3Table 13.18, Minimum x/Q by Stability ClassA-B C D E-FRG 1.45 (1) 0.004164 0.004351 0.004484 0.004572RG 1.45 (2) 0.013801 0.024114 0.047368 0.122098RG 1.45 (3) na na 0.071051 0.122098The limiting value for x/Q is 0.122.F (3) Source Term Release RateAs described in Chapter 9, the reactor bay ventilation system is designed to provide at least 2 airchanges per hour (2.29 m3 s1, 2.29E6 cm3 s-1), and actually produces about 5 air changes perhour. Also described, a control system secures ventilation when the atmospheric contaminationthe reactor bay reaches a fraction of a DAC. Effluent is then driven by building leakage. Buildingleakage with the HVAC system secured is driven by pressure differential across porous barriersdeveloped by winds, with the low pressure developed from a building wake. The ASHRAEHandbook of Fundamentals (2009) suggests a simple model for building leakage (SAE units) is:Q =2160. A .-TWhereQ is building leakage in cfm2160 is a conversion factorA is the "net open crack area of the room"AP is differential pressure between the room and the surrounding environmentAll door openings to the reactor bay are facing the same direction, all doors exit to buffer areasdesigned to support differential pressures. There is essentially no potential for differentialpressure at the reactor bay openings from environmental conditions.The equipment hatch has two hinged doors, 66 in. X 132.5 in. sealed in the center. All threepersonnel doors are 36 in. X 72 in. Therefore the total linear perimeter of all openings with theexception of the HVAC system is 1177 in. The HVAC system ductwork (separate supply andreturn) is approximately 2 ft. X 3 ft. at the inlet, 3 ft. XC 3 ft. at the outlet. The perimeter of theduct work is therefore 264 in. If a large % in. gap is assumed at all openings, the open surfacePage 13-22 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 13 -Aarea for all doors is 1.02 ft2.The duct surface area with the HVAC system operating is 15.00 ft2,with the HVAC system secured is 0.22 ft2.Total area open for air flow with the system operating is the sum of the HVAC surface area andthe door fittings, 16.02 ft2.Total open area with the system secured is the leakage around thedampers, 0.22 ft2.The ratio of surface areas in the two conditions is 1.37%. It is assumed thatthe pressure drop across the HVAC fan is at least a factor of 10 greater (than a pressuredifference from ambient conditions) when the fan is operating. The ratio of flow rate while theHVAC system is operating can reasonably be expected to be reduced by a factor consisting of theratio of the two specified areas and the square root of the ratio of the differential pressuresdriving flow, or reduction by a factor of 0.434%.Leakage around system dampers and door-penetrations is not measured, but all openings areequipped with gaskets. While the ventilation system is a confinement system and notcontainment, reactor bay openings are equipped with rubber seals which are periodicallychecked for function. Therefore the simple model is likely to overestimate building leakage by alarge margin.The most limiting atmospheric dispersion factor (0.122098) and the conservative estimate of thebuilding leakage factor (0.004343) provide a reduction in the airborne concentration of fissionproducts as indicated in Table 13.10A and 13.10B as they are released (from the reactor bay) bya factor of 5.30E-4.The concentration of the reactor bay fission product inventory (reduced by the minimumatmospheric dispersion in transit to unrestricted areas) is compared to the effluent limit, withthe results provided in Table 13.19.Table 13.19, Effluent Limit Ratio to Release Concentrations1 MW 2 MWGaseous Particulate Gaseous Particulate3.5 MWGaseous ParticulateSeconds 1 0.5594 0.1214 1.1183 0.1559 1.9546 0.1946Minutes 30 0.5488 0.1214 1.0977 0.1559 1.9191 0.1946Hours 1 0.5425 0.1214 1.0849 0.1554 1.8963 0.19468 0.4741 0.1214 0.9481 0.1548 1.6576 0.1935Days 1 0.3956 0.1209 0.7912 0.1548 1.3835 0.19307 0.1766 0.1204 0.3526 0.1538 0.6167 0.190930 0.0239 0.1188 0.0483 0.1506 0.0838 0.186190 1.45E-04 0.1167 2.84E-04 0.1458 4.91E-04 0.1782180 1.06E-05 0.1145 1.54E-05 0.1432 2.07E-05 0.1734365 1.01E-05 0.1129 1.48E-05 0.1405 1.96E-05 0.1697Weighted average 0.0056 0.1145 0.0112 0.1429 0.0196 0.1734Page 13-23 CHAPTER 13, ACCIDENT ANALYIS 12/2011In all cases for the maximum hypothetical accident, when the HVAC system is secured theannual effluent concentration limit is met.The most significant conservatism in this analysis is the assumption that meteorologicalconditions maintain the lowest possible dilution factor for a year. This is obviously notsupported in reality; any changes to meteorological conditions will increase dilution and reducethe concentration of the effluent.As in the ALl analysis, this analysis is conservative in assuming a burnup of 10 grams 235U and acontinuous operating history. A slightly more realistic assumption of 6 grams burnup reducesthe inventory of the long lived 90Sr by approximately 60%, interpolating between the particulateratios indicates that the DAC ratio is 1 at or below 3 MW, and a less aggressive operatingschedule reduces shorter lived radionuclides considerably.This analysis is conservative in assuming less than 100% dilution in the reactor bay volume, with10% of the volume is occupied by equipment. Increasing the volume decreases nuclideconcentration.This analysis is conservative in assuming that the radionuclide inventory is not decreased in thetransport from the reactor bay to the environment.The reactor bay HVAC control system is designed to automatically secure ventilation ondetecting a preset level of airborne contamination, and there is some delay before theradionuclides buildup to the trip level. During this interval the reactor bay continues to exhaustby design 34.3 m3 s-1, and an actual 57.2 m3 s-1. A reduction in reactor bay inventory reduces theradionuclide inventory to be released.The 1% release value is unrealistically conservative. Ambient flow from a building is motivatedby a pressure difference that is the result of wind. All reactor bay penetrations are on the southside of the building, and there are no access points on the east, west or north sides. Thereforeany wind developed pressure will be constant across the openings, and there is no differentialpressure to develop flow.Therefore although the DAC values in the reactor bay are exceeded for the 2 and 3.5 kW case ofthe maximum hypothetical accident under extremely conservative assumptions, dilution ofradionuclides at the point of release ensures 10CFR20 Appendix B effluent limits are met.13.3.3 Results and ConclusionsAlthough the initial radionuclide release inventory exceeds the Annual Limit on Intake, there isno conceivable means of delivering the total inventory to a single worker. Although theradionuclide inventory in the reactor bay exceeds the limiting DAC values in the limiting case,Page 13-24 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 13access control as required under the Radiation Protection Program would prevent exposure ofany individual exceeding 10CFR20 limits. Effluent limits are met.13.4 Insertion of Excess ReactivityRapid compensation of a reactivity insertion is the distinguishing design feature of the TRIGAreactor. Characteristics of a slow (ramp) reactivity insertion are less severe than a rapid transientsince temperature feedback will occur rapidly enough to limit the maximum power achievedduring the transient. Analyses of plausible accident scenarios reveal no challenges to safetylimits for the TRIGA. The fuel-integrity safety limit, according to Simnad (1980), may be stated asfollows:Fuel-moderator temperature is the basic limit of TRIGA reactor operation. Thislimit stems from the out-gassing of hydrogen from the ZrHx and the subsequentstress produced in the fuel element clad material. The strength of the clad as afunction of temperature can set the upper limit on the fuel temperature. A fueltemperature safety limit of 1150 TC for pulsing, stainless steel U-ZrH1.6s ... fuel isused as a design value to preclude the loss of clad integrity when the cladtemperature is below 500TC. When clad temperatures can equal the fueltemperature, the fuel temperature limit is 950 'C. ....Two reactivity accident scenarios are presented. The first is the insertion of 2.8% Ak/k ($4.00)reactivity at zero power (i.e., less than 1 kW) by sudden removal of a control rod. The second isthe sudden removal of the same 2.8% Ak/k ($4.00) reactivity with the core operating at themaximum power level permitted by the balance of core excess reactivity. Maximum excessreactivity permitted is 4.9% Ak/k ($7.00); if 2.8% Ak/k ($4.00) is allocated to the reactivitytransient, then 2.1% Ak/k ($3.00) supports power operation; 880 kW operation corresponds to a2.1% Ak/k ($3.00) reactivity deficit. Analysis shows that in neither scenario does the peak fueltemperature exceed the temperature limit. The nearest approach occurs if the reactor isoperating at a steady power of 880 kW, an action prevented both by administrativerequirements and by interlocks, but there is adequate margin to the temperature limit forcladding that has a temperature less than 500°C. Chapter 5 shows that steady statetemperatures are much less than 500°C.13.4.1. Initial Conditions, Assumptions, and ApproximationsThe following conditions establish an extremely conservative scenario for analysis of insertion ofexcess reactivity:For the first scenario, the reactor is critical below 1 kW, with reactor and coolant ambient(zero power) temperature 270C.Page 13-25 CHAPTER 13, ACCIDENT ANALYIS 12/2011* For the second scenario, the reactor is operating at a steady state power level supportedby core excess reactivity minus 2.8% Ak/k ($4.00) reserved for pulsing.* Maximum pulsed reactivity insertion is 2.8% ($4.00)0 The time over which heat is generated and causes fuel temperature to rise is muchshorter than heat transfer time constants for removal of the heat, so that analysis usesadiabatic conditions* The reactivity addition is assumed to be instantaneous and greater than the delayedneutron fraction so that the contribution of delayed neutrons is small and thereforeneglected in analysis0 A Control rod interlock preventing pulsing operations from power levels greater than amaximum of 10 kW is not credited0 Operation at 880 kW with the pulse rod decoupled and fully inserted is assumed to bepossible0 The core is assumed to contain 85 fuel elements to maximize the power generated in thehottest fuel elementHot channel factors calculated in Chapter 5 are used13.4.2 Computational Model for Power ExcursionsAs noted in Chapter 4, TRIGA fuel has a strong negative temperature coefficient. Operating atpower causes fuel to heat up, and the increase in temperature then contributes negativereactivity. The temperature increase is nearly instantaneous as fission products transfer kineticenergy to the fuel matrix, increasing the average kinetic energy/temperature compared to theheat transfer time constants for fuel and coolant. Rapid changes in core reactivity are thereforenearly adiabatic until the system has time to respond. Large, nearly instantaneous reactivityadditions, pulses, in a TRGIA reactor are therefore characterized by a power excursion, fuelheatup, and power reduction (associated with the heatup) over a short time interval. Thecontribution of delayed neutrons is limited by the transient time interval.Temperature is related to energy (power over an interval) through the specific heat capacity.Specific heat capacity of the fuel material (J/kg°K) at temperature T (TC) is given by:Cpf = C+ + C1

• TWhere c, is 340.1 and cl is 0.6952. With there are N fuel elements in the core, each with massmf(kg), then the core heat capacity (a, units of J/*K) is given byPage 13-26 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 13C=-[N in1](c0 + C! T) =C0 + C, TThe following relationships are for the Fuchs-Nordheim model, modified by Scalletar, for powerexcursions, as described for the TRIGA reactor by West et al. (1967). The inhour equation for theprompt critical condition (if the delayed neutron time constants do not play a significant role intime dependent behavior) reduces to:dP(t) p-/fldt)- p-8.P(t)dt tWhere t is the prompt neutron lifetime (41 Its for TRIGA II reactors). The rise in temperature isrelated to the change in reactor power from some initial power level (P(to)) to a power level atsome incremental time after the initial power (P(t)):C.- '5 = (P(t +.5t) -PQto)).dThe reactivity from fuel temperature (Ak/k) is characterized by the temperature coefficient (a)and the temperature (T):Ak = a.TLeading to:dP -(5k-aT).(Co +C, .) T)d -= f .(p -po)On integration (where T=O at P=Po):t- P -o)-o-n -T. 5k.Cý-a C -,,k -2 -,*-(0C 2C 3ck .-Minimum and maximum temperatures occur at pulse initiation and after completion. For P=P0,T.(k.C0o-(a.C0o-C,. 8k) T_ -C1 TC02 3Which simplifies to:Page 13-27 CHAPTER 13, ACCIDENT ANALYISI12/2011T'+ -_Skl3.T-3. CO=LC a J2 a*C1With the quadratic equation solution:2C, a 2 C, 2+4.3C1Only the positive half of the solution has physical meaning. Although there are nonlinear termsin the model, calculation of temperature change as a function of temperature shows a nearlylinear response.Pulsed Reactivity Versus Fuel Temperature.. .......... ... .. ... .. ....... ....... ..... .. .... .......................... T..... .. ...... .. .. ....... .... .................... .. ........ ......... ... ............. ..T...................... ....................... ... ...................i .................. .... .............. ........ .. ............... .................... ..i ............. ..... .. .......... .... ... .... ............ .......... ..._w ....... ................... ............................................... ....... ...... .............. ... ............. i..... ..... .... ...... .. ...... ...... ........ ..... ..... .. ..........aI-.........0 ..... .0 ......3 .-.0 .0 .4 ... .0 ....... ....0. ...00...Pulsed Reactivity. .. .Figure 13.3, Fuel Temperature and Pulsed ReactivityThe average core temperature rise in response to a reactor pulse was calculated and tabulatedin Table 13.20, along with the maximum temperature rise based on a radial peaking factor of n/2as provided in Chapter 4. Maximum permissible pool temperature during reactor operations is48.9°C; fuel temperatures that result from pulsed operations at the limiting pool temperatureare provided in Table 13.20 based on the increase in fuel temperature caused by the pulse overpool temperature.Table 13.20, Low Power Pulsed Reactivity ResponseRho $ ATFueI,Ave ATFuelPeak TFueiPpeak0.007000.007860.008821.0001.1231.260123137153193216241242264290Page 13-28 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 13Table 13.20, Low Power Pulsed Reactivity ResponseRho $ ATFuel,Ave ATFuel,Peak TFuel,Peak0.00990 1.414 171 269 3180.01111 1.588 192 301 3500.01247 1.782 214 336 3850.01400 2.000 239 375 4240.01572 2.245 267 419 4680.01764 2.521 298 468 5160.01981 2.829 332 522 5700.02223 3.176 370 582 6310.02495 3.565 413 649 6970.02801 4.002 460 723 7720.03144 4.492 513 806 8550.03529 5.042 571 898 9470.03962 5.660 637 1000 10490.04447 6.353 709 1114 1163At the maximum permissible pulsed reactivity insertion of 2.8% Ak/k, peak fuel temperature is772°C, approximately 18.7% below the safety limit of 950'C with cladding temperature above500°C and 32.9% below the safety limit of 1150'C with cladding temperature below 5000C. Itshould be noted that in pulsing from low power operations, the cladding temperature isdetermined by the pool water temperature so that the 1150°C safety limit applies.The second case assumes 2.8% Ak/k ($4.00) of reactivity is reserved for a pulse, and the reactoris operating at the maximum steady state power level that can be supported by the balance ofthe excess reactivity (2.1% Ak/k, $3.00). Power level is conservatively assumed to be 880 kW.Assuming an 85 element core (initial criticality for the UT TRIGA) and a n/2 peaking factor, thehottest element produces 17.3 kW. As noted in Chapter 4, the temperature difference acrossthe fuel matrix is calculated by:AT=q rk2.-kfWhere q" is the heat flux across the outer cladding surface, kf is the fuel conductivity (18 W m-1K1) and r, is the fuel diameter. The temperature difference from fuel center to the outer surfaceof the element is 195"C. The temperature difference for 17.3 kW from bulk water to the innersurface of the cladding was calculated in Chapter 4 to be 120C. Fuel temperature is therefore2070C. Using data previously calculated for temperature rise from pulse reactivity values, thepeak fuel temperatures were calculated for pulsed reactivity values from $1.00 to approximately$4.5 and reported in Table 13.21.Page 13-29 CHAPTER 13, ACCIDENT ANALYIS112/2011CHAPTER 13, ACCIDENT ANALYIS 12/ 2011Table 13.21, Initial Power 880 kW Pulsed Reactivity Responseak/k/k0.007000.007860.008820.009900.011110.012470.014000.015720.017640.019810.021000.022230.024950.028010.031441.0001.1231.2601.4141.5881.7822.0002.2452.5212.8293.0003.1763.5654.0024.492ATFueI,Ave123137153171192214239267298332351370413460513ATFueI,Peak193216241269301336375419468522551582649723806TFuel,Peak4004234484765085435826266757297587898569301013Pulsing to $3.00 from 880 kW, the hot channel has a margin of 20.1 below the safety limit of950'C with cladding temperature above 5000C and 34.0% below the safety limit of 11500C withcladding temperature below 5000C. Pulsing from $4.00 the margin to the safety limit of 950°Cwith cladding temperature above 5000C is only 2%, but the margin to the safety limit of 1150°Cwith cladding temperature below 500°C is 19.1%. Chapter 4 shows that for steady stateoperations the cladding temperature is below the fuel temperature, well below 500°C, so thatthe 1150°C limit and the 19.1% margin applies.The postulated scenarios do not result in fuel damage, but physical aspects of system preventthese scenarios from occurring. It is not possible to achieve full power operation with the pulserod fully inserted; since the pulse rod is partially withdrawn with air applied to the pulsesolenoid, it physically cannot be pulsed. Although not required to ensure the safety of thereactor, an interlock prevents pulsing from power levels greater than a maximum of 10 kW.13.4.3 Results and ConclusionsInsertion of the maximum possible reactivity of $4.00 without initial temperature feedback (i.e.,fuel temperature is too low to limit core available reactivity) results in a peak hot spot wellbelow the safety limit. Insertion of the $4.00 maximum possible reactivity with the reactoroperating at power providing initial temperature feedback results in a peak hot spot fueltemperature well below the safety limit for cladding temperature greater than 500'C.Page 13-30 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 1313.5 Loss of Reactor Coolant AccidentAlthough total loss of reactor pool water is considered to be an extremely improbable event,calculations have been made to determine (1) the maximum fuel temperature rise and (2) themaximum radiation dose that could be expected to result from such an event taking place afterlong-term operation at power levels up to 2 MW.A TRIGA fuel element (with 8.5% uranium) useful life ends about 6 grams of burnup; a 10 gramburnup is used for the end of life as a conservative assumption. Slightly more than 1 gram isdepleted per MWD of burnup, corresponding to 10 MWD per element. Calculations wereperformed using the T-6 depletion sequence SCALE to determine decay heat at shutdown fromTRIGA fuel operated to 10 MWD per element (input file in Appendix I). TRIGA specific crosssection libraries were generated as part of the sequence for use in determining the gammasource term from fission product decay. Scale is a comprehensive modeling and simulation suitefor nuclear safety analysis and design developed and maintained by Oak Ridge NationalLaboratory under contract with NRC and DOE to perform reactor physics, criticality safety,radiation shielding, and spent fuel characterization for nuclear facilities andtransportation/storage package designs. ORIGIN ARP was used to determine time dependentdecay heat and gamma energy-spectrum intensity. Fuel temperature is calculated using thedecay heat and first principles modeling of cooling. Radiation dose rates for receptor locationsare modeled using MCNP with the gamma spectrum as a source term.Discharge flow rate from a tank at atmospheric pressure (Streeter, V. L., E. B. Wylie, and K. W.Bedford, 1998, Fluid Mechanics. McGraw-Hill, Inc. 9ed, Daugherty, R. L., J. B. Franzinin, and E. J.Finnemore. 1985; Fluid Mechanics with Engineering Applications. Mc Grawe Hill, Inc. 8ed) isgiven by:Q a a.C. 2. g.hWhere:a is the diameter of a circular (drain) openingC is the loss coefficient associated with the openingh is the water height, subscripted i for initial and f for finalg is the acceleration of gravityFlow from a tank with a constant cross sectional area A is also characterized by:dh0=-A.-AdtPage 13-31 CHAPTER 13, ACCIDENT ANALYIS12/2011The time to drain a tank open to atmosphere between an initial level (Hi) and a final level (Hf) iscalculated by substituting the differential into the first equation and integrating between theinitial and final heights, with the result:a.CAs described in Chapter 4, the pool has a composite surface are of a circle with radius of 39 in.(0.9906 m) and a 39 in. X 78 in. (0.9906 m X 1.9626 m) rectangle. Normal pool height is 8.1 m,with a reactor scram at 7.8 m. The loss coefficient is a dimensionless number between 0 and1.0, with high turbulence as in a sharp edge losing more (61%) than from a short tube (80%).Since the discharge from the pool through a beam port travels through about 3 m with multipleabrupt changes in diameter, significant additional loss can be expected; for conservatism, a lossfactor of 0.61 is assumed. If a beam port shears and falls completely out of the flow path whilethe beam port shutter is open and no shielding or obstructions to flow are in the beam line, aminimum of 5.0 minutes will be required to drain the pool coolant from 7.8 m to the top of theactive fuel (47.25 in, 1.200 m above the pool floor). Therefore cooling analysis assumes a decaytime of 5 minutes prior to uncovering fuel. Reduced shielding capability occurs as the waterfalls, but normal levels are adequate for full power operations and most of the radiationexposure source term during operation is from fissions, falling by a factor of about 0.053 atshutdown. Since (1) shielding requirements are significantly reduced and (2) the calculation ofthe time to drain the pool to the top of the reactor core, a 5 minute decay time is assumed forsource term calculation.This section demonstrates under extraordinarily conservative assumptions that maximum fueltemperature reached in a loss of coolant accident is well below any safety limit for TRIGA reactorfuel. Conservatism notwithstanding, the margin between computed temperature and designlimits is sufficiently great to accommodate a design margin of at least a factor of two. Limitingdesign basis parameters and values for cooling consideration are addressed by Simnad (1980) asfollows:Fuel-moderator temperature is the basic limit of TRIGA reactor operation. This limitstems from the out-gassing of hydrogen from the ZrH, and the subsequent stressproduced in the fuel element clad material. The strength of the clad as a function oftemperature can set the upper limit on the fuel temperature. A fuel temperature safetylimit of 1150 °C for pulsing, stainless steel U-ZrH1.65 ... fuel is used as a design value topreclude the loss of clad integrity when the clad temperature is below 500 9C. When cladtemperatures can equal the fuel temperature, the fuel temperature limit is 950 'C. Thereis also a steady-state operational fuel temperature design limit of 750'C based onconsideration of irradiation- and fission-product-induced fuel growth and deformation....Page 13-32 THE UNIVERSITY OF TEXAS TRIGA 11 RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 1313.5.1 Initial Conditions, Assumptions, and ApproximationsThe following conditions establish the scenario for analysis of the loss of coolant accident.* The reactor is assumed to have been operating with 100 fuel elements for infinite time atpower P, = 2,000 kW when coolant is lost.* Decay heat is calculated using a SCALE depletion sequence (T-6) based on core burnup to10 MWD per element followed by decay over intervals" Coolant loss is assumed to occur in 5 minutes." Reactor shutdown is assumed to occur with initiation of coolant loss.* Decay heat is from fission product gamma and x rays, beta particles, and electrons.Effects of delayed neutrons are neglected." Thermal power is distributed across the core with a radial peak-to-average ratio of n/2.In individual elements, thermal power is distributed axially according to a sinusoidalfunction." Cladding and gap resistance are assumed to be negligible, i.e., cladding temperature isassumed to be equal to the temperature at the outside surface of the fuel matrix." Cooling of the fuel occurs via natural convection to air at inlet temperature T1 = 3000K.Radiative cooling and conduction to the grid plates are neglected." Heat transfer in the fuel is one dimensional, i.e., axial conduction is neglected, and fuel isassumed to be uniform in thermodynamic and physical properties." Heat transfer in the fuel is treated as pseudo-steady-state behavior, i.e., at any oneinstant, heat transfer is described by steady-state conduction and convection equations.113.5.2 Heat Transfer to AirFundamental relationships between buoyancy driven differential pressure and pressure lossesfrom friction provide a means to calculate fuel and cladding temperature that results from thedecay heat source, related by:(5Pb = '5P f +(P + (5p, + 5P+WhereSee Todreas & Kazemi (1990) or EI-Wakil (1971) for steady-state conduction equations.Page 13-33 CHAPTER 13, ACCIDENT ANALYIS 12/2011(5Pb is the buoyancy force5P f is the pressure difference from friction developed across the fuelP,5P is the pressure difference across the exit restrictioniJP, is the pressure difference across the inlet restriction&Pý is the pressure drop from acceleration\A. Buoyancy ForcesIf pi and P. are respectively the densities of air at the inlet and outlet temperatures,2 thedistance between the center of the zone (1/2 the fuel length Lf) in which the air is heated (inlettemperature) and the center of the zone in which the air is cooled by full mixing (outlettemperature) is 10 hydraulic diameters above the core exit, the distance from the heated lengthto the core exit is Lt, the acceleration of gravity (9.8 m S-2) is g, the buoyancy pressure differenceis given by:Apb = (p, -p).g jL+ L, + 10.D D,B. Friction LossesFriction losses across the lower unheated length, heated length and upper unheated length aregiven by:P 4. L, w24 4.Lf W2 + f L, w2..= f.___ ____ + fl,_"4"L_De 2 .g.p,.A2 ,fFj De 2"g'pf'A,f De 2"g'p ,eWhereLx is the length of the x componentfF,x is the friction factor (23.46/Re) for the x section(x is lower, heated, and upper lengths)A, flow area for the x section per elementC. Losses from Flow RestrictionsInlet and exit losses are calculated by:2 Density at 1 atm, for air as an ideal gas, is given by p (kg/m3) = 353.0/T(°K). Heat capacity, from 300 to 700 'K is1030 J/kgK +/- 3% (Incropera and DeWitt, 1990).Page 13-34 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 13w2.k,.[ ] A, A,[~SPe.+S,= + 1A, 2 g+1 22"- g"- pi,"AC.f 2.- g .,o,. Acceleration losses are given by:'5'aUsing the definition for Reynolds number and values for elevation previously described:r 0.700 °.1491104.42 F 0.153-A + + 0.153/l 102 +LP .P2 I L PI Pave P2A[1.25.p +1.25.p1-1.25.po]=0Where flow w is in lb/h and viscosity M is in units of lb/h-ft. the properties for air for use in theequation are expressed as:40PX -TAnd p is:= 5.739x10-3 + 7.601x10-5 -1 ].78xl 0-" T2Where T is in units of °R, the heat transfer coefficient is calculated through:N, = 6.3 Ra < 1000N.0.806.R°2976 Ra > 1000Where the Nusselt number is:= h.DkThe Rayleigh number is:R e .p2 .g./ .T.cpu.k.LThermal conductivity (from a least squares fit to data presented by Etherington) is:k = 2.377x] 0-4 + 2.995x] 0-5 *T + 14.738-9. T2Page 13-35 CHAPTER 13, ACCIDENT ANALYIS_ .12/2011The specify heat capacity for air (also least squares fit, Etherington) is:cP = 2.413x10-'- 1.780x10-6. T + 1.018x10-8. T2Volumetric expansion coefficients is 13, 6T is the temperature rise over the channel. Theexpression for the Nusselt number was developed from the work of Sparrow, Loeffler, andHubbard for laminar flow between triangular arrays of heated cylinders. The parametersderived above were used as input data for a General Atomics 2-dimensional transient-heattransport compute code used for calculating the systems temperatures after a loss of poolwater. Maximum temperature reached by the fuel are plotted as a definition of operatingpower density in Fig. 13.4A for several cooling or delay times between reactor shutdown andloss of coolant from the core.105.mCJCDa-C'Cooling time. after reactor shutdoaonetCa~se" to limit fu4.W ael tempera-ture veraux potter den.iryTn = M4AX. FUEL TEMIPERATUREm AFTER WATER LOSS (-C),020 30 20 30 4o 50 60 70POWER DENSITY-KW/ELEMENTFigure 13.4A, Cooling TimePage 13-36 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 13I12/2011200018001600COOLING TIME (Sio /Nazlami fuel taspereture versus power densityafter loss of coolant for various cooling.imsa between reactor shutdown and coolant lossLUI-tCLI80060040020000 510 15 20 25 30OPERATING POWER DENSITY-KW/ELEMENTFigure 13.4B, Cooling Time and Power DensityFor reactor operation with maximum power density of 18 kW/element or less, loss of coolantwater immediately upon reactor shutdown would not cause the maximum fuel temperature toexceed 750'C. Operation at maximum power densities above 18 kW/element will not cause fueltemperatures above 7500 C if coolant loss occurs sometime after shutdown, with the decay timerequired depending on power density. Therefore, minimum 5 minutes to effect loss of poolwater adds additional margin.In Fig. 13.4B, data was developed to show time required for natural convective cooling afterreactor shutdown to produce temperatures less than a given value. For instance, temperatureless than 950°C after operating with a maximum power density of 27 kW/element requires aPage 13-37 CHAPTER 13, ACCIDENT ANALYISI12/2011shutdown interval of 3730 s (1.04 h) after shutdown, when decay heat will be low enough thatair cooling is adequate. A 65 minute delay time applies to power density corresponding to a 90element core, and is negligible for power density in a 100 element core.13.5.7 Radiation Levels from the Uncovered CoreAlthough there is only a very remote possibility that the primary coolant and reactor shieldingwater will be totally lost, direct and scattered dose rates from an uncovered core following1,000, 2,000 and 3,500 kW operations are calculated. This section describes calculations of on-site and off-site radiological consequences of the loss-of-coolant accident. Extremelyconservative assumptions are made in the calculations, namely, operation at 2,000 kW for oneyear followed by instant and simultaneous shutdown and loss of coolant. The SCALE depletionsequence (previously referenced for decay heat calculation) is used to generate TRIGA specificcross section libraries for use in ORIGIN ARP for operation over the life of the core (10 MWD perelement). Gamma-ray source strengths, by energy group, are determined by an ORIGEN ARPcalculation. Radiation transport calculations use the MCNP code.Table 13.22, Gamma Source TermMeV0.010.0250.03750.05750.0850.1250.2250.3750.5750.851.251.752.252.753.5579.51Sec1.80E199.09E186.62E182.17E182.12E181.60E189.64E176.54E175.68E75.59E172.69E171.49E175.82E163.70E161.43E164.53E155.22E141.02E1230Min7.11E183.30E182.74E188.30E177.93E175.90E173.48E172.58E172.09E172.60E179.11E165.21E161.57E169.62E151.13E151.63E144.27E108.30E11Hours5.94E182.79E182.40E187.01E176.07E175.34E172.87E172.06E171.76E172.34E176.72E164.22E161.08E167.18E158.01E149.41E131.50E108.22E183.25E181.63E181.61E184.15E173.31E173.89E171.69E171.37Ei71.05E171.58E171.70E161.97E161.70E151.10E153.10E132.73E122.14E98.22E11Days2.32E181.18E181.30E183.09E172.61E173.05E171.38E179.93E167.69E161.26E178.41E151.57E165.79E146.39E148.51E122.26E104.30E78.22E17 30 90 3651.33E185.47E175.86E171.55E171.32E178.03E165.86E163.15E163.90E167.20E162.61E151.06E161.71E144.50E146.19E123.99E68.78E28.17E18.57E173.39E172.70E178.93E165.93E163.22E163.01E168.72E151.76E165.23E167.44E143.02E159.76E131.30E141.86E123.80E68.59E27.98E15.48E172.19E171.58E175.46E163.90E162.07E161.57E163.34E155.19E153.27E162.02E141.76E147.98E136.62E121.62E113.39E68.16E27.58E12.58E171.02E177.54E162.58E161.90E161.02E165.99E151.60E159.38E146.30E159.37E133.49E134.10E139.61E115.42E102.04E67.15E26.63E1Modeling ofthe reactor core (Appendix 13.2) was performedusing approximate geometrydescribed in Fig. 13.5 and 13.6. The TRIGA reactor core is approximated as a right circularcylinder with the outer diameter of the G ring and a fuel region 0.381 m (15 in.) high. Axialzones are defined above and below the fuel, and at the grid plate elevations. The zones aredescribed in Table 13.23, including the height of the zone and identification of materials inPage 13-38 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 13I12/2011locations defined by fuel positions (FUEL POS) and materials outside the(CHANNEL).fuel positionsTable 13.23, Height/Thickness Dimensions of Unit CellZone THICKNESS/LENGTH CHANNEL FUEL POS1 LOWER GRID PLATE 3.27 cm AL VOID, SS2 LOWER ELEMENT 12.70 cm VOID(1) End Cap (Lower) 5.09 cm VOID VOID, SS(2) Graphite 8.36 cm VOID Gr, SS3 FUEL 38.10 cm VOID FUEL, SS4 UPPER ELEMENT 11.58 cm VOID(1) Graphite 8.36 cm VOID Gr, SS(2) End Cap (Upper) 3.22 cm VOID VOID, SS5 UPPER GRID PLATE 1.59 cm AL VOID, SSMass fractions of material components are calculated assuming a unit cell based on the fuelelement pitch. A unit cell is the total area defined by the section of three fuel elements that liewith in the area formed by connecting three fuel center points (Table 13.24, Unit cell Area).Materials within the unit cell are eight fuel, graphite (assumed to have the same cross section asfuel), cladding, or void. The areas are listed in Table 13.24.Table 13.24, Unit Cell AreasUNIT CELLUnit cell area 8.2071 cm2Fuel 4.7886 cm2Cladding 0.3397 cm2Channel void 3.0788 cm3Materials in the volumes described by the heights/thicknesses Table 13.23 and the areas inTable 13.24 are homogenized based on material characterizations in Table 13.25.Table 13.25, MaterialCharacterizationCOMPONENT VALUE UNITFUELU235U238ZrHSS 304Fe38.00156.872052.3845.36GGGG0.6993 %Page 13-39 CHAPTER 13, ACCIDENT ANALYISIl12/2011CHAPTER 13, ACCIDENT ANALYIS 12/2011CrNiMnSiPSGraphiteAluminumSS3040.19000.10000.02000.01000.00050.00032.252.708.03g/ccg/ccg/ccZone 5Zone 4Reflector Zone 3 ReflectorZone 2Zone 1Figure 13.5, Core ModelBiological shielding is approximated as a two-section concrete cylinder based on dimensions inChapter 4. The structure was simplified as rectangular for this calculation, and the top deckneglected.< Co.Figure 13.6A, Bay Model Top ViewFigure 13.6B, Bay Model Cross SectionThe site boundary is about 75 m at its nearest approach to the north wall of the reactor bay(87.5 m from the core center), with a fence erected 70 m from the reactor bay wall (82.5 m fromPage 13-40 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, CHAPTER 13I12/2011the core center). Receptor locations for dose calculations inside the reactor bay were set at 1foot from (1) the ground floor personnel door, (2) the center of the truck door, (3/4) in line withthe core at the north and west walls, (5) the top floor personnel door, and directly over the core.Receptor locations for dose calculations outside the reactor bay were set 1 foot outside thewalls of the reactor bay in line with the core on the three sides with exterior faces. Additionalpoints were set 80 and 90 meters from the core center.The building geometry is simplified to single thickness walls, and the floor structures areneglected. The colocation boundary extends about 4 meters into the ground below the reactorbay, and spherically to approximately 700 meters.< \~] T/////L II J I iFigure 13.7C, Top ViewFigure 13.7A, Building Model Figure 13.7B, MCNP Side ViewThe results of the calculation are provided in Table 13.26.Table 13.26, Post LOCA DosesSec Min. hours days1 30 1 8 1 7 30 90 365R/hLower bay door 3.66 0.28 0.228 0.084 0.064 0.106 0.042 0.025 0.010Lower bay east wall 3.64 0.28 0.207 0.101 0.065 0.111 0.045 0.023 0.011Lower bay west wall 4.75 0.35 0.264 0.120 0.083 0.123 0.058 0.032 0.013Mid-truck door 4.65 0.36 0.286 0.112 0.087 0.154 0.077 0.042 0.013Top deck over core 14801 948 754 301 206 324 135 73 28Top deck door 26.58 1.82 1.590 0.704 0.489 0.720 0.311 0.190 0.068mR/hOutside east wall 0.906 0.0712 0.0529 0.0107 0.0079 0.0165 0.0067 0.0035 0.0015Outside west wall 1.547 0.1062 0.0923 0.0341 0.0204 0.0374 0.0144 0.0083 0.0026Outside north wall 1.035 0.0678 0.0590 0.0232 0.0107 0.0167 0.0069 0.0047 0.0040Approx. Parking Lot 2.475 0.0732 0.0659 0.0145 0.0151 0.0180 0.0070 0.0053 0.0019Approx. Fence Line 1.615 0.0722 0.0540 0.0121 0.0132 0.0134 0.0063 0.0056 0.0023Page 13-41 CHAPTER 13, ACCIDENT ANALYIS 12/201113.5.8 Results and ConclusionsAlthough a loss of pool water is considered to be an extremely improbable event, calculationsshow the maximum fuel temperature that could be expected to result from such an event (afterlong-term operation at full power of 2,000 kW is 7500C, well below any safety limit for TRIGAreactor fuel.Maximum possible dose rates resulting from a complete loss of pool water permit mitigatingactions. The area surrounding the reactor is under control of the University of Texas, andexposures outside the reactor bay environment can be limited by controlling accessappropriately. The University of Texas has complete authority to control access to campuslocations.13.6 Loss of Coolant Flow13.6.1 Initialing Events and ScenariosLoss of coolant flow could occur due to failure of a key component in the reactor primary orsecondary cooling system (e.g., a pump), loss of electrical power, blockage of a coolant flowchannel, or operator error.The UT TRIGA reactor pool tank holds 40.57 m3 (10717 gallons) of water, or about 40570 kg ofwater. At a steady-state power level of 1 MW, the bulk water temperature would increaseadiabatically at a rate of about 20.74-C MW-' h-'.Under these conditions, the operator has ample time to reduce the power and place the heat-removal system back into operation before a high temperature is reached in the reactor bulkwater. Control console instruments indicate pool temperature, heat exchanger inlet and outlettemperature. Alarms are provided for heat exchanger low differential pressure (pool to chillwater), pool water temperature, and abnormal water level (hi or low). A reactor scram occurs atlow-low water level. These indicators allow the operator to observe an abnormal condition andmake corrections or secure operations, and prevent operating the reactor with low water poolwater level.13.6.2 Accident Analysis and Determination of ConsequencesIf the UT TRIGA was operated without coolant flow for an extended period of time, and therewas no heat removal by the reactor coolant systems, voiding of the water in the core couldoccur and the water level in the reactor tank would decrease because of evaporation. Thesequence of events postulated for this very unlikely scenario is as follows:The reactor would continue to operate at a power level of 1 MW (provided the rods wereadjusted to maintain power) and would heat the tank water at a rate of about 0.35TC m-1 forPage 13-42 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 13approximately 66 minutes until the tank water reached the maximum allowed operatingtemperature. It is considered inconceivable that such an operating condition with the attendantalarms and indications would not be undetected.If it is assumed that the operator or automatic control system continued to maintain power at 1MW, and assuming that the system is adiabatic except for the evaporation process, pool waterwould evaporate until the pool low level scram setpoint is reached and the reactor wouldshutdown.13.7 Mishandling or Malfunction of Fuel13.7.1 Initiating Events and ScenariosEvents which could cause accidents at the UT TRIGA in this category include:* Simple failure of the fuel cladding due to a manufacturing defect or corrosion) and* Fuel handling accidents where an element is dropped underwater and damaged severelyenough to breach the cladding,Overheating of the fuel with subsequent cladding failure during steady-state or pulsingoperations.In the experience at UT, cladding failures from manufacturing defects occur before the elementhas enough operating history to generate a significant quantity of fission products.13.7.2 AnalysisGaseous fission product releases in water are delayed or partially retained (because of gassolubility) slowed (in the case of gas). Particulate fission product releases are substantiallyretained. Therefore a cladding failure under water is bounded by cladding failure in air, themaximum hypothetical accident.13.8 Experiment Malfunction13.8.1 Accident Initiating Events and ScenariosImproperly controlled experiments involving the UT TRIGA reactor could potentially result indamage to the reactor, unnecessary radiation exposure to facility staff and members of thegeneral public, and unnecessary releases of radioactivity into the unrestricted area. Mechanismsfor these occurrences include the production of excess amounts of radionuclides withPage 13-43 CHAPTER 13, ACCIDENT ANALYIS 12/2011unexpected radiation levels, and the creation of unplanned pressures in irradiated materials.These materials could subsequently vent into the irradiation facilities or into the reactor roomcausing damage from the pressure release or an uncontrolled release of radioactivity. Othermechanisms for damage, such as large reactivity changes, are also possible.13.8.2. Analysis and Determination of ConsequencesThere are two main sets of procedural and regulatory requirements that relate to experimentreview and approval. These are the UT Reactor Procedures and the Technical Specifications.These requirements are focused on ensuring that experiments will not fail, and they alsoincorporate requirements to assure that there is no reactor damage and no radioactivityreleases or radiation doses which exceed the limits of 10 CFR 20, should failure occur. Forexample, the detailed procedures call for the safety review and approval of all reactorexperiments.A. Administrative ControlsSafety related reviews of proposed experiments require the performance of specific safetyanalyses of proposed activities to assess such things as generation of radio nuclides and fissionproducts, and to ensure evaluation of reactivity worth, chemical and physical characteristics ofmaterials under irradiation, corrosive and explosive characteristics of materials, and the need forencapsulation. This process is an important step in ensuring the safety of reactor experimentsand has been successfully used for many years at research reactors to help assure the safety ofexperiments placed in these reactors. Therefore, this process is expected to be an effectivemeasure in assuring experiment safety at the UTTRIGA reactor.B. Reactivity ConsiderationsA Technical Specifications limit of $1.00 has been placed on the reactivity worth of non-securedexperiments. This is designed to prevent an inadvertent pulse by experiment manipulation, andis well below the maximum reactivity limit analyzed in the insertion of excess reactivity of 13.4.A Technical Specifications limit of $1.00 has been placed on the reactivity worth of any singleexperiment. This is designed to prevent an inadvertent pulse by experiment manipulation whileoperating at power, and is well below the maximum reactivity limit analyzed in the insertion ofexcess reactivity of 13.4. Since these experiments are secured the transient that occurs fromremoval while operating at power will be less severe, and reactor protective systems areexpected to terminate operations.A Technical Specifications limit of $3.00 has been placed on the reactivity worth of allexperiments during an operation. Removal of all experiments while operating is bounded by thepositive reactivity addition analysis.Page 13-44 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011C. Fueled Experiment Fission Product InventoryLimiting the generation of certain fission products in fueled experiments ensures thatoccupational radiation doses as well as doses to the general public, due to experiment failurewith subsequent fission product release, will be within the limits prescribed in 10 CFR 20. DACratio, as previously used, indicates the radionuclide concentration to which an exposedindividual can receive 5 rem TEDE in a 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> exposure. The DAC ratio for the activity of aspecific nuclide (Ar) of an element distributed in a volume (V) is defined by:DAC.,The sum of the fractions for all nuclides determines an effective DAC fraction which meets DACrequirements if the sum is less than or equal to 1. For a fission product distribution yield acrossan element, if the yield is defined as Y% then the fraction can be calculated:AE/'./.y%F,= /VDAC.Therefore the total DAC fraction for the element is calculated:AE Y%V For a target DAC fraction, activity can be calculated:AE = F.-' Y%xDACrThe ORIGEN source term calculations were used to calculate fractional fission product yields foriodine and strontium. The calculation assumes a 5 minute decay time after the reactor is shutdown until the source term calculations are initiated; this is conservative from a practicalperspective in considering the removal process. The weighted elemental yield fraction, and theweighted yield normalized to reactor bay volume is provide in Table 13.27.Page 13-45 CHAPTER 13, ACCIDENT ANALYIS12/2011CHAPTER 13, ACCIDENT ANALYIS 12/2011Table 13.27, Calculations Supporting Limits onFueled ExperimentsIsotope Isotope Weightedisotope Yield DAC Yieldi125 6.6E-15 3.OE-8 2.2E-7i128 1.6E-5 5.OE-5 3.3E-1i129 8.7E-9 4.OE-9 2.2i130 9.2E-5 3.OE-7 3.1E2i131 1.OE-1 2.OE-8 5.2E6i132 1.5E-1 3.OE-6 5.1E4i133 2.4E-1 1.OE-7 2.4E6i134 2.8E-1 2.OE-5 1.4E4i135 2.2E-1 7.OE-7 3.2E58.OE+06VOL/SUM 4.66E2sr85 1.41E-11 6E-7 2.36E-5sr85m 7.65E-12 3E-4 2.55E-8sr87m 2.20E-8 5E-5 4.39E-4sr89 2.67E-1 6E-8 4.45E6sr90 6.96E-2 2-9 3.48E7sr9l 3.28E-1 1E-6 3.28E5sr92 3.35E-1 3E-6 1.12E53.97E7VOL/SUM 9.35E1For a 2-hour evacuation period, the DAC fraction is 1000; therefore a total iodine activity of4.66E5 IICi will allow an individual to meet the annual 1OCFR20 dose limits for radiation workersassuming a 2-hour evacuation period, and 9.32E5 GCi will allow an individual to meet the annual10CFR20 dose limits for radiation workers assuming a 1-hour evacuation period. Similarly, a9.35E4 VICi strontium inventory is acceptable for a 2-hour evacuation period. Therefore, limitingexperiment radioiodine and strontium inventories in experiments will assure that there isadequate time for taking corrective actions.D. ExplosivesProjected damage to the reactor from experiments involving explosives varies significantlydepending on the quantity of explosives being irradiated and where the explosives are placedrelative to critical reactor components and safety systems. If in the reactor tank, the UT TRIGAreactors Technical Specifications limit the amount of explosive materials, such as gunpowder,TNT, nitroglycerin, or PETN, to quantities less than 25 milligrams. Also, the TechnicalSpecifications state that the pressure produced upon detonation of the explosive must havebeen calculated and/or experimentally demonstrated to be less than the design pressure of thecontainer. The following discussion shows that the irradiation of explosives up to 25 milligramscould be safely performed if the containment is properly chosen. A 25-milligram quantity ofexplosives, upon detonation, releases approximately 25 calories (104.6 joules) of energy, withPage 13-46 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 13the creation of 25 cm3 of gas. For the explosive TNT, the density is 1.654 g/cm3, so that 25 mgrepresents a volume of 0.015 cm3.If the assumption is made that the energy release occurs asan instantaneous change in pressure, the total force on the encapsulation material is the sum ofthe two pressures. For a 1 cm3 volume, the energy release of 104.2 joules represents a pressureof 1,032 atmospheres. The instantaneous change in pressure due to gas production in the samevolume adds another 25 atmospheres. The total pressure within a 1 cm3 capsule is then 1,057atmospheres for the complete reaction of 25 mg of explosives. Typical construction materials ofcapsules are stainless steel, aluminum, and polyethylene; Table 13.28 lists the mechanicalproperties of these encapsulation materials.Table 13.28, Material StrengthsMaterial Yield Strength Ultimate Strength Density (g/cm3)(Kpsi) (Kpsi)Stainless Steel (304) 35 85 7.98Aluminum (6061) 40 45 2.739Polyethylene 1.7 1.4 0.923Analysis of the encapsulation materials determines the material stress limits that must exist toconfine the reactive equivalent of 25 mg of explosives. The stress limit in a cylindrical containerwith thin walls is one-half the pressure times the ratio of the capsule diameter-to wall thickness.This is the hoop stress. The hoop stress is 2 times the longitudinal stress, and hence hoop stressis limiting. Thus:p.d2.tWhereaxiS the maximum hoop stress in the container wallp is the total pressure in the containerd is the diameter of the container, andt is the container wall thicknessWhen evaluating an encapsulation material's ability to confine the reactive equivalent of 25 mgof explosives, the maximum stress in the container wall is required to be less than or equal tothe yield strength of the material:p2d< .ieSolving this equation for d/t provides an easy method of evaluating an encapsulation material:d < 2

• aQieldt pPage 13-47 CHAPTER 13, ACCIDENT ANALYIS ..12/2011Assuming an internal pressure of 1,057 atmospheres (15,538 psi), the maximum values of d/t forthe encapsulation materials are displayed in Table 13.28. The results indicate that apolyethylene vial is not a practical container since its wall thickness must be at least 4.5 timesthe diameter. However, both the aluminum and the stainless steel make satisfactory containers.As a result of the preceding analysis, a limit of 25 mg of TNT-equivalent explosives is deemed tobe a safe limitation on explosives which may be irradiated in facilities located inside the reactortank, provided that the proper container material with appropriate diameter and wall thicknessis used.Table 13.29, Container Diameter to Thickness RatioMaterial d/tStainless Steel (304) 4.5Aluminum (6061) 5.1Polyethylene 0.2213.9 Loss of Normal Electric Power13.9.1 Initiating Events and ScenariosLoss of electrical power to the UT TRIGA reactor could occur due to many events and scenariosthat routinely affect commercial power.13.9.2 Accident Analysis and Determination of ConsequencesSince the UT TRIGA does not require emergency backup systems to safely maintain core cooling,there are no credible reactor accidents associated with the loss of electrical power. Backuppower for lighting is provided by an emergency diesel on the Pickle Research Campus, and thereare emergency exit lights and hand-held battery-powered lights located throughout the facilityto allow for inspection of the reactor and for an orderly evacuation of the facility. Loss of normalelectrical power during reactor operations requires that an orderly shut down is to be initiatedby the operator on duty. The backup power supply will allow monitoring of the orderly shutdown and verification of the reactor's shutdown condition.13.10 External Events13.10.1 Accident Initiating Events and ScenariosHurricanes, tornadoes, and floods are virtually nonexistent in the area around the UT TRIGAreactor. Therefore, these events are not considered to be viable causes of accidents for thereactor facility. In addition, seismic activity in the area as indicated in Chapter 2 is acceptablylow.Page 13-48 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 1313.10.2 Accident Analysis and Determination of ConsequencesThere are no accidents in this category that would have more on-site or off-site consequencesthan the MHA previously analyzed, and, therefore, no additional specific accidents are analyzedin this section.13.11 Experiment Mishandling or Malfunction13.11.1 Initiating Events and ScenariosNo credible accident initiating events were identified for this accident class. Situations involvingan operator error at the reactor controls, a malfunction or loss of safety-related instruments orcontrols, and an electrical fault in the control rod system were anticipated at the reactor designstage. As a result, many safety features, such as control system interlocks and automatic reactorshutdown circuits, were designed into the overall TRIGA Control System (SAR Chapter 7). TRIGAfuel also incorporates a number of safety features (SAR Chapter 4) which, together with thefeatures designed into the control system, assure safe reactor response, including in some casesreactor shutdown. Malfunction of confinement or containment systems would have thegreatest impact during the MHA, if used to lessen the impact of such an accident. However, nosafety considerations at the UT TRIGA depend on confinement or containment systems. Loss ofpool water was previously addressed. Although no damage to the reactor occurs as a result ofthese leaks, the details of the previous analyses provide a more comprehensive explanation.13.10.2 Accident Analysis and Determination of ConsequencesSince there were no credible initiating events identified, no accident analysis was performed forthis section and no consequences were identified.Page 13-49 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, APPENDIX 13.1'Input generated by GeeWiz SCALE 6.1 Compiled on Mon Jun 6 11:04:33 2011=t6-depl parm=(addnux=0,MAXDAYS=800)TRIGA FUEL BURN TO GENERATE DECAY AND RADIONUCLIDE SOURCE TERMS238groupndf5Mixture Compositionsread compositionwtptss304 4 7.8 82600067.8524000 18280009.825055 1.814000 16000 0.815031 0.45160000.31300 end'Graphite for axial then radial reflectorsgraphite 5 1300 endgraphite 6 1300 end'Aluminum for sheet then smeared for RSR volumealuminum 7 1300 endaluminum 8 0.2 300 endwtptair 9 0.00123 27014 80.08016 20.01 300 endwtptrods 10 2.5 35010 16.05011 64.06012 20.01 300 endend compositionPage 13.1-1 APPENDIX 13.1, T-6 DEPLETION ANALYSIS INPUT FILE FOR SCALE CALCULATION 12/2011Depletion Specificationsread depletion1end depletionread burndataend burndataread keeporigenend keepread opustitle="uranium isotopes (grams/mtihm)"symnuc=U-235 U-238 endunits=gramstime=dayssort=nonrank=2title="uranium isotopes (grams/mtihm)"matl=1 endnew casetitle="fission products (curies)"units=curiestime=daysnrank=100new casetitle="decay heat (watts)"units=wattstime=daysend opusread modelRun-time Parametersread parametergen=250npg=1000nsk=50htm=yesPage 13.1-2 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, APPENDIX 13.1end parameterGeometryread geometryunit 1com="UNIT 1: FUEL CHANNEL"com="Upper axial reflector (inside 10)"cylinder 10 1.8222 31.909 19.053)")"hexprism 30 2.177 31.909 -31.909media 5 1 10media 1 1 11 -12media 5 1 12media 5 1 13media 3 1 30 -20media 4 1 20 -10 -11 -13boundary 30unit 2com="UNIT 2: GRAPHITE ROD CHANNEL"Com="Graphite (inside 10)"cylinder 10 1.8222 31.909 -31.909Com="Cladding (inside 11, not in 10)"cylinder 11 1.873 31.909 -31.909media 6 1 10media 4 1 11 -10media 3 1 20 -11boundary 20unit 3com="UNTI 3: WATER CHANNEL"Page 13.1-3 APPENDIX 13.1, T-6 DEPLETION ANALYSIS INPUT FILE FOR SCALE CALCULATION 12/2011Com="Fuel cell boundary, filled with water"hexprism 10 2.177 31.909 -31.909media 3 1 10boundary 10unit 4com="UNIT 4: STANDARD CONTROL ROD"Com="aluminum above boron region"cylinder 10 1.619 31.909 19.05 ORIGIN X=O Y=0 Z=20Com="Boron region"cylinder 11 1.619 19.05 -19.05 ORIGIN X=0 Y=0 Z=20Com="al spacer between boron and fuel"cylinder 12 1.619 -19.05 -21.59 ORIGIN X=0 Y=0 Z=20media 7 110-11 -12media 10 111-12media 7 1 12media 1113 -14media 2 1 14media 4 1 20 -10 -11 -12 -13media 3 1 30 -20boundary 30Unit 5I com="UNIT 5: PULSE ROD"Com="aluminum above boron region (inside 10)"cylinder 10 1.519 31.909 19.05 ORIGIN X=0 Y=0 z=38.1Com="Boron region (inside 11)"cylinder 11 1.519 19.05 -19.05 ORIGIN X=0 Y=0 Z=38.1Com="Aluminum spacer (inside 12)"cylinder 12 1.519 -19.05 -21.59 ORIGIN X=0 Y=0 Z=38.1Com="Air in follower (inside 13)"cylinder 13 1.519 -21.59 -64.61 ORIGIN X=0 Y=0 Z=38.1Com="Cladding (inside 20, not in 10, 11, 12, Or 13)"cylinder 20 1.59 31.909 -64.61 ORIGIN X=0 Y=0 Z=38.1Com="Control rod cell boundary (water inside 30, not in 20)"Page 13.1-4 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, APPENDIX 13.1hexprism 30 2.177 31.909 -31.909media 7 110media 10 1 11media 7 1 12media 9 113media 7 120 -10 -11 -12 -13media 3 130 -20boundary 30media 6 1 40 -24 -11 -50 -51 -52 -53media91 50 40 -11media 9 1 51 40 -11media 9 1 52 40 -11Page 13.1-5 APPENDIX 13.1, T-6 DEPLETION ANALYSIS INPUT FILE FOR SCALE CALCULATION 12/2011media91 53 40 -11RSRmedia 9 120 -11media 7 121 -20media 7 122 -21media 8 123 -22media 7 124 -23boundary 40end geometryread arrayara=1 nux=15 nuy=15 nuz=1 typ=shexagonalfill'INITIAL CRITICALITY 3/16/1992'LHS RHS'BOTTOM OF ARRAY'AX are apexes, W water, F fuel, G graphite, PR pulse rod,'RR reg rod, SX shim rod'S source is' not modeled -position is water filled'TOP OF ARRAYPage 13.1-6 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, APPENDIX 13.1end fillend arrayend dataend modelend#shellcopy ft71fO01 "%RTNDIR%"\TRIGA.ft71endPage 13.1-7 THE UNIVERSITY OF TEXAS TRIGA I RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, APPENDIDIX 13.2'This SCALE input file was generated by'OrigenArp Version 6.1 Compiled on Thu Oct 7 11:31:00 2010#shellcopy "C:\NEW SCALE CALCS\TRIGA GAMMA SOURCE TERM.f71" "ft7lf00l"end#origens0$$ all71etDecay Case3$$ 21 1 10 a16 2 a33 18 et35$$ 0 t54$$ a8 1 all 0 e56$$ a2 10 a6 I alO 0 a13-21 a15 3 a17 2 e57** 0 a3 le-05 e95$$ 0 tcase 10 MTU60** 0.001042 0.003125 0.009375 0.028125 0.041667 0.125 0.375 1 3 961** fO.0565$$'Gram-Atoms Grams Curies Watts-All Watts-Gamma3z 1 0 0 3z 3z 3z 6z3z 1 0 0 3z 3z 3z 6z3z 1 0 0 3z 3z 3z 6z81$$ 2 0 26 1 e82$$ 2 2 2 2 2 2 2 2 2 2 e83**1.1000000e+07 8.0000000e+06 6.0000000e+06 4.0000000e+06 3.0000000e+062.5000000e+06 2.0000000e+06 1.5000000e+06 1.0000000e+06 7.0000000e+054.5000000e+05 3.0000000e+05 1.5000000e+05 1.0000000e+05 7.0000000e+044.5000000e+04 3.0000000e+04 2.0000000e+04 0.O000000e+00 et56$$00alO let56$$00alO 2et56$$00alO 3et56$$00alO 4et56$$00alO 5et56$$00alO 6et56$$00alO 7et56$$00alO 8et56$$0OalO 9et56$$OOalO 10et54$$ a8 1 all0 e56$$ a2 5 a6 1alO 10 a15 3 a17 2 ePage 13.1-1 APPENDIX 13.2, ORIGEN ARP INPUT 12/201157** 9 a3 le-05 e95$$ 0 tCase 20 MTU60** 27 30 90 180 36561** fO.0565$$'Gram-Atoms Grams Curies Watts-All Watts-Gamma3z 1 0 0 3z 3z 3z 6z3z 1 0 0 3z 3z 3z 6z3z 1 0 0 3z 3z 3z 6z81$$ 2 0 26 1 e82$$ 2 2 2 2 2 e83**1.1000000e+07 8.0000000e+06 6.0000000e+06 4.0000000e+06 3.0000000e+062.5000000e+06 2.0000000e+06 1.5000000e+06 1.0000000e+06 7.0000000e+054.5000000e+05 3.0000000e+05 1.5000000e+05 1.0000000e+05 7.0000000e+044.5000000e+04 3.0000000e+04 2.0000000e+04 0.O000000e+00 et56$$00alO let56$$00alO 2et56$$00alO 3et56$$00alO 4et56$$OOalO 5et56$$ fO tend=opusLIBUNIT=21TYPARAMS=N UCLIDESUNITS=WATTSLIBTYPE=ALLTIME=DAYSNPOSITION=22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 endend=opusLIBUNIT=21TYPARAMS=NUCLIDESUNITS=CURIESLIBTYPE=ALLTIME=DAYSNPOSITION=22 23 24 25 26 27 28 29 30 31 endend#shellcopy ft7lf001 "C:\NEW SCALE CALCS\TRIGA GAMMA SOURCE TERM.f71"Page 13.2-2 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, APPENDDIX 13.2del ft71fO01endPage 13.1-3 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, APPENDX 13.3c Created on: Tuesday, May 25, 2010 at 16:20c *****************CELL CARDS ***********************1 0 75 :-66C12 8 -1.6104 -75 -33 66 $ DIRT BELOW EVEYTHING13 9 -0.001205 -75 45 $ AIR ABOVE EVERTHING14 8 -1.6104 -75 (13 :31 :-2 :-14 )-75 33 -34 $ DIRT OUTSIDE FTPRT15 9 -0.001205 -75 (13 :31 :-2 :-14 )-75 34 -45 $ AIR OUTSIDE FTPRTC16 8 -1.6104 14 -18 6 -13 -34 33 $ DIRT WEST OF BAY17 9 -0.001205 14 -18 6 -13 34 -45 $ AIR WEST OF BAYC18 8 -1.6104 22 -25 2 -6 -34 33 $ DIRT SOUTH OF RX WING19 9 -0.001205 22 -25 2 -6 34 -45 $ AIR SOUTH OF RX WINGC20 8 -1.6104 25 -29 2 -4 -34 33 $ DIRT SOUTH OF OFFICE WING21 9 -0.001205 25 -29 2 -4 34 -45 $ AIR SOUTH OF OFFICE WINGC22 8 -1.6104 29 -31 2 -6 -34 33 $ DIRT SOUTH OF STAIRS23 9 -0.001205 29 -31 2 -6 34 -45 $ AIR SOUTH OF STAIRSC24 8 -1.6104 26 -31 11 -13 -34 33 $ DIRT NORTH OF OFFICE WING25 9 -0.001205 26 -31 11 -13 34 -45 $ AIR NORTH OF OFFICE WING2c ABOVE BUILDING26 9 -0.001205 26 -29 4 -11 43 -45 $ AIR OVER OFFICE WING27 9 -0.001205 25 -26 4 -6 43 -45 $ Accounts for wall interface28 9 -0.001205 29 -316 -11 43 -45 $ AIR OVER OFFICE WING STAIRPage 13.3-1 APPENDIX 13.3, MCNP INPUT FOR LOCA DOSESI12/2011APPENDIX 13.3, MCNP IN PUT FOR LOCA DOSES 12/201129 9 -0.001205 14 -22 2 -6 41 -45 $ AIR OVER N-GEN ROOMc INSIDE BUILDING30 10 -2.3 14 -22 2 -6 33 -41 #31 $ n-gen room shell31 9 -0.001205 15 -21 3 -8 33 -40 $ n-gen room volumec3233C343536C373810 -2.3 18 -26 6 -10 33 -45 #33 $ rx wing shell9 -0.001205 19 -25 7 -10 33 -44 $ rx wing volume10 -2.3 (26 -29 4 -11 33 -43) #36 $ office wing shell10 -2.3 (25 -26 4 -6 33 -43)9 -0.001205 (26 -28 5 -10 33 -42)10 -2.3 29 -31 6 -11 33 -43 #38 $ office wing stairwell she9 -0.001205 29 -30 7 -10 33 -42 $ office wing stairwell volc14 px -609.6 $ n-gen room walls15 px -487.68 $ n-gen room interior27px 1880 $ office wing wallPage 13.3-2 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTORSAFETY ANALYSIS REPORT, APPENDX 13.3-1 12/2011C282930313233343536373839404142px 4260 $ wallpx 4300.92 $ lab wing wallpx 4840 $ building wallpx4876.8 $ buildingpz -365 $ concrete pad/foundation 12 ftpz 0 $ bay floor ALREADY SURFACE 12pz 366 $ ground at side of buildingpz 183 $ ground at parking lotpz 365 $ office wing basement ceilingpz 396 $ office wing 1st floor floorpz 762 $ office wing 1st floor ceilingpz 792 $ office wing 2nd floor floorpz 945 $ n-gen room ceilingpz 1240.526 $ n-gen room roof 633.984 cm < baypz 1158 $ office 2nd floor ceiling43 pz 1183.4 $ office 2nd floor roof 411.48 cm < bay44 pz 1840.52 $ bay roof45 pz 1865.92 $ bay roofC POOL AND POOL WALL CORE CENTER (x,y) = (655,655)46 1 px-9947 1 px 9948 1 py 049 1 py-9950 2 cz 9951 2 c/z-70-70 9952 1 pz853.4453 1 px-99.654 1 px 99.655 2 cz 99.656 2 c/z -70 -70 99.6c LOWER SHIELD (pz=0 to pz=548.64)57 1 px-342.958 1 px 342.959 1 py 316.2660 1 py-415.2661 1 pz 548.64c STEM (pz=548.64 to pz=853.44)62 1 px -190.563 1 px 190.564 1 py 190.565 1 py-289.56cPage 13.3-3 APPENDIX 13.3, MCNP INPUT FOR LOCA DOSESI12/201166 pz -500CC75 s 655 655 0 21000*trl 655 655 0 135 315 90 45 45 90 90 90 0 $ 45 degree at (655,655)*tr2 655 655 0mode pnps 10000C LGP 3.378563ml 26000 -0.346163366 24000 -0.09405940628000 -0.04950495 25000 -0.0099009914000 -0.004950495 15000 -0.00022277216000 -0.000148515 13000 -0.4950495052800014000-0.061516567-0.00615165725000 -0.01230331315000 -0.000276825Page 13.3-4 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, APPENDX 13.316000 -0.00018455 13000 -0.372531019c composition of radial reflectorm6 6000. -1 $MATC material: nominal soil d=1.6104 g/cmA3; .05 bound water contentc .20 free water contentm8 1000. -0.02331 $MAT8000. -0.55922 14000. -0.22259 13000. -0.0652826000. -0.04015 20000. -0.02915 19000. -0.020811000. -0.02272 12000. -0.01678 18000. -0.0128m9 6000. -0.000124 $ Air 0.001205 g/cc7014. -0.755268 8016. -0.231781 18000. -0.012827ml0 1001. -0.0221 $ Normal concrete 2.3 g/cc6012. -0.002484 8016. -0.57493 11023. -0.01520812000. -0.001266 13027. -0.019953 14000. -0.30462719000. -0.010045 20000. -0.042951 26000. -0.006435mll 1000. -0.003585 $ Barite concrete 2.8 g/cc (up to 3.5 g/cc)8000. -0.311622 12000. -0.001195 13000. -0.00418314000. -0.010457 16000. -0.107858 20000. -0.05019426000. -0.047505 56000. -0.4634imp:p 0 1 36r $ 1, 36c

• RUN CARDS***********************sdef cel=9 erg=dl axs=O 0 1 pos=655 655 98.91 rad=d2 ext=d3S12 0 25.4si3 -19.05 19.05cc ** 3.5 MW 0.12E-04 dayssil a1.OOOE-02 2.500E-02 3.750E-02 5.750E-02 8.500E-021.250E-01 2.250E-01 3.750E-01 5.750E-01 8.500E-011.250E+00 1.750E+00 2.250E+00 2.750E+00 3.500E+005.OOOE+00 7.OOOE+00 9.500E+00spi1.798E+19 9.090E+18 6.621E+18 2.174E+18 2.124E+181.599E+18 9.635E+17 6.540E+17 5.683E+17 5.588E+172.685E+17 1.486E+17 5.819E+16 3.698E+16 1.432E+164.528E+15 5.217E+14 1.020E+12cF5:p 25 25 2001fl5z:p 566 1500 1 566 1800 1 566 4900 1 566 8000 1566 9000 1 566 10000 1 566 15000 1 566 20000 1ccfsl5 4 10 13 15 18 22 26 52 61cPage 13.3-5 APPENDIX 13.3, MCNP INPUT FOR LOCA DOSES 12/2011fm5 9.749E17fc5 Tally multiplied by 2.708E14 photons/s per kWtimes 3600 s/h to yield Sv/h per kilowattfm15 9.749E17fc15 Tally multiplied by 2.708E14 photons/s per kWtimes 3600 s/h to yield Sv/h per kilowattcc Ambient dose conversion (Sv cmA2) -ICRP 51, 1987de0 0.01 0.015 0.02 0.03 0.04 0.05 0.06 0.08 0.100.15 0.20 0.30 0.40 0.50 0.60 0.80 1 1.523456810dfO 0.0769E-12 0.846E-12 1.O1E-12 0.785E-12 0.614E-120.526e-12 0.504E-12 0.532E-12 0.611E-12 0.890E-121.18e-12 1.81E-12 2.38E-12 2.89E-12 3.38E-124.29e-12 5.11E-12 6.92E-12 8.48E-12 11.1E-1213.3e-12 15.4E-12 17.4E-12 21.2E-12 25.2E-12cPhys:p 10 1 1prdmp 3j 3 ljprintPage 13.3-6 THE UNIVERSITY OF TEXAS TRIGA 11 RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, CHAPTER 1515.0 FINANCIAL QUALIFICATIONS15.1 Financial Ability to Operate a Nuclear Research ReactorThe University of Texas is a State owned entity, as documented in Appendix 15.1. UT hasoperated a TRIGA nuclear research reactor since 1967. In 1998, UT decided to decommission a250 kW TRIGA located on the main campus and construct a new 1.1 MW TRIGA on the PickleResearch Campus. The PRC facility has operated successfully, continuously since granted afacility operating license in 1991. Recent facility budgeting and expenditures was used todevelop an estimate of operating costs and income for the next five years (Appendix 15.2).15.2 Financial Ability to Decommission the FacilityThe University of Texas intends to renew the facility operating license. Whenever a decision ismade to terminate operations and decommission the facility, the university will seek legislativeappropriations of funds from the State of Texas, as indicated in indicated in Appendix 15.2.15.3 BibliographyNUREG/CR-1756 "Technology, Safety, and Costs of Decommissioning Reference NuclearResearch and Test Reactors," U.S. Nuclear Regulatory Commission, March 1982; Addendum,July 1983.

THE UNIVERSITY OF TEXAS TRIGA II NUCLEAR RESEARCH REACTORSAFETY ANALYSIS REPORT, APPENDIX 15.1I12/2011EXCERPTS FROM THE TEXAS EDUCATION CODE FOR THE GOVERNMENTOF THE UNIVERSITY OF TEXAS SYSTEM AND RULES 10501 AND 20201 FROMTHE RULES AND REGULATIONS OF THE BOARD OF REGENTS OF THEUNIVERSITY OF TEXAS SYSTEM FOR THE GOVERNMENT OF THEUNIVERSITY OF TEXAS SYSTEMEDUCATION CODETITLE 3. HIGHER EDUCATIONSUBTITLE C. THE UNIVERSITY OF TEXAS SYSTEMCHAPTER 67. THE UNIVERSITY OF TEXAS AT AUSTINSUBCHAPTER A. GENERAL PROVISIONSSec. 67.01. DEFINITIONS. In this chapter:(1) "University" means the University of Texas at Austin.(2) "Board" means the board of regents of The University of TexasSystem.Acts 1971, 62nd Leg., p. 3159, ch. 1024, art. 1, Sec. 1, eff. Sept. 1, 1971.Sec. 67.02. THE UNIVERSITY OF TEXAS AT AUSTIN. The University ofTexas at Austin is a coeducational institution of higher education within TheUniversity of Texas System. It is under the management and control of the board ofregents of The University of Texas System.Acts 1971, 62nd Leg., p. 3160, ch. 1024, art. 1, Sec. 1, eff. Sept. 1, 1971.EDUCATION CODETITLE 3. HIGHER EDUCATIONSUBTITLE C. THE UNIVERSITY OF TEXAS SYSTEMCHAPTER 65. ADMINISTRATION OF THE UNIVERSITY OF TEXAS SYSTEMPage 15.1-1 APPENDIX 15.1, STATUTES AND EXCERPTS REGARDING UT 12/2011SUBCHAPTER A. GENERAL PROVISIONSSec. 65.02. ORGANIZATION. (a) The University of Texas System iscomposed of the following institutions and entities:(1) The University of Texas at Arlington, including:(A) The University of Texas Institute of Urban Studies atArlington; and(B) The University of Texas School of Nursing at Arlington;(2) The University of Texas at Austin, including:(A) The University of Texas Marine Science Institute-(B) The University of Texas McDonald Observatory at MountLocke; and(C) The University of Texas School of Nursing at Austin;(3) The University of Texas at Dallas;(4) The University of Texas at El Paso, including The University ofTexas School of Nursing at El Paso;(5) The University of Texas of the Permian Basin;(6) The University of Texas at San Antonio, including the Universityof Texas Institute of Texan Cultures at San Antonio;(7) The University of Texas Southwestern Medical Center at Dallas,including:(A) The University of Texas Southwestern Medical School atDallas;(B) The University of Texas Southwestern Graduate School ofBiomedical Sciences at Dallas; and(C) The University of Texas Southwestern Allied HealthSciences School at Dallas;(8) The University of Texas Medical Branch at Galveston, including:(A) The University of Texas Medical School at Galveston.;(B) The University of Texas Graduate School of BiomedicalSciences at Galveston;(C) The University of Texas School of Allied Health Sciencesat Galveston;(D) The University of Texas Marine Biomedical Institute atGalveston;(E) The University of Texas Hospitals at Galveston; and(F) The University of Texas School of Nursing at Galveston;-2-Page 15.1-2 THE UNIVERSITY OF TEXAS TRIGA II NUCLEAR RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, APPENDIX 15.1.(9) The University of Texas Health Science Center at Houston,including:(A) The Universityof Texas Medical School at Houston;(B) The Universityof Texas Dental Branch at Houston;(C) The University of Texas Graduate School of BiomedicalSciences at Houston;(D) The University of Texas School of Health InformationSciences at Houston;(E) The University of Texas School of Public Health atHouston;(F) The University of Texas Speech and Hearing Institute atHouston; and(G) The University of Texas School of Nursing at Houston;(10) The University of Texas Health Science Center at San Antonio,including:(A) The University of Texas Medical School at San Antonio;(B) The University of Texas Dental School at San Antonio;(C) The University of Texas Graduate School of BiomedicalSciences at San Antonio;(D) The University of Texas School of Allied Health Sciencesat San Antonio; and(E) The University of Texas School of Nursing at San Antonio;(11) The University of Texas M. D. Anderson Cancer Center,including:(A) The University of Texas M. D. Anderson Hospital;(B) The University of Texas M. D. Anderson Tumor Institute;.and(C) The University of Texas M. D. Anderson Science Park;and(12) The University of Texas Health Science Center--South Texas,including The University of Texas Medical School--South Texas, if established underSubchapter N, Chapter 74.(b) The University of Texas System shall also be composed of such otherinstitutions and entities as from time to time may be assigned by specific legislativeact to the governance, control, jurisdiction, or management of The University ofTexas System.-3-Page 15.1-3

...APPENDIX 15.1, STATUTES AND EXCERPTS, REGARDING UT 1 .... 2/2:011 :Added by Acts 1973, 63rd Leg., p. 1186, ch. 435, Sec. 1, eff. Aug. 27, 1973.Amended by Acts 1989, 71st Leg., ch. 644, Sec. 2, eff. June 14,1989; Acts 2001,77th Leg., ch. 325, Sec. 1, eff. Sept. 1, 2001.Amended by:Acts 2009, 81st Leg., R.S., Ch. 1341, Sec. 5, eff. June 19, 2009.SUBCHAPTER B. ADMINISTRATIVE PROVISIONSSec. 65.11. BOARD OF REGENTS. The government of the universitysystem is vested in a board of nine regents appointed by the governor with theadvice and consent of the senate. The board may provide for the administration,organization, and names of the institutions and entities in The. University of TexasSystem in such a way as will achieve the maximum operating efficiency of suchinstitutions and entities, provided, however, that no institution or entity of TheUniversity of Texas System not authorized by specific legislative act to offer a four-year undergraduate program as of the effective date of this Act shall offer any suchfour-year undergraduate program without prior recommendation and approval by atwo-thirds vote of the Texas Higher Education Coordinating Board and a specific actof the Legislature.Acts 1971, 62nd Leg., p. 3144, ch. 1024, art. 1, Sec. 1, eff. Sept. 1,1971. Amendedby Acts 1973, 63rd Leg., p. 1188, ch. 435, Sec. 2, eff. Aug. 27, 1973; Acts 1989,71st Leg., ch. 644, Sec. 3, eff, June 14,1989.SUBCHAPTER C. POWERS AND DUTIES OF BOARDSec. 65.31. GENERAL POWERS AND DUTIES. (a) The board isauthorized and directed to govern, operate, support, and maintain each of thecomponent institutions that are now or may hereafter be included in a part of TheUniversity of Texas System.(b) The board is authorized to prescribe for each of the componentinstitutions courses and programs leading to such degrees as are customarilyoffered in outstanding American universities, and to award all such degrees. It is theintent of the legislature that such degrees shall include baccalaureate, master's, anddoctoral degrees, and their equivalents, but no new department, school, or degree-program shall be instituted without the prior approval of the Coordinating Board,Texas College and University System.-4-Page 15.1-4 THE UNIVERSITY OF TEXAS TRIGA II NUCLEAR RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, APPENDIX 15.1(c) The board has authority to promulgate and enforce such other rules andregulations for the operation, control, and management of the university system andthe component institutions thereof as the board may deem either necessary ordesirable. The board is specifically authorized and empowered to determine andprescribe the number of students that shall be admitted to any course, department,school, college, degree-program, or institution under its governance.(d) The board is specifically authorized to make joint appointments in thecomponent institutions under its governance. The salary of any person who receivessuch joint appointment shall be apportioned to the appointing institutions on thebasis of services rendered.(e) The board is specifically authorized, upon terms and conditionsacceptable to it, to accept, retain in depositories of its choosing, and administer gifts,grants, or donations of any kind, from any source, for use by the system or any ofthe component institutions of the system.(f) No component institution which is not authorized to offer a four-yearundergraduate program shall offer a four-year undergraduate program without thespecific authorization of the legislature.(g) The board by rule may delegate a power or duty of the board to acommittee, officer, employee, or other agent of the board.Acts 1971, 62nd Leg., p. 3145, ch. 1024, art. 1, Sec. 1, eff. Sept. 1,1971. Amendedby Acts 1971, 62nd Leg., p. 3360, ch. 1024, art. 2, Sec. 37, eff. Sept. 1, 1971; Acts1983, 68th Leg., p. 5010, ch. 900, Sec. 1, eff. Aug. 29, 1983; Acts 1995, 74th Leg.,ch. 213, Sec. 1, eff. May 23,1995.Rule 10501 Delegation to Act on Behalf of the Board (last amended 2/5/10)1. TitleDelegation to Act on Behalf of the Board2. Rule and RegulationSec. 1 Identification of Significant Contracts or Documents. Institutionalpresidents and executive officers at U. T. System Administrationare responsible for identifying contracts, agreements, and otherdocuments that are of such significance to require the priorapproval of the Board of Regents. Each such matter so identified-5-Page 15.1-5 APPENDIX 15.1, STATUTES AND EXCERPTS REGARDING UT _ 12/2011shall be presented to the Board by the Chancellor as an agenda ordocket item at a meeting of the Board.Sec. 2 Compliance with Special Instructions. All authority to execute anddeliver contracts, agreements, and other documents is subject tothese Rules and Regulations and compliance with all applicablelaws and special instructions or guidelines issued by theChancellor, an Executive Vice Chancellor, and/or the ViceChancellor and General Counsel. Special instructions or guidelinesby the Chancellor, an Executive Vice Chancellor, or the ViceChancellor and General Counsel may include without limitationinstructions concerning reporting requirements; standard clauses orprovisions; ratification or prior approval by the Board of Regents orthe appropriate Executive Vice Chancellor; review and approval bythe Office of General Counsel; and recordkeeping.Sec. 3 Contracts or Agreements Requiring Board Approval. The followingcontracts or agreements, including purchase orders or vouchersand binding letters of intent or memorandums of understanding,must be submitted to the Board for approval or authorization.3.1 Contracts Exceeding $1 Million. All contracts oragreements, with a total cost or monetary value to the U. T.System or any of the institutions of more than $1 million,unless exempted in Section 4 below. The total cost ormonetary value of the contract includes all potential contractextensions or renewals whether automatic or by operation ofadditional documentation. For purposes of this Rule, allcontracts with unspecified amounts of payments with a termof greater than four years are presumed to have a total valueof greater than $1 million.3.2 Contracts with Foreign Governments. Contracts oragreements of any kind or nature, regardless of dollaramount, with a foreign government or agencies thereof,except affiliation agreements and cooperative programagreements, material transfer agreements, sponsoredresearch agreements and licenses, or other conveyances ofintellectual property owned or controlled by the Board ofRegents prepared on an approved standard form orsatisfying the requirements set by the Office of the GeneralCounsel, or agreements or contracts necessary to protectthe exchange of confidential information or nonbinding lettersof intent or memorandums of understanding executed inadvance of definitive agreements each as reviewed andapproved by the Vice Chancellor and General Counsel.3.3. Contracts Involving Certain Uses of Institution Names,Trademarks, or Logos. Except as specifically allowed under-6-Page 15.1-6 THE UNIVERSITY OF TEXAS TRIGA 11 NUCLEAR RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, APPENDIX 15.1existing contracts entered into between the Board ofRegents and nonprofit entities supporting a U. T. Systeminstitution, agreements regardless of dollar amount that grantthe right to a non-U. T. entity to use the institutional name orrelated trademarks or logos in association with the provisionof a material service or in association with physicalimprovements located on property not owned or leased bythe contracting U. T. System institution.3.4 Contracts with Certain Officers. Agreements, regardless ofdollar amount, with the Chancellor, a president, a formerChancellor or president, an Executive Vice Chancellor, aVice Chancellor, the General Counsel to the Board, or theChief Audit Executive are subject to the applicable provisionsof Texas Education Code Section 51.948.3.5 Insurance Settlements.(a) Settlements in excess of $1 million must have theapproval of the Board.(b) Settlement claims from insurance on money andsecurities or fidelity bonds of up to $1 million shall beapproved by the Executive Vice Chancellor for BusinessAffairs.(c) If a loss is so extensive that partial payments in excessof $1 million are necessary, the Chancellor is delegatedauthority to execute all documents related to the partialpayment or adjustment. Final settlement of claimsin excess of $1 million will require approval by the Board.3.6 Settlement of Disputes. Settlements of any claim, disputeor litigation for an amount greater than $1 million requireapproval. The settlement may also be approved by theappropriate standing committee of the Board of Regents.The Vice Chancellor and General Counsel shall consult withthe institution's president and appropriate Executive ViceChancellor, or Vice Chancellor with regard to all settlementsin excess of $150,000 that will be paid out of institutionalfunds.Sec. 4 Contracts Not Requiring Board Approval. The following contractsor agreements, including purchase orders and vouchers, do notrequire prior approval by the Board of Regents regardless of thecontract amount.4.1 Construction Projects. Contracts, agreements, anddocuments relating to construction projects previously-7-Page 15.1-7 APPENDIX 15.1, STATUTES AND EXCERPTS REGARDING UT 12/2011approved by the Board of Regents in the CapitalImprovement Program and Capital Budget or Minor Projects.4.2 Construction Settlements. All settlement claims anddisputes relating to construction projects to the extentfunding for the project has been authorized.4.3 Intellectual Property. Legal documents, contracts, or grantproposals for sponsored research, including institutionalsupport grants, and licenses or other conveyances ofintellectual property owned or controlled by the Board ofRegents as outlined in Rule 90105 of these Rules.4.4 Replacements. Contracts or agreements for the purchaseof replacement equipment or licensing of replacementsoftware or services associated with the implementation ofthe software.4.5 Routine Supplies. Contracts or agreements for thepurchase of routinely purchased supplies.4.6 Group Purchases. Purchases made under a grouppurchasing program that follow all applicable statutory andregulatory standards for procurement.4.7 Approved Budget Items. Purchases of new equipment orlicensing of new software or services associated with theimplementation of the software, identified specifically in theinstitutional budget approved by the Board of Regents.4.8 Loans. Loans of institutional funds to certified nonprofithealth corporations, which loans have been approved asprovided in The University of Texas System AdministrationPolicy UTS166, Cash Management and Cash Hand/inoPolicy and The University of Texas System AdministrationPolicy UTS167. Banking Services Po/icy concerning depositsand loans.4.9 Certain Employment Agreements. Agreements withadministrators employed by the U. T. System or any of theinstitutions, so long as such agreements fully comply withthe requirements of Texas Education Code Section 51.948including the requirement to make a finding that theagreement is in the best interest of the U. T. System or anyof the institutions.4.10 Energy Resources. Contracts or agreements for utilityservices or energy resources and related services, if any,-8-Page 15.1-8 THE UNIVERSITY OF TEXAS TRIGA II NUCLEAR RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, APPENDIX 15.1which contracts or agreements have been approved inadvance by the Chancellor or the Chancellor's delegate.4.11 Library Materials. Contracts or agreements for the purchaseor license of library books and library materials.4.12 Athletic Employment Agreements. Contracts with athleticcoaches and athletic directors except those: with total annualcompensation of $250,000 or greater, as covered by Rule20204.4.13 Bowl Games. Contracts or agreements related topostseason bowl games, subject to a requirement that thecontract or agreement has been submitted to the ExecutiveVice Chancellor for Academic Affairs and is in a formacceptable to the Vice Chancellor and General Counsel.4.14 Property or Casualty Losses. Contracts or agreements witha cost or monetary value to the U. T. System or any of theinstitutions in excess of $1 million but not exceeding$10 million associated with or related to a property orcasualty loss that is expected to exceed $1 million may beapproved, executed, and delivered by the Chancellor. TheChancellor shall consult with the institutional president, ifapplicable.4.15 Health Operations. Contracts or agreements for theprocurement of routine services or the purchase or lease ofroutine medical equipment, required for the operation orsupport of a hospital or medical clinic, if the services orequipment were competitively procured.4.16 Increase in Board Approval Threshold. An institution's dollarthreshold specified in Section 3.1 may be increased to up to$5 million by the Vice Chancellor and General Counsel, afterconsultation with the General Counsel to the Board ofRegents, if it is determined that the institution has theexpertise to negotiate, review, and administer suchcontracts. Unless approved in advance by the ViceChancellor and General Counsel, any increase will not applyto contracts or agreements designated as Special ProcedureContracts by the Vice Chancellor and General Counsel.4.17 Group Employee Benefits. Contracts or agreements foruniform group employee benefits offered pursuant toChaoter 1601, Texas Insurance Code.Sec. 5 Signature Authority. The Board of Regents delegates to theChancellor or the president of an institution authority to execute and-9-Page 15.1-9 APPENDIX 15.1, STATUTES AND EXCERPTS REGARDING UT1 12/2011deliver on behalf of the Board contracts and agreements of anykind or nature, including without limitation licenses issued to theBoard or an institution. In addition to other primary delegates theBoard assigns in the Regents' Rules and Regulations, the Boardassigns the primary delegate for signature authority for thefollowing types of contracts.5.1 System Administration and Systemwide Contracts. TheBoard of Regents delegates to the Executive ViceChancellor for Business Affairs authority to execute anddeliver on behalf of the Board contracts or agreements:(a) affecting only System Administration,(b) binding two or more institutions of the U. T. System withthe concurrence of the institutions bound, or(c) having the potential to benefit more than one institutionof the U. T. System so long as participation is initiatedvoluntarily by the institution.5.2 Contracts Between or Among System Administration andInstitutions. The Board of Regents delegates to theExecutive Vice Chancellor for Business Affairs authority toexecute on behalf of the Board contracts or agreementsbetween or among System.Administration and institutions ofthe U. T. System for resources or services. Any suchcontract or agreement shall provide for the recovery of thecost of services and resources furnished.5.3 Contracts with System Administration or Between or AmongInstitutions. The Board of Regents delegates to thepresident of an institution authority to execute on behalf ofthe Board contracts or agreements with SystemAdministration or between or among institutions of the U. T.System for resources or services. Any such contract oragreement shall provide for the recovery of the cost ofservices and resources furnished.5.4 Contracts for Legal Services and Filing of Litigation. TheBoard of Regents delegates to the Vice Chancellor andGeneral Counsel authority to execute and deliver on behalfof the Board contracts for legal services and such otherservices as may be necessary or desirable in connectionwith the settlement or litigation of a dispute or claim afterobtaining approvals as may be required by law. Litigation tobe instituted under these contracts on behalf of the Board,System Administration, or an institution of U. T. System must-10-Page 15.1-10 THE UNIVERSITY OF TEXAS TRIGA II NUCLEAR RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, APPENDIX 15.1have the prior approval of the Vice Chancellorand GeneralCounsel.5.5 Settlements of Disputes. Except as provided in Section 5.6below, the Board of Regents delegates to the ViceChancellor and General Counsel authority to execute anddeliver on behalf of the Board agreements settling any claim,dispute, or litigation. The Vice Chancellor and GeneralCounsel shall consult with the institutional president and theappropriate Executive Vice Chancellor or Chancellor withregard to all settlements greater than $150,000 that will bepaid out of institutional funds. Settlements greater than$1,000,000 will require, the approval of the Board as outlinedin Section 3.5 above. The Vice Chancellor and GeneralCounsel shall consult with the Office of External Relationswith respect to settlement of will contests and other mattersrelating to gifts and bequests administered by that Office.5.6 Construction Settlements. The Board of Regents delegatesauthority to execute all documents necessary or desirable tosettle claims and disputes relating to construction projects tothe System Administration or institution official designated inthe construction contract to the extent funding for the projecthas been authorized.5.7 Assurance of Authority to Act. The officer or employeeexecuting any document on behalf of the Board .of Regentsshall be responsible for assuring that he or she has authorityto act on behalf of the Board and that such authority isexercised in compliance with applicable conditions andrestrictions. Documents executed on behalf of the Boardpursuant to authority granted under these Rules andRegulations shall not require further certification orattestation.5.8 Institutional Agreements for Dual Credit. The Board ofRegents delegates the authority to approve and executedual credit partnership agreements for the academicinstitutions to the Executive Vice Chancellor for AcademicAffairs.Sec. 6 Delegation Process. The primary delegate identified in theseRules and Regulations or in an official Board action may furtherdelegate his or her delegated authority to a secondary delegateunless otherwise specified. Any such further delegation of authoritymust be made in writing and the primary delegate shallpermanently maintain, or cause to be maintained, evidence of allsuch delegations. A secondary delegate of the primary delegatemay not further delegate such authority.-11 -Page 15.1-11 APPENDIX 15.1, STATUTES AND EXCERPTS REGARDING UT 12/20116.1 Delegate's Responsibilities. The primary delegate identifiedin these Rules and Regulations as authorized to execute anddeliver on behalf of the Board of Regents various types ofcontracts, agreements, and documents shall maintain, orcause to be maintained, necessary and proper records withregard to all contracts, agreements, and documentsexecuted and delivered pursuant to such delegatedauthority, in accordance with any applicable recordsretention schedule or policy adopted by the Board, the U. T.System Administration, or the institution.Sec. 7 Actions of the Board as Trustee. Authority delegated by the Boardof Regents in these Rules and Regulations includes actions thatmay be taken by the Board in its capacity as trustee of any trust tothe extent such delegation is permitted by law.Sec. 8 Power to Authorize Expenditures. No expenditure out of fundsunder control of the Board shall be made and no debt or obligationshall be incurred and no promise shall be made in the name of theSystem or any of the institutions or of the Board of Regents by anymember of the respective staffs of the U. T. System or any of theinstitutions except:8.1 In accordance with general or special budgetaryapportionments authorized in advance by the Board ofRegents and entered in its minutes; or8.2 In accordance with authority specifically vested by theBoard of Regents in a committee of the Board; or8.3 In accordance with authority to act for the Board of Regentswhen it is not in session, specifically vested by these Rulesand Regulations or by special action of the Board.Sec. 9 Power to Establish Policies. No employee of the U. T. System orany of the institutions, as an individual or as a member of anyassociation or agency, has the power to bind the System or any ofthe institutions unless such power has been officially conferred inadvance by the Board of Regents. Any action which attempts tochange the policies or otherwise bind the System or any of theinstitutions, taken by any individual or any association or agency,shall be of no effect whatsoever until the proposed action has beenapproved by the president of an institution concerned, if any, theappropriate Executive Vice Chancellor, and the Chancellor, andratified by the Board.Sec. 10 Exceptions. This Rule does not apply to any of the following:-12-Page 15.1-12 THE UNIVERSITY OF TEXAS TRIGA 11 NUCLEAR RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, APPENDIX 15.110.1 UTIMCO. Management of assets by UTIMCO, which isgoverned by contract and the provisions of Rule 70101,70201, 70202, and 70401 ofthese Rules and Regulations.10.2 Acceptance of Gifts. The acceptance, processing, oradministration of gifts and bequests, which actions aregoverned by Rule 60101, 60103, 70101, and 70301 of theseRules and Regulations and applicable policies of the Boardof Regents.10.3 Statutory. Any power, duty, or responsibility that the Boardhas no legal authority to delegate, including any action thatthe Texas Constitution requires be taken by the Board ofRegents.3. DefinitionsSettlement -the amount of the settlement shall mean the amount that mightbe reasonably expected to be recoverable by the U. T. System or any of theinstitutions but not received pursuant to the settlement or, in the case of aclaim against the U. T. System, the total settlement amount to be paid by theU. T. System.Group Purchasing Program -for purposes of this Rule, a purchasing programestablished by (1) a state agency that is authorized by law to procure goodsand services for other state agencies, such as the Texas Procurement andSupport Services Division of the Texas Comptroller of Public Accounts andthe Texas Department of Information Resources, or any successor agencies,respectively; or (2) a group purchasing organization in which the institutionparticipates, such as Novation, Premier, Western States Contracting Alliance,and U.S. Communities Government Purchasing Alliance.4. Relevant Federal and State StatutesTexas Education Code Section 51.928(b) -Written Contracts or AgreementsBetween Certain InstitutionsTexas Education Code Section 51.948- Restrictions on Contracts withAdministratorsTexas Education Code Section 65.31 (Q) -Delegation by the BoardTexas Government Code Section 618.001 -Uniform Facsimile Signature ofPublic Officials ActTexas Government Code Sections 669.001 -669.004 -Restrictions onCertain Actions Involving Executive Head of State Agency-13-Page 15.1-13 APPENDIX 15.1, STATUTES AND EXCERPTS REGARDING UT 1 12/2011Texas Insurance Code, Chapter 1601 -Uniform Insurance Benefits Act forEmployees of The University of Texas System and The Texas A&M UniversitySystem6. Relevant System Policies, Procedures, and FormsThe University of Texas System Administration Policy UTS166. CashManacqement and Cash Handling PolicyThe University of Texas System Administration Policy UTS 167, BankingServices PolicyRegents' Rules and Regulations, Rule 20204 -Determining andDocumenting the Reasonableness of CompensationRegents' Rules and Regulations, Rule 60101 -Acceptance andAdministration of GiftsRegents' Rules and Regulations, Rule 60103 -Guidelines for Acceptance ofGifts of Real PropertyRegents' Rules and Regulations, Rule 70101 -Authority to Accept andManage AssetsRegents' Rules and Regulations, Rule 70201 -Investment PoliciesRegents' Rules and Regulations, Rule 70202 -Interest Rate Swap PolicyRegents' Rules and Regulations, Rule 70401 -Oversight Responsibilities forUTIMCOLitigation Approval Request FormSpecial Procedure Contracts6. Who Should KnowAdministrators7. System Administration Office(s) Responsible for RuleOffice of the Board of Regents8. Dates Approved or AmendedFebruary 5, 2010November 12, 2009August 20, 2009-14-Page 15.1-14 THE UNIVERSITY OF TEXAS TRIGA II NUCLEAR RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, APPENDIX 15.1Editorial amendment to add Subsection 4.17 (Group Employee Benefits)back into the Rules made August 6, 2009Editorial amendment to Number 4 made January 5, 2009November 13, 2008May 15, 2008Editorial amendment to Sec. 3.3 made March 17, 2008Editorial amendment to Number 3 made January 28, 2008May 10, 2007February 8, 2007May 12, 2005December 10, 20049. Contact InformationQuestions or comments regarding this rule should be directed to:0 borutsystem.eduRule 20201 Presidents (last amended 8/23/07)1. TitlePresidents2. Rule and RegulationSec. 4 Duties and Responsibilities. Within the policies and regulations ofthe Board of Regents and under the supervision and direction ofthe appropriate Executive Vice Chancellor, the president hasgeneral authority and responsibility for the administration of thatinstitution. Specifically, the president is expected, with theappropriate participation of the staff, to:4.1 Develop and administer plans and policies for the program,organization, and operation of the institution.4.2 Interpret the System policy to the staff, and interpret theinstitution's programs and needs to the SystemAdministration and to the public.4.3 Develop and administer policies relating to students, andwhere applicable, to the proper management of services topatients.4.4 Recommend appropriate operating budgets and superviseexpenditures under approved budgets.-15-Page 15.1-15 APPENDIX 15.1, STATUTES AND EXCERPTS REGARDING UT12/2011I4.5 Appoint all members of the faculty and staff, except asprovided in Rule 31007, concerning the award of tenure, andmaintain efficient personnel programs.4.6 Ensure efficient management of business affairs andphysical property; and recommend additions and alterationsto the physical plant.4.7 Serve as presiding officer at official meetings of faculty andstaff of the institution, and as ex officio member of eachcollege or school faculty (if any) within the institution.4.8 Appoint, or establish procedures for the appointment of, allfaculty, staff, and student committees.4.9 Cause to be prepared and submitted to the appropriateExecutive Vice Chancellor and the Vice Chancellor andGeneral Counsel for approval, the rules and regulations forthe governance of the institution and any relatedamendments. Such rules and regulations shall constitute theHandbook of Operating Procedures for that institution. Anyrule or regulation in the institutional Handbook of OperatingProcedures that is in conflict with any rule or regulation in theRegents' Rules and Regulations is null and void and has noeffect.(a) Input from the faculty, staff, and student governancebodies for the institution will be sought for all significantchanges to an institution's Handbook of OperatingProcedures. The institutional Handbook of OperatingProcedures will include a policy for obtaining this inputthat is in accordance with a model policy developed bythe Office of General Counsel.(b) Sections of the Handbook of Operating Procedures thatpertain to the areas of faculty responsibility as defined inRegents' Rules and Regulations, Rule 40101 titledFaculty Role in Educational Policy Formulation will beexplicitly designated in the Handbook of OperatingProcedures. The president, with the faculty governancebody of the. campus, shall develop procedures to assureformal review by the faculty governance body beforesuch sections are submitted for approval. The formalreview should be.done within a reasonable timeframe (60days or less).4.10 Assume initiative in developing long-range plans for theprogram and physical facilities of the institUtion.-16-Page 15.1-16 THE UNIVERSITY OF TEXAS TRIGA I I NUCLEAR RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, APPENDIX 15.14.11 Assume active leadership in developing private fund supportfor the institution in accordance with policies and proceduresestablished in the Regents' Rules and Regulations.4.12 Develop and implement plans and policies to ensure that theinstitution remains in compliance with any accreditationrequirements appropriate to the institution or its programs,including, for the health institutions and those academicinstitutions with student health services, the accreditation ofhospitals, clinics, and patient-care facilities.4.13 The president of each general academic institution of TheUniversity of Texas System that engages in intercollegiateathletic activities shall ensure that necessary rules andregulations are made so as to comply with the currentGeneral Appropnations Act.3. DefinitionsNone4. Relevant Federal and State StatutesCurrent General Appropriadons Act5. Relevant System Policies, Procedures, and FormsModel Policy -Handbook of Operatinq Procedures (HOP) AmendmentApproval Process6. Who Should KnowAdministratorsFacultyStaffStudents7. System Administration Office(s) Responsible for RuleOffice of Academic AffairsOffice of Health Affairs8. Dates Approved or AmendedAugust 23, 2007August 10, 2006May 11,2006-17-Page 15.1-17 APPENDIX 15.1, STATUTES AND EXCERPTS REGARDING UT 12/2011March 10, 2005December 10, 20049. Contact InformationQuestions or comments regarding this rule should be directed to:* bortbutsgstem.edu-18-Page 15.1-18 UNIVERSITY OF TEXAS TRIGA II NUCLEAR RESEARCH REACTORSAFETY ANALYSIS REPORT, APPENDIX 15.2I12/2011FIVE-YEAR OPERATING COST ESTIMATEInitial year expenses in relevant categories are summarized from monthly expense records. Projectedexpenses are based on an average 3% rate of inflation.NOTE[1]: Return on UT investment portfolio, consequently fluctuates UNIVERSITY OF TEXAS TRIGA II NUCLEAR RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, APPENDIX 15.30;1`1(iTE IEiESDN AI .FIIi FINANCIAL OFFICER,4 I TH E LINIVERSITY OF TEXAS AT AUSTINA.0. Box 8179 *A~wti,, 7'exas 78713-8179(512)471-1422 ( 512471-7742December 1, 2011Mr. A. Jason LisingProject ManagerDivision of Policy and Rule MakingResearch and Test Reactor Licensing BranchWashington, DCRE: License R-129Docket 50-602Dear Mr. Lising:This concerns the ultimate decommissioning of the University of Texas TRIGA II nuclearresearch reactor, currently licensed for operation by the University until January 17, 2012.Pursuant to the Code of Federal regulations, title 10, Part 50, this is to assure that theUniversity an entity of the State of Texas will obtain funds for decommissioning when it isnecessary.It is our intention to propose renewal of the current facility operating license. Nevertheless,whenever a decision to decommission the facility is made, the University will requestlegislative appropriation of funds sufficiently in advance of decommissioning to prevent delayof required activities,As Chief Financial Officer for the University, I have the authority to sign this statement ofintent.ice PresidHegartyi~fce KPresident and Chief Financial Officerc: Dr. Juan M. Sanchez, UT Austin, VP for ResearchMr. Paul Michael Whaley, UT Austin, NETL UNIVERSITY OF TEXAS TRIGA II NUCLEAR RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, APPENDIX 15.4DECOMMISSIONING COST ESTIMATENUREG/CR-1576 analyzes data from decommissioning of a 0.1 MW university reactor (OSU/AGN-201), a0.01 university facility (NCSUR-3), a 0.2 MW (1 MW forced flow) commercial facility (B&W, LPR), a 250kW Army facility (DORF), and a 5 MW heavy water moderated DOE facility (ALRR).Table 15.x, Summary of NUREG/CR-1576 ValuesBASE COSTFACILITY POWER OWNER YE ($00YEAR ($1000)OSU/AGN-2011"'1 0.1 W Oregon State University 1980 10NCSUR 10 kW NC State University -- 33/partLPR 200 kW/1 MW Babcock & Wilcox 1982 86DORF 250 kW A.S. Army 1980 336ALRR 5 MW Department of Energy 1981 4,292The ALRR was a more complex installation than the UT TRIGA, and would not be expected to have thecomparable labor demands in decommissioning. The cost for decommissioning the UT reactor istherefore expected to be biased more towards the LPR and DORF; DORF decommissioning costs aretherefore used for comparison of total costs, distributed according to NUREG recommended disposalcost estimation:C,98,,adjj,,ed = (X)- {(L) * (L,.) + (R) .(R,) + (0) .(Oa)}Where:C1981,adjusted is the current value based on the 19981 valuesL is the labor cost as a fraction of total decommissioning costsLa is the adjustment of labor costs from 1981 valuesR is the radwaste burial costs as a fraction of the total decommissioning costsRa is the adjustment to account for changes between 1981 and the current year0 is the factor of all other coasts as a fraction of the total decommissioning costsOa is the adjustment to account for changes between 1981 and the current yearThe average cost of labor is 44.72% of the total cost. There are two outliers in the data, 64% for a verylow power reactor (where the remainder of the costs were disproportionally low), and a universityreactor that minimized costs with student labor. With these outliers removed, the average value is43.9% with a deviation of 1.9% from the aggregate average indicating the average value may berepresentative of the 1.1 MW UT TRIGA.Page 15.4-1 APPENDIX 15.4, DECOMMISSIONING COST ESTIMATE .12/2011The average of the unspecified ("other") costs is 50.7% of the total cost. The influence of the outliersadds some bias but the average excluding the outliers is 52.0% (a deviation of about 2% from theaggregate), indicating the average value may be representative of the 1.1 MW UT TRIGA.The cost of waste disposal ranges from 1% to 9.4%, probably because of the large variation in thevolume of waste in the cases examined. The volume of waste ranges from 1157 m3 for the largestfacility to a negligible quantity for the smallest. The average fraction for waste disposal across all casesis 4.6%, with 4.2% excluding the outliers. The two highest power levels have fractions significantlydifferent, 3.9% for the 5 MW kW facility and 1.6% for the 250 kW facility, suggesting the average maynot be as representative of the 1.1 MW UT TRIGA; the 4.2% value is used.The three individual fractions are normalized to get a valid distribution, so the fractions are (L) 44.8%,(0) 50.9% and (R) 4.2% for labor, non-specified and rad-waste disposal costs respectively.The Consumer Price Index calculator (http://www.bls.gov/data/inflation calculator.htm) indicates thatthe current value for the original $336,000 decommissioning cost is $836,936. Assuming an annual rateof 3% inflation, the decommissioning cost at the end of the new 20 year license will be $888,609.Page 15.4-2 THE UNIVERSITY OF TEXAS TRIGA II NUCLEAR RESEARCH REACTORSAFETY ANALYSIS REPORT, APPENDIX 15.512/2011,.STANDARD 4ESARCHSUo§-CXTRACT NO IooP. W026"'REACTOR.. L ASSISTANSCEAND FUELELEMENTS"Battelle Energy (BEA).2525 Fremont Avenue:. 0; Box 1625ID..&3415-39OSubcoitfractor: Contractor's Procurement RepresentativeTne Texasat AusatiP.O. Box 7.7,26 Lynda KellerAustii.,'TX 7837.31!726 Subcontract AdministratorTo. _Sus.m.yatt .Sedwk 208r526-5597Pl:. Sean O'0ilUy" 208-526-5780Lynda,Keller.inl.govPeriod 6f P"efokmainaee- Award Ainowit:I..li2OS4 ..... o0Auguntl3.L200 _-A. ... .......' -... .bloductlowsThis.is a and develop-ent work, notrelated tonuclea, .Cemiai, biologicall dr-fad.logica1 weapos:ofmass destruction ortheproduotionofspecial nuclear material-for usein- weaona-of mass detruction. This Sibcontract.isbetween'BattelleEneryA~igu, LC A~Con~ctr) nivrit~y. of Texas at Austin (Subcontractor). The$ubcontract is aiuedmunder-P-ime Contra& No. DE_,AC7-05lD14517 betweenuthe Contractor andthe.Uhited"$taRts )ep artmeant y (DOE) for the r. nanegeent and operation offthe IdahoNAtioftal Laboratory (INI).The parties agree etre* sefa 'e obligtionsj icwrdance with the tems and. conditionsgfethe %l.Werd"Pi.visios A Wb l ather doumnents attached or incorporated by reference,which togethe, constitute.the:entiroSubcontract aid' s du'ersedellprio discssons rgotiotons,rýepresentations, and Ao~enents..RATTELEE 4EMRGV ALAMNCE, LC-.(BEA)Na.me:. Lynida Keller....Title:. Sul'ontracf Am istratorDat. 44UNIVERSITY OF TEXAS AT AUSTINNarie:. Jenette Holmes .Title:: Associate DirectorDate: Office Of. Sponsored rPojecDEC 112009Page 15.5-1 APPENDIX 15.5, FUELS ASSISTANCE CONTRACT -12/2011Mate oa Altace LBeimm A111an LLCSdwionreh suonatcl No.m0007206PegE2_of 12.SCHEDULE- OF ARTICLES1. Statement of WorkThe Subcontr acor shall fitrnishothe following intc*ordtcewith.the.require nts, tenms4od condiiionespeditied orrernditlsubnfa.Pýovide for ulizafion ofth. reactor owned by .te Subcontrdctorin aprogram-of educationand hraig'o.f~students:ino nuclearfscience and egineering, and for faculty and studentresearch. The Subcontract provides. for the continued possssion and use of Department ofEnetgli (D.OE)-owned .nmclear Imaterials, enriched uranium, in reactor fuel withoutincreentalcharge of use, burtn-p, and reprocessing while.used for reseaach, eduoption andtrliainin purposes.TheDOB-owned nuclear materials were rigindlly provided to Subcoatractor underSubcntract~b No.0-110G742-002- The nuclear mateiials- will noW reside with thisSubcontract No.,00d78206.The Subcontractor'.'rincipal Investigator assiped to this work is Sean O'Kely. ThePrincipalInvestigatot shall nat be replacedor reassigned w.thout the advance written approval of the'Contactor's Subcontract A ia2. stRepor-s dDFata Re7uitMeuestsL Disribuion othe OWNR~or~lL~iilear ate l Tranaction Report;shall ihdludeJS G/MM Copies of DOE/NRC Forms 742, Material Balanceand:142C, Physical Inventory Listing, shall be sent to the Contractro* ahrmrncleerhnateal mangeent and accountability,2. Annually, i conjunction vith subimittal ofthe Material Balance Repoyr: andPlihysidel Inventory Listing the requtired to submit.irmation wsoth ath. Contraztor.can rmeetDOE'requirenents~frjuDOE Order 5:660.ItB, Manag~remetof f.NuclearN Mie .ls requited fo notifyýtheC.Qotractor. ofthe following:Fulusage in grams Uranium 235 and auabenofire tiems.2.2. CUrrnt .in to of uradlated telelements in storage.2;3.. C Qriven prY f el elements in ore.Z4. pwreanHv.enoory of useabla irradated the) eldmrents outside o .ftoroý215 nh e shipmeE-6. Thjected'fanj needs trihte naflivejyers2:,7. infventQry of ia other .nuc cleanmaj-ialitems.under daho Fidd.Office(DOE-4WYasganedprojet.identification number; le., those project numbersbegin2ng with theo haraotor. "]".28.& CuretSubcntrcto p/oint-o f-contact nmaterial accoountability.Page 15.5-2 THE UNIVERSITY OF TEXAS TRIGA II NUCLEAR RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, APPENDIX 15.5Hrattelte'Encrs AttwnceL1WStbndantRe661tch WOW tt W,0075206Pge 3.pf .12b. Final Repo t.The Subcotractor shal funish within.6.iMoths after the. shipment orall remainingma teral under this Subonnoot,.areport indicaing the amount otmateial returned andwhether:additional material fequests are planned.3. aiod of Pe-brulraneeThe work dsaiib~ed in. theStatenent of Workis efftcthe August 1, 200S%.and shall be completed onorbefote.August 31,2013..4. React.r iel .Spedatl Froduieus8.a. Tidctleaau special fuclearmaterials-loaned to the Sdbnotractor ýunderthiis8.ubcontract -sheatat all times be and remain the Unitd. States.Government.b. TheCnrco ilntchreteSbolrc for naateriala(l)- consumedi hoperation of expation of thisSdbcontu_.ct and (2) Mrerioc ing.subsqe1attto the'uhidiate -return of teepecial nuclears materia,.c, As a Nuclear Li -.epseq, the Subcontractor shal, 'Inadditiontocom-plying with .0CFR 73.7 :and73.7 be.respomnsible for performing (orcontracting others the actions necessary for compliance with the Ordc.for.tSeeity Cmpensadtot Measures on theTrasportation of Spentthen 100 gream s-.podified by the NRC from time to time, Ifrequitet arrgeiteats famed esctathetesppnsibility cftSe: Subotacoro4. If the Sub.contrctor deis to retuwn-MaWtR provideunerbi Sibnintrsea hSubcontractor shall. kbmita reqiuest to the Contracthr, preferably withia.1.months.butno later than6 mont f-om tihe time whic'the. Subcontrigtor desires-toretartheSubotecod rest return. The-Contractor will provide requiremnin s fordocuran-tion ad-inuiore for retmniug'the niairit. At thS Conh#trtWrs 60piothe ~ ~ d Co 1Wl~t~~e adhippn- 'c taint and prov ifus 'irectly to aCarrier,;.dund'era Separate.P ed-t to.negotin4td cost linitatiofii,.the Confatb'r witlle rebus the ucciracofronlercia shiipping cdnftififrntoa use dfaC.kr ancothet costs fo% .activitie.incident to the .shipmet odf-ihimntetiaf, The:Sqbcontraco:has. o:re.poaoibiliy'foi rectip.ta a. *oqrpsu'iofs.dscrix.terxal- The.Subcontractor't obligationdis to return matedialindthp form defined, as.affected by the adiviii listed above.I Aicle 1.e. E:xcept aaotherwis provided heri,.the Suboontractour-is responsible for and will paythe AY. stwg 8impose by theCqnrtraetorfor material delivred-to-the-Subontractor and not: utimately retumdtc to.the: Contiactor..N~otwitbstanduataa o -epoi~sion of ibis Sidicontrad, the Contractor or theG .o vemment responsiblerfor orhave any-oblgatiou.to the .for d aitati isioi (-&D) Of Any 6f the Subc6ntractorsfacilities.1lT- TheSubconittpdt..is.responsible-fot the matagemeat. accountabillty.aand -ontrol~ofDE-o.wned nuclear mataeriainiits possession. Ncletarmaterialtiipplied utdertliisSubontactbyt DQ shllanupy'It the fbllowingtc~quhlrnmats:Page 15.5-3 APPENDIX 15.5, FUELS ASSISTANCE CONTRACT12 0112/2011I.-Standard.Sitch SubconuuctNo. 00078206l.. Nuclear material jitscuxted for with ..w0.digi alphanumeic, budget andportingproje:ideifidationn..be whichis assigned-ancl controlled byIdaho. Ooperations (NE-ID)> The Stb-ontractor is not allowed to:make changes.to this number..2. The pruject identification number must be recorded in the Project Numbet. field0ttti.he DOE/NRC Foiina741, 'Nuclear Matenial Transaction Report", involvingany activity, e.g., receipts, removal and adjustments (Reference NUREG BR-0006, "Instructions.for Completing Nuclear Material Transaction Reports");and'DOE/NRCForm. 742C, "Physical Inventory Listing" (Reffrence NUJREG9R-0007, fo* thebPreparation and. Distribution of Material Status'eports").i. In the efvnt -the terms and condi#ons :f this Subcontract ar e not mae t wiiRCrufl and rqgulatobs,: the NRCrhequirements willtake precedence.S. SubcontractAdminhstrationa, TheContractor's SubhontratAdmiristrator for this Sucontract-is LyndaKelcr. TheSubp ntrat41Adin.stratr.is t-e oriyp-rson.authorized:to make.chages in-herequi'wnmentsofthis Sibcgniract-crnenosld.catins. totis. ubcontract,-includgdfiatis orind~pii eh Stteen0o Workan~tdth Schedule. The Siteontractor"-lldirect &noficetafdteqtueswts.-for approval required-by this Subcontract to tidAny notic#s by thSubcona frio the Contractor to theSubcontrctor:shal be issve by the Subcontract Administrator.b. The Contractor's Teclnical Representative for this Subcontract is D. Morrell. The.Technical Ropresentativexis the person designated to monitori the SuTbcontract work and to.iiibterprt and clarify he-technical dr.quir ents of the Statemejit of Worl- The Technical.'eantative "inot. au i to make changes to the work or mocdify thi Subcontract.c40T;eCovtntrctos.Ma rials%6mag Wt.a~nd-Accoountability representactive fortis:iiieonnrcis M.. Wilklns 'Progress repor*as 2a. shall beproided.t.the.eprqresena *tive accrdnlo the establishedby DOE and NRC.ThMe trator foe this Subcontract is.Dr, Susan WyattSdwick.:6,. SpplierPerformance. Evahaflonr ystem: C4jntrtkrb rva avakatetalbconsrtmatot performna.c in.-acordance with the SPES. The Subcontractorshall. be;.Tnolly evaluated no'less th quater as-ipoiicable, and.upon comple.ion of the WOrk. Atiiimiumtcd6 .of.80.pointo out InformatioRconcerning the SPES I's avaflable for review at: huttp:wvw.'oinl.ovloroctrernentlfonmsaihtml. Select..I1l. Supplier ManagemrentProgram,7.. Lower-tier Subcontractors$ub,coniactor dshl ].nt isbcontract pru-rýflr ce of !y...p. tion.of.the dWork being: 0 rfoed .at.theINL .vhu .ý .a4yandeA wfttex4ppruval ikf Contractor1(exoluding material deliveries). Lower-tierPage 15.5-4 THE UNIVERSITY OF TEXAS TRIGA II NUCLEAR RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, APPENDIX 15.5*tand ,d hsr: Subontract No; 00078206PageS of 12.ubcohtracts and ptrchase ordelt must inludeprovisions-to..secure all.rights andzremedies ofContractor and the-Co.vermnentprovided under this Subcohtract,..and inust impose upon the lower-tier suklbntqactor altof the gen-eal 4uties: and. obligations.required to fuill this Subcontract.Subontractoris responsible fbrtlh performanc.and oversight ofall lower-tier subcontractors8. " Ordet of PrecedenceIn the of ay irikonsistency between provisions oftihis Subcontract, the inconsistency shall beresolveýby ghiing-pre.dence as f6llows:. (a> Subcontracit hange ddcuments, if any, (b)Subcontract, (c) :Speoifcations o $State nt o f Work, (d)- Geneal.Provisions, and (e) otherpr.sibns thiftsiSubcontractwhtliherinoorporataedby refereucýorotherwise. However,Sub cnnt notify-,Jontracto: prior to ,erfoxo~w~o~k based oh.iesolution of any-incopsistenc, ý.'ha bj 6dr'ofprecedence setforth. hereiný9. Appliable DocumeutsThe following documents are applicable to Subcontrtcta. 10 CFR 73,37 ad b. Order for Safeguads and Security:Compep.etory.ealt7,utaents on the Tranportationofspent Nue'l*~ Fuel;..c. mfNC'Fonn741, N'ea Maei r al& Transaction Rtpbn.DOE-NRC ftrm 742,. Material Balance Report.e. DOE/NRC Form:7420, Ph4ysical inventory Listtsg.f. N1U.EG BR-0006, Instructionsfor .Completing Nuclear. Material Transaction reports-* ; N .G BIR-W.0.7,u"bstmctions'for the Pteparation and Ditrisbation of Material StatusReports.h D DO¶.2drw5660. %Mngmn fNula aeisPage 15.5-5 APPENDIX 15.5, FUELS ASSISTANCE CONTRACT 12/2011.* (8tiad~sc SucnmtN.478206:CENFAALý PROVISIONSCLAUSE I- A, The Subcontractor shall closely coordiate with the Contractor's Technical Representative'regarding any proposed. sientific, technical 4r. professional publication of the results of the. workpebrined ot any data developed under this Subcontract. 'The Subcontractor shall provide theContrator An to revjeW antyproposed-manuscripts in. w0le or in part,'thestf the .ed or anydata-developed.under this Subcontract at leetfforty-five.(45)days :prior su imssion for publicatipnm .The Con.trato .wil review the proposedpublication .Ad piovide comments. A. response shall: be. piovided to the Subcontractor -within.forty-five (45) days; otherwise, the Subcontractor may asare that the Contractor has nocomments. Subjgct to the. of Clause 9, the Subcontractor agrees to address anyconerns or isues. identified by the lContractor prior to submission for publication.B Subcontractor ni6.y Contratdor and of the work asappMOP~ito.A. Tle'-Stbcatractor hall -nmcdia+/-ely notify the.Contractorts. SubcontractA4ministrator inw *ig*o. (1) any acot., including any proc beoe an adrin'strativ6 agency, filed agaii~t theSubcontractor arising out of ýthe performance ofthis-Subcontract, and (2) any claim against theSubcontractor, the cost and-expense.of which is alloW1abl6uder the.terms ofthi SubcontractB, ay time durig the .erfrmauce of this Subcontact the Subcontractor becobaeg aware of,alty CdhIrCUtaiinC whic ay.jeopiadze its *-"&erma f aV or any p~rtibn df the Subcontract,it shall lnneadtly. oto:y Athe Contractor's. Siubcontract- Administrator in- writing of suchc .cumstAnces;, ards.heSnbhb.imi dia'ake whateveactin is §esar to, cre si& defectThe assign tfis Subcontract to.te. Goverment or its designee(s). Except as to.assignment. o.f payment due, the¢ Subcontractor. shall :Wave, no:. right o0 assigh. or. m.o1tgage 'thisoAr any part'f ithu.. t ih pror .i0tte-._aprovl of the Contractor's Subcontract.Adininistrý tr,. oeptýýrstb= rot s aleady, idenifiediir-te tSecontmctor-s proposal.,CLAIUSE4- DISPUTES.A.: Informnal Res~lutiqn1W parties to. a, dispute shltm attempt to resolve it. in gpod fait by dirn, infousfal,All. egotiation: shall Pending resolitioa of the dispute, the.Subcontractor .shall proceeddigeon tly. With the performance of this Subontract, in..cordane..withits. ternis and.conditi~ns.2..1W-patties, lpon mutual agteme, may seek, the.. assistance of a.nejtral third.pa'ty a1. anytirIe "but th-ey must seeOk sc assistia.ce n jaiter. than 120 ds after the date of theCoiiAmwtor .recewp , la. 'The.requiremet t see t .eas.t .qof..nem thirdparty' may be- walved or- Modified. only with -the consemt of all parties. T~he paifict teaPage 15.5-6 THE UNIVERSITY OF TEXAS TRIGA II NUCLEAR RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, APPENDIX 15.5S-andid Ricar"ch Subea inut aNo.:00078206fI'request .the-istance of an.-ctablished Ombudsman Program, where availabIe, or hire .amutually.agreeableiiediator, or sk:the DOE Office.of Disp06.Resolution to assistlhem inseleecing a mutually agreeable mediator. The cost of mediation shall be shared. equally byboth parties. If requested by both p4ries, the.neutral third party may ofer-e non-bindingopi0Ion as to .setleradit. All..discussions with the neutmrL third party shall beconfidential.3. In the eovent the parties are -unable, to resjv._ tiiedispute. by using a .ieutral third party orwpivyethe i'emq nt to. s3k such assistance, theContractor will issue a written decision-on.:tha lanB. Formal leSoMon1. If a,.dispte has tiot.been resolved..by informal:resolution, itmay be submitted to. bindingupon. agrepment of both partis, b.- and ini accordance with the CommercialArbitratio .Rides of theAmeiian.Axbitration Association (AAA).. If arbitration is agreed to'by boti- pares, such. 4ceision.. is irievocable:..tl. th outcoe of te..rbitration shall be21. Bapar.y to th:e-adbitrtn shall pay its.pro rata-shar6eof the arbitration fees,. notincludingcounsel .fes orwitness:fees or othevr ep.mes incurred y. the party .for its on benentfi*3. udnent.:on the ..award. rendered -by the .arbitraor maybe. eantered m ny court havingjmasi~ctimi aycor vnA £1.1gationtItarbitration is Aehed such dis'ts the partieswa uru litjgatio-ih any court of competentD. GoPv6ningLawThbis. Subcoontract shall beointeirpreted mid. governed:in ac~oodance withall api~cablefederalnd state:laws .ad elf applicable federal rules, and reguldtions,The- -par~es Amerstasal that :materials -40d informthion rsWot from. the perforrmance of this~16- tc to;-expboi eoatrol laws anmthak each party- is responsitble for ýt. owecompliancewit such..w5s.CLAUtSE fl--COST "C~OUNT-IN6 STANDARlDS _(CA&M)LIABILITYtApplcablk'to :Subeotitacts-exceeding *$009,OOCUausM&eA3boW incotporates lntoL teseGEMER-AL PROVISMOS clauses entitled, "COSTAC&W GSTAMN4D9" and 7ADMM R 0AVOPN. COSTAC .oU7 .sADAkD&"MiAW9 ding the~pfovislon of thes clat.1s bofayoteprvisionofteS cnra hSubcontractor-shall beliable.to _the .Govermentfor any infasd fcoa, -interrte n, resutitngfroim anyfWiltte wit eý pedt car-ied:.on at the site of the.work, or ofa subcontractor, to Comply witL applicablecostaccounting.standards or to follow any practicesdisrlosedpursanttot Page 15.5-7 APPENDIX 15.5, FUELS ASSISTANCE CONTRACT 12/2011(Smnderd No. W007820&CLAUSE 7-DISCLOSURE AND USE ýRSTRICTIONS FOR LIMITED RIGHTSDATAGenerally, delivery of Limited Rights Data (or Restricted Computer Software). should not benecesary. However, qnly if LixniteERights Daft Will be. used in.meeting the delivery requirements.of th subconract, the follow.ing disclosure.and use restrictidns, shall apply to and shall:be inserted in,-Oy yFAR 52227-14 Liri-tedFights.Notice on any Lhiiteid Rights Data furnished or .deliveredby theSubcontractor.or alower-tier subcontractorA. These "Limited Rights Data" may be disclosed for evaluation purposes under the restriction thatthe "Limited Rights Data"be retained iq confidence and not be further disclosed;B. These "%Lmited Rights Data' may. be disclosed to other contractors participating irs theof which this is a part forkitforrnation or use in. connectionwith ifh work pexformed under their contracts and under the. restriction that the'"Limited RightsDatai" be retaijiedih confidence and notibe fi.rher disilosed; and.C. these "Limited Rights. Datae" .may be used by the -Govnmmnt or others on its. behalf foremrgency .repair br bvebaul work utnder -the restrictior that the "Limited Fights Data" beretained confidence end not.be fin-her disclosed.CLAUSEM -ORDER OF PRECEENCEAny inconsistencies -in the documents comprising this Subcontract shall be resolved by givingprece4.ene in the. following order: -(a -the, SCHEDul OF ARTICLES and this SubcontractSign-aitur. Page;- ():theice GENERAL.fPROVISIONS- (0) other refere documen.its, exhibits,.andattachents;ý and (ri) any dferenced speaification or Staftmnn. 6f WorkCLAUSE 9-SCURI REOU EMSNTA. This Subbontractis intended foridlassified publicly releasablereseerch or. development work..TheC~ontra~ dor not results ofthe.researchprojet will involve classifiedor .Unclaml-hd ontroled nfomation (UGCNi).(See 10 CFR.yIt 10.17).However,.-ihe Contrator may reviewtheeeath worklgeaated ander: is Subcontraet at any.h_44tef., in. f it.reuires cls~ii~catiombr tatrtol.as .B. If, sbsequentito the date.of..thiSuboiitraqt, areview of theinformation reveals thgt.classified*irfo ai0. b :genera "ted~under-this .ubcopnact, then the s-ityrequiremnts ofthisSubcontact.:muet-be changed, If such changes causean increase:or decrease in costs orotherwise affiut-any other terni or condition of ths Suibcontract, the Subcontract shall be subjectto an equitable.adustmmentasif the changes weredirected under the Changes clause of this* Subcotrc*C. .1fthe security requirements are changed,.the Subcontractor shal.exert every reasonable effdrt.-nile, -with poicies.to co~nuet.p erfoomance of work underthe-Subcontraot in compliance witht-changain the security requirements. If the Subcontractordetf ='-iet th o' thetwouderthisSbna , s not praicabe because of thedaangein seeit req uiremenis the..Subwntractor.shall.notify the Contractor's ProcurementRepretenteaive.ih writing, presentative provides diecon,:the Subbcntrkctor shall protec thaiaterial %As directed by contractor.D), he writtens nobficaionrthe Contractor'. aro m Repr tative all-explowvthý cretnttaeswouniding the proposed change in security requirements and shallPage 15.5-8 THE UNIVERSITY OF TEXAS TRIGA II NUCLEAR RESEARCH REACTOR 12/2011SAFETY ANALYSIS REPORT, APPENDIX 15.5Bottefl Rn&*MyiAJ1nctLtCSmdnt~ r t.ub*fccnnrNo. 00.18206~Page 9.1f 12a mutgaly satiahtorynthod to. allow the Subcontraefor to continucperorpance. ofw7ork undx .this Subcontract.B. Witbin 15 days of .otif.. tibn 6fthe Subccntraetots.stated. inahility-to.p0 d..th ontracoo's: Representatiiemust detemmine.whether(l) thes.e securityrequirements do notapptis contract or (2) amwtually oatisfaeryrnetod for.enlioningperformance- of work under tbis:Subcontract dan be:agreed upoc. If this. determination is notmade, the Subcontractor may request.the Contractor's Procuremeat Representative to.terminatethe-Subcontract in whole or in part. The Contractor's.Procur ment Representativ.e 'shal.teninatethe Subcontract in whole or in part, as. may-be appropriate, and the ternination shall be deemed- atenradon undertheterms of.the .for the, Convenienc-of the ýGovernment clause,CLAUSE. PD.- _CLAUjSES INCORORA-TED BY REERENCE,The. FEDERAL ACQUISMITON RGLATION. (FAR).and.tfe U.S. D -PARTMENT OF ENERGY'ACQUISfMION !REGULATION. (DEAR) clauses listed below, which are located in Chapters 1 and9*, respectively, of Tlile ,48.of Oe Co4e ofl!ederal.a.egulations, areincorporated by this reference,as apar of thescOGNERAL. PROVISIONS with he'.same forcc-and effect as if they. were given in fulltext as-prescriý"bed ow.T ill textl. the olaus.:may.- e accessed electrodcally at h'bflg ff.AR> and* (bEAR).As used&in. the causes, the- erm shall ma.,nthisS'bcontracý the fter shallthe :tdnj: 1 .0.bAtract.r" sh9l the co andtheterms "Government"' exceptmin'FAR. lause52.227-44, and DEAR clauses-97(15227-4, 952.22.7-Il, 97065232-3 and 52.245-5 Alternate I, in:whieh clauses "Government' shall mean. the United; States Government. and "Contracting Officer'l.mea the DOEINNSA Cojtracting Officer for Prime.Contract DR-ACO7-05114517 with the4s ued in .rEAR cjlauses 952,204-72 and '952.227-9, the ter 'VOE" shall meahDOEINNSA or-the Contracfor.The inodifiatians at these tens are. intended to appropriately, identify the and:cstablshaf contrac~aeand admrninst~radve.etpdrtitlg rejatiohsh1ip,-andj§ha not- apply tofthe extent'they would U.5. Government's righfs. The Suhcnntado rshall inqlude the listed clausesin.itasbcoftratsany tier, to the extent.9pp1icqable,APPLICABLE TO ALL SUBCONEA~r NES TEWSEIDCTE)BLWDEAR 9$Z. 4-71t FR ,IGN N ATIN CONTROL.(APR. 1994),. Applis ifýthe. Subctontract is for-unclassffiedreneatch lirwotvigg nuclear technology..PAYENT- (DE2002). &xbstitt-31.3-insubcoctetrcsvioth educati~n~hin sti osr.312 in paragraph (a),FAR 52.12-16-15 PREDETERMNED IDIRECT COSTSRATES APR. 1.998).FAIA 52-222-21 PROIlTflON OF SEGkEGATED (FEB. 1999)..R. 5222-26 EQLALOPPORTUNITY (APR 2002).FAR.52-223-a 4US MATERIALIDENTIFICATION.AND MATERI.ALSAFETY DATA SU(IAN 1_.997) AND ALTERNAATE".I. ApplieszonlyMffSi0_cornmc t detive,. f lntdelis-satermas.Page 15.5-9 APPENDIX 15.5, FUELS ASSISTANCE CONTRACTI12/2011APPENDIX 15.5, FUELS ASSISTANCE CONTRACT 12/2011Batik egy A. ancet w".SwedanP Reseagch-Subeinterc No.. 007506Paga lOo6flE.FAR 52.22.5-1-3DEAR 970.5M247DEAR 952.127-1DEAR 952.227-ItFA23S22Z7-l4FAR 52.2273.FAR 5-2.2320FAR 52.22-22IFAR 52.2 42.5PAR $Z2a441RESTRICTIONS ONCEERTAIN FOREIGN PURCHASES (DEC:2003).AUTHORIZATTON AND CONSENT (AUG 200M), Paragraph (a).REFUND OF ROYALTIES (B.F 1995). Applies if "royalties" of morethan..$250 arepaid by.a-subcontractor at any tier.PATENT RIG4TS -RETBNTIONBY THE CONTRACTOR (SHORTFORM) (PEE 1995).. (Applies onlyffSubcontractoris a nonprofitorganization as 27.301, If Scbeontracior does notiquaa-y' lA-accordaice.with. 48 CFR 27.301, it may request .a pateat waiverpursuant t" 10 CFR.7.84;):[ ChI'ebkprovisipn .bapw tt applies. OR include p.oly applicable prision] .__ kRIGHTS IN DATA-GENERAL (JUN 1987) with ALTERNATE Vand DFAR. 927.409 Paragraphs (a) and (d)(3): Applies if the Subcontract isfor development work, or for basic and applied research where computersoftware is specified: as a Deliverable ii the Statemneat of Work or-otherspecial, ceircumstaaoesapply as specified in the agreement.jL :.RIGHS IN DATA-GENERAL.(JUN :1987) widiALTEIRNATEIV, subpamrafph'.( )4)-and DRAR..927.409, subaragraph (a) Definitions..Applies if~lhSbontrct Is for.basicor applied resewch and co utersoflware isot apoc.i~fe..as-a Deliverable in-theStatembnt of Wo*i cnd noother special circumstances apply :per DEAR 927.409.RIGHTS:T.OPROPOSALDATA (T'EC-HNICAL) (JNE .1987). Applies ifthe Subcontraisbased upon a-technical proposal.STATE RECEIPTS AND TAX.(ApR.2(03)., Applies.if aypart.qf this Subconract is to be-perfonned in theState Ort40w Me4oo.I4M1TATION OP COST.(A,.R.:. 1 .Applies if the Sjbcontiact-is fl.yfinidecLIMITATION OF FUNDS (APR 1984). Applies if i$e Subcontractisincrementally funded.STOP-WORK ORDER.AUG 1989) with ALTERNATE-I (APR 1984).C .ANGES.- 1981), WITHALTERNATE VSUBCO.NTRA .ACTS 1998.) ,. Insert in:Paragraph (e): .'Anysubcontratpu rpchýise:order.f& athe than."con=u-eial itens. ex ctedin¢lh:simplified acqpsition'threshotd. ("Cormmercialitem" has the meaning&mtained in FAR 52ZQZD~flitons7.) Applies only:if th&earesubcontracts.undert1hsCbntrat.DEAR.97005245q PROPER7Y.E 200).FAR :52.-24649: N QOF RES AND DFVELOPMENT (SHORT FOM(APR. -84).iA.R.52<247463 PREFERENCE OR U. S.l'AG.AIRCARRER (JUNE 2003). Appliesfifthe Subcontract-invoalves- international air! transportation.Page 15.5-10 THE UNIVERSITY OF TEXAS TRIGA II NUCLEAR RESEARCH REACTORSAFETY ANALYSIS REPORT, APPENDIX 15.5I12/2011f---9=datlell EitnyA1Ian eV,,LW.~ wtiof. 12FAR.52.247ý64.DEAR.96224 7-70ýFAR 52.24.5-DEAR, 952.217-70MEAR V0S.232-4.PREFERENCE IOR PRIVATL OWNEiD US-FLAG .COMMERCIAL.VESSELS. (APR 2 003).FOREIGN TRAVEL (DEC 2000).TERMINATION FOR CONVENiENcE OF THE GOVERNMENT(EDUCATIONAL AND OTHIERNONPROFIT INSTrrUTIONS) (SEP1996).ACQUISITION OF REAL (.APR 1994).. Applies ifthe.S.~ntrmdt i:voves l -ased space-that is reimbirseRECO"S,.AND .INSPECTION (DEC 2000)AP.LICABLE IF.E SUBCONTRACtTS F.OR 1i0,000 OR'MORE:.FAIR 52.222-i35NAASI 2-22-36FAR `5212Z.72-3PEQUAL OPPORTUNITY FOR SPECIAL DISABLED VETERANS.VETERANS OF. THE VIETNAM ERA AND OTHER ELIGIBLEVETERANS (DEC 2001).APRMA-1V0 ACTIONF VFkW.RXERSV A ITH DSABILrrIES(JUNE 1998).)EPLOYMENT R:EPODRTS ON SPECiIA DISABLE-D VETERANS,VEI'EWAS.OF THE VIETNAIMETA AND OTHER ELIGIBLEVETERANS- (DP-C2001).AIPTICABLE IF TffiSUBCONTRACT EXCEEDS $100,00:"AR5-2203-5 COVENANT AGAINST CONTINGENT EES.(APR 1984)FAR-S5.20;:6 REtS*9TRXCTNN S.UBCONTRACTOR SALES TO THEANTI-XCBACKPROCEDUME (jULY I )95), excluding.Pa ragraph'WillFAR52.20-12FAA5iý6I42FAR52.222-904DEAR 97Q.522.7,5PICE F ORFEE ADJUSTMENT FOR ILLEGAL OR IROPER~ACIVVIY(JAN1997).LIMITATION O'N PA-Y)MENTS to INFLUjENCE CERTAIN FEDERALTI~ANSCTINS MUN2003).UT-ILIZAT1ON OF SMALL BUSINESS CONCERNS (MAY 2004).:CON~TR-ACrWOR HOURS AND -SAFETY STANARS-ACT -OVERTIME :COMPENSATION(SE?.2000).NOTlCE. AND ASSISTANCE RGARDING PATENT. ANDCOPYI14T INM.RoNGEMENT : .(AUG 2002)..APPt-CAL :IF- TE[KSUBCONTT EXt.EDS $500,O0O-FAR 52.215: 10 PRIC-E.EDCT.INt FOR DEFECTiVE COST OR PRICING DATAý(OCT 199\bif S"5.,0000..Page 15.5-11 APPENDIX 15.5, FUELS ASSISTANCE CONTRACT 12/2011treflo,,Enpry,ýli=anoe,LLC6S00jdad~il R0M~h No. 0078206.Pa6c.12f 612FAR 52.215-1 PRICE. RDlUCTION FORkDEFECTIVE COST ORPRICING DATA-MODIFICATIONS (OCT 19.97) not us-dwben 52.215410 is included. Insubcontracts:greater than $550,000..FAR 52_24%,512 SUBCONTRACTOR COST OR PRICLNG DATA (OCT-1997.. Applies if52.2I-5-,160applies.RAR 52,21543 SUB-ONIR "AcOro CING DATA-MODIFICATIONS(OC19.97). Applies if52.2l:5-l.1. app ies.FAR 52.2!9-9 SMALL BUSINE$S-SUBCONTRACTING PLAN (JAN 2002). Appliesutiess there are no subcontracting possibilities.FAR 52.22746 ADDITIONAL DATA REQUIREME S (JUNE 1987).FAR 52.230-2 COST ACCOUNTING-STANDARDS-(APR 1998), excluding.paragraph(b),. Appies o nonprofit orgmsizati.0m if they:are subject to fiJl CAScoverage. as:set forth:in 48 CFR.Chapter 99, Subp.at 9903.201-2 (FARAppehdixB)7FAR. 52.2,3 DISCLoSLRANT)-CONSTENCY-OF COSTACCOUNMIGPRACTICES (,4PR 1995), eJ(Gdingpatagrapji (): plbhitpbl0oganizafint .if the are 'Ubject taomodifted CAS coveageas set forth in48 CFR9Chbater,99, S*arO903.201-2 (FAR Appeadix B).FAR 52-230-5 COST ACCOUNTING:STMibARDS! -EWCATIONAL INSTITUTION(APR 1909), picludlng pxtagraphý (b),FAR 52.2306 ABthlNSTRATION' OP COST AGC(UNTING STANDARDS (NOV( O.1999).Page 15.5-12

• .ANETLTechnicalSpecificationsThe University of Texas at AustinNuclear Engineering Teaching LaboratoryTRIGA Mark II Nuclear Research ReactorLicense R-129Docket 50-60212 December 2011The University of Texas at AustinNuclear Engineering Teaching Laboratory10100 Burnet Rd, Bldg 159Austin, TX 78758 UT TRIGA II TECHNICAL SPECIFICATIONSTable of Contents1. DEFINITIONS ........................................................................................................... TS-52. SAFETY LIM ITS AND LIM ITING SAFETY SYSTEM SETTINGS ................................... TS-102.1 Fuel Elem ent Tem perature Safety Lim it ......................................................... TS-102.1.1. Applicability ........................................................................................... TS-102.1.2. Objective ............................................................................................... TS-102.1.3. Specification .......................................................................................... TS-102.1.4. Actions ................................................................................................... TS-102.1.5. Basis ...................................................................................................... TS-102.2 Lim iting Safety System Settings ...................................................................... TS-122.2.1. Applicability ........................................................................................... TS-122.2.3. Objective ............................................................................................... TS-122.2.4. Specification .......................................................................................... TS-122.2.5. Actions ................................................................................................... TS-122.2.6. Basis ...................................................................................................... TS-123. LIM ITING CONDITIONS FOR OPERATIONS ............................................................. TS-143.1 CORE REACTIVITY ............................................................................................ TS-143.1.1. Applicability ........................................................................................... TS-143.1.3. Objective ............................................................................................... TS-143.1.4. Specification .......................................................................................... TS-143.1.5. Actions .................................................................................................. TS-143.1.6. Basis ................................................................................................................ TS-153.2 PULSED M ODE OPERATIONS .......................................................................... TS-173.2.1. Applicability ........................................................................................... TS-173.2.3. Objective .............................................................................................. TS-173.2.4. Specification .......................................................................................... TS-173.2.5. Actions ................................................................................................... TS-173.2.6. Basis ...................................................................................................... TS-173.3 M EASURING CHANNELS ................................................................................. TS-183.3.1. Applicability ........................................................................................... TS-183.3.3. Objective ............................................................................................... TS-183.3.4. Specification .......................................................................................... TS-183.3.5. Actions ................................................................................................... TS-183.3.6. Bases ..................................................................................................... TS-213.4. SAFETY CHANNEL AND CONTROL ROD OPERABILITY ..................................... TS-233.4.1. Applicability ........................................................................................... TS-233.4.3. Objective ............................................................................................... TS-233.4.4. Specification .......................................................................................... TS-233.4.5. Actions ................................................................................................... TS-243.4.6. Basis ...................................................................................................... TS-243.5 GASEOUS EFFLUENT CONTROL ....................................................................... TS-253.5.1. Applicability ........................................................................................... TS-25TS-1 12/2011 TECHNICAL SPECIFICATIONS3.5.3. Objective ............................................................................................... TS-253.5.4. Specification .......................................................................................... TS-253.5.5. Actions ................................................................................................... TS-263.5.6. Basis ...................................................................................................... TS-273.6 LIM ITATIONS ON EXPERIM ENTS ..................................................................... TS-293.6.1. Applicability ........................................................................................... TS-293.6.3. Objective ............................................................................................... TS-293.6.4. Specification .......................................................................................... TS-293.6.5. Actions ................................................................................................... TS-293.6.6. Basis ...................................................................................................... TS-303.7 FUEL INTEGRITY ............................................................................................. TS-313.7.1. Applicability ........................................................................................... TS-313.7.3. Objective ............................................................................................... TS-313.7.4. Specification .......................................................................................... TS-313.7.5. Actions ................................................................................................... TS-313.7.6. Basis ...................................................................................................... TS-323.8 REACTOR POOL W ATER .................................................................................. TS-333.8.1. Applicability ........................................................................................... TS-333.8.3. Objective ............................................................................................... TS-333.8.4. Specification .......................................................................................... TS-333.8.5. Actions ................................................................................................... TS-333.8.6. Basis ...................................................................................................... TS-343.9 Retest Requirem ents ...................................................................................... TS-363.9.1. Applicability ........................................................................................... TS-363.9.3. Objective ............................................................................................... TS-363.9.4. Specification ........................................ TS-363.9.5. Actions ................................................................................................... TS-363.9.6. Basis ...................................................................................................... TS-364. SURVIELLANCES ..................................................................................................... TS-374.1 CORE REACTIVITY ............................................................................................ TS-374.1.1. Objective ............................................................................................... TS-374.1.2. Specification .......................................................................................... TS-374.1.3. Basis ...................................................................................................... TS-374.2 PULSE M ODE ................................................................................................... TS-384.2.1. Objective ............................................................................................... TS-384.2.2. Specification .......................................................................................... TS-384.2.3. Basis ...................................................................................................... TS-384.3 M EASURING CHANNELS ................................................................................. TS-394.3.1. Objective ............................................................................................... TS-394.3.2. Specification .......................................................................................... TS-394.3.3. Basis ...................................................................................................... TS-404.4 SAFETY CHANNEL AND CONTROL ROD OPERABILITY ..................................... TS-414.4.1. Objective ............................................................................................... TS-4112/2011 TS-2 UT TRIGA II TECHNICAL SPECIFICATIONS4.4.2. Specification .......................................................................................... TS-414.4.3. Basis ...................................................................................................... TS-414.5 GASEOUS EFFLUENT CONTROL ....................................................................... TS-434.5.1. Objective ............................................................................................... TS-434.5.2. Specification .......................................................................................... TS-434.5.3. Basis ...................................................................................................... TS-434.6 LIM ITATIONS ON EXPERIM ENTS ...................................................................... TS-444.6.1. Objective .............................................................................................. TS-444.6.2. Specification .......................................................................................... TS-444.6.3. Basis ...................................................................................................... TS-444.7 FUEL INTEGRITY .............................................................................................. TS-454.7.1. Objective ............................................................................................... TS-454.7.2. Specification .......................................................................................... TS-454.7.3. Basis ...................................................................................................... TS-454.8 REACTOR POOL W ATER .................................................................................. TS-464.8.1. Objective ............................................................................................... TS-464.8.2. Specification .......................................................................................... TS-464.8.3. Basis ...................................................................................................... TS-464.9 M AINTENANCE RETEST REQUIREM ENTS ........................................................ TS-474.9.1. Objective ............................................................................................... TS-474.9.2. Specification .......................................................................................... TS-474.10.3. Basis .................................................................................................... TS-475. DESIGN FEATURES .................................................................................................. TS-485.1 REACTOR FUEL ................................................................................................ TS-485.1.1. Applicability .......................................................................................... TS-485.1.2. Objective ............................................................................................... TS-485.1.3. Specification .......................................................................................... TS-485.1.4. Basis ...................................................................................................... TS-485.2 REACTOR FUEL AND FUELED DEVICES IN STORAGE ....................................... TS-485.2.1. Applicability .......................................................................................... TS-485.2.2. Objective ............................................................................................... TS-485.2.3. Specification .......................................................................................... TS-495.2.4. Basis ...................................................................................................... TS-495.3 REACTOR BUILDING ........................................................................................ TS-495.3.1. Applicability ........................................................................................... TS-495.3.2. Objective ............................................................................................... TS-495.3.3. Specification .......................................................................................... TS-495.3.4. Basis ...................................................................................................... TS-505.4 EXPERIM ENTS ................................................................................................. TS-505.4.1. Applicability ........................................................................................... TS-505.4.2. Objective ............................................................................................... TS-505.4.3. Specification .......................................................................................... TS-505.4.4. Basis ...................................................................................................... TS-51TS-3 12/2011 TECHNICAL SPECIFICATIONS6. ADMINISTRATIVE CONTROLS ................................................................................. TS-536.1 ORGANIZATION AND RESPONSIBILITIES OF PERSONNEL ............................... TS-536.2 REV IEW AN D A UDIT ........................................................................................ TS-586.3 PRO C ED U RES .................................................................................................. TS-606.4 REVIEW OF PROPOSALS FOR EXPERIMENTS .................................................. TS-616.5 EMERGENCY PLAN AND PROCEDURES ........................................................... TS-626.6 OPERATOR REQUALIFICATION ........................................................................ TS-626.7 PHYSICAL SECURITY PLAN .............................................................................. TS-636.8 ACTION TO BE TAKEN IN THE EVENT A SAFETY LIMIT IS VIOLATED ............... TS-636.9 ACTION TO BE TAKEN IN THE EVENTOF A REPORTABLE OCCURRENCE .................................................................... TS-636.10 PLANT OPERATING RECORDS ......................................................................... TS-646.11 REPORTING REQUIREMENTS ................................................................................. TS-6612/2011TS-4 UT TRIGA II TECHNICAL SPECIFICATIONS1. DEFINITIONSThe following frequently used terms are defined to aid in the uniform interpretation ofthese specifications. Capitalization is used in the body of the Technical Specifications toidentify defined terms.ACTION Actions are steps to be accomplished in the event a requiredcondition identified in a "Specification" section is not met, as statedin the "Condition" column of "Actions."In using Action Statements, the following guidance applies:* Where multiple conditions exist in an LCO, actions are linked tothe failure to meet a "Specification" "Condition" by letters andnumber.* Where multiple action steps are required to address acondition, COMPLETION TIME for each action is linked to theaction by letter and number." AND in an Action Statement means all linked steps need to beperformed to complete the action; OR indicates options andalternatives, only one item needs to be performed to completethe action." If a "Condition" exists, the "Action" consists of completing allsteps associated with the selected option (if applicable) unlessthe "Condition" is corrected prior to completion of the stepsANNUAL 12 months, not to exceed 15 monthsBIENNIAL Every two years, not to exceed a 30 month intervalCHANNELCALIBRATIONCHANNEL CHECKCHANNEL TESTA channel calibration is an adjustment of the channel so that itsoutput responds, with acceptable range and accuracy, to knownvalues of the parameter that the channel measures.A channel check is a qualitative verification of acceptableperformance by observation of channel behavior. This verificationshall include comparison of the channel with expected values, otherindependent channels, or other methods of measuring the samevariable where possible.A channel test is the introduction of an input signal into a channelTS-512/2011 TECHNICAL SPECIFICATIONSCONFINEMENTCONFINEMENTISOLATIONCONTROL ROD(STANDARD)CONTROL ROD(TRANSIENT)DAILYENSUREEXCESSREACTIVITYEXPERIMENTEXPERIMENTALFACILITYIMMEDIATEto verify that it is operable. A functional test of operability is achannel test.The enclosure which controls the movement into and out of thereactor bayCondition for reactor bay ventilation where:(1) dampers controlling confinement ventilation are closed, and(2) confinement ventilation fans are secured(3) the reactor bay fume/sort hood fans are secured(4) the reactor bay fume/sort hood dampers are closedThe purge system may be operated in manual overrideA standard control rod is one having an electric induction orstepper motor drive coupled to the control rod by anelectromagnet, with scram capability.A transient rod is one that is pneumatically coupled to the controlrod drive, is capable of initiating a power pulse, is operated by amotor drive, and/or air pressure operated and has scram capability.Prior to initial operation each day (when the reactor is operated), orbefore an operation extending more than 1 dayVerify existence of specified condition or (if condition does notmeet criteria) take action necessary to meet conditionThat amount of reactivity above the critical condition which wouldexist if all the control rods were moved to the maximum positivereactivity conditionAn EXPERIMENT is (1) any apparatus, device, or material placed inthe reactor core region (in an EXPERIMENTAL FACILITY associatedwith the reactor, or in line with a beam of radiation emanating fromthe reactor) or (2) any in-core operation designed to measurereactor characteristics.Experimental facilities are the beamports, pneumatic transfersystems, central thimble, rotary specimen rack, and displacementof fuel element positions used for EXPERIMENTS (single-elementpositions and the multiple element positions fabricated in theupper grid plate displacing 3, 6 or 7 elements).Without delay, and not exceeding one hour.12/2011TS-6 UT TRIGA II TECHNICAL SPECIFICATIONSINITIAL STARTUPLIMITINGCONDITION FOROPERATION (LCO)LIMITING SAFETYSYSTEM SETTING(LSSS)MEASUREDVALUEMEASURINGCHANNELMOVABLEEXPERIMENTOPERABLEOPERATINGNOTE:IMMEDIATE permits activities to restore required conditions for upto one hour; this does not permit or imply deferring or postponingactionA reactor startup and approach to power following:1 Fuel element or control rod relocations or installations withinthe reactor core region2 Relocation or installation of any experiment in the core regionwith a reactivity worth of greater than one dollar3 Recovery from an unscheduled (a) shutdown or (b) significantpower reductions, or4 modifications to reactor safety or control rod drive systems.The lowest functional capability or performance levels ofequipment required for safe operation of the facility.Settings for automatic protective devices related to those variableshaving significant safety functions. Where a limiting safety systemsetting is specified for a variable on which a safety limit placed, thesetting shall be chosen so that the automatic protective action willcorrect the abnormal situation before a safety limit is exceeded.The measured value of a parameter is the value as indicated at theoutput of a MEASURING CHANNEL.A MEASURING CHANNEL is the combination of sensor, lines,amplifiers, and output devices that are connected for the purposeof measuring the value of a process variable.A MOVABLE EXPERIMENT is one the EXPERIMENT may be movedinto, out-of or near the reactor while the reactor is OPERATING.A system or component is OPERABLE when it is capable ofperforming its intended function in a normal mannerA system or component is OPERATING when it is performing itsintended function in a normal manner.The reactor is in the PULSE MODE when the key switch is in the"on" position, the reactor mode selection switch is in the pulseposition and the reactor display indicates pulse mode.PULSE MODETS-712/2011 TECHNICAL SPECIFICATIONSREACTOR SAFETYSYSTEMREACTORSHUTDOWNREFERENCE CORECONDITIONSAFETY CHANNELSAFETY LIMITSSECUREDEXPERIMENTSHALL(SHALL NOT)SEMIANNUALSHUTDOWNMARGINSTANDARD FUELELEMENTNOTE:In the PULSE MODE, reactor power may be increased on a period ofmuch less than I second by motion of the transient control rod.The REACTOR SAFETY SYSTEM is that combination of MEASURINGCHANNELS and associated circuitry that is designed to initiate areactor scram or that provides information that requires manualprotective action to be initiated.The reactor is shutdown if it is subcritical by at least the minimumrequired amount of reactivity (shutdown margin) in the REFERENCECORE CONDITION with the reactivity worth of all experimentsincluded.The condition of the core when it is at ambient temperature (cold)and the reactivity worth of xenon is negligible (<$0.30)A safety channel is a MEASURING CHANNEL in the REACTOR SAFETYSYSTEM.Limits on important process variables which are found to benecessary to protect reasonably the integrity of the principalbarriers (i.e., fuel element cladding) which guard against theuncontrolled release of radioactivity.A secured EXPERIMENT is an EXPERIMENT held firmly in place by amechanical device or by gravity providing that the weight of theEXPERIMENT is such that it cannot be moved by forces (1) normalto the operating environment of the experiment or (2) that mightresult from credible failures.Indicates specified action is required/(or required not to beperformed)Every six months, with intervals not greater than 7 1/2 monthsThe shutdown margin is the minimum shutdown reactivitynecessary to provide confidence that the reactor can be madesubcritical by means of the control and safety systems, startingfrom any permissible operating condition, and that the reactor willremain subcritical without further operator actionA standard fuel element is a single TRIGA element of standard type,U-ZrH clad in stainless steel with nominal hydrogen to zirconiumratio of 1.6.12/2011TS-8 UT TRIGA II TECHNICAL SPECIFICATIONSINTSTRUMENTEDFUEL ELEMENTSTEADY-STATEMODETECHNICALSPECIFICATIONVIOLATIONAn instrumented fuel element (IFE) is a stainless steel clad fuelelement containing three sheathed thermocouples embedded inthe fuel element.The reactor is in the steady-state mode when the key switch is inthe "on" position, the reactor mode selector pushbutton switch hasrequested either the manual, automatic, or square wave positionand the reactor display indicates manual, automatic, or squarewave.(1) A violation of a Safety Limit occurs when the Safety Limit valueis exceeded.(2) A violation of a Limiting Safety System Setting or LimitingCondition for Operation) occurs when a "Condition" existswhich does not meet a "Specification" and the corresponding"Action" has not been met within the required "CompletionTime."A violation has not occurred if the "Action" statement of (1) anLSSS or LCO is completed or (2) the "Specification" is restoredwithin the prescribed "Completion Time,"NOTE"Condition," "Specification," "Action," and "Completion Time" referto applicable titles of sections in individual Technical SpecificationsTS-912/2011 TECHNICAL SPECIFICATIONS2. SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS2.1 Fuel Element Temperature Safety Limit2.1.1 ApplicabilityThis specification applies when the reactor in STEADY STATE MODE and the PULSEMODE.2.1.2 ObjectiveThis SAFETY LIMIT ensures fuel element cladding integrity2.1.3 Specification2.1.4 ActionsCONDITION REQUIRED ACTION COMPLETION TIMEA. Stainless steel clad,high-hydride fuel A.1 ENSURE SHUTDOWNelement temperature conditionexceeds 11500C.OR ANDB. Fuel temperatureexceeds 7500C in A.2 Report per Section 6.8 A.2 Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />steady state conditions2.1.5 BasesSafety Analysis Report Chapter 4 (4.2.1 B) identifies design and operating constraints forTRIGA fuel that will ensure cladding integrity is not challenged.NUREG 1282 identifies the safety limit for the high-hydride (ZrH1.6) fuel elements withstainless steel cladding based on the stress in the cladding (resulting from the hydrogenpressure from the dissociation of the zirconium hydride). This stress will remain belowthe yield strength of the stainless steel cladding with fuel temperatures below 11500C.A change in yield strength occurs for stainless steel cladding temperatures of 500°C, but12/2011TS-10 UT TRIGA II TECHNICAL SPECIFICATIONSthere is no scenario for fuel cladding to achieve 500°C while submerged or in air;consequently the safety limit during reactor operations is 1150'C.Therefore, the important process variable for a TRIGA reactor is the fuel elementtemperature. This parameter is well suited as a single specification, and it is readilymeasured. During operation, fission product gases and dissociation of the hydrogen andzirconium builds up gas inventory in internal components and spaces of the fuelelements. Fuel temperature acting on these gases controls fuel element internalpressure. Limiting the maximum temperature prevents excessive internal pressuresthat could be generated by heating these gases.Fuel growth and deformation can occur during normal operations, as described inChapter 4 (4.2.1 Z). Damage mechanisms include fission recoils and fission gases,strongly influenced by thermal gradients. Limiting steady state operating fueltemperature to less than 7500C limits potential fuel growth.TS-1 112/2011 TECHNICAL SPECIFICATIONS2.2 Limiting Safety System Settings (LSSS)2.2.1 ApplicabilityThis specification applies when the reactor in STEADY STATE MODE2.2.2 ObjectiveThe objective of this specification is to ensure the safety limit is not exceeded.2.2.3 SpecificationsA Power level SHALL NOT exceed 1100 kW (th) in STEADY STATE MODE of operationB Instrumented elements in the B or C ring SHALL indicate less than 550'C2.2.4 ActionsCONDITION REQUIRED ACTION COMPLETION TIMEA.1 Reduce power to less than A.1 IMMEDIATE1100 kW (th)A. Steady state powerlevel exceeds 1100 kW OR(th)A.2. ENSURE REACTORSHUTDOWN conditionB.1. ENSURE REACTOR B.2. IMMEDIATEB. An INTSTRUMENTED SHUTDOWN conditionFUEL ELEMENT in the Bor C ring indicates ORgreater than 550'CB.2 VERIFY the MEASURED B.2 IMMEDIATEVALUE is not correct2.2.5 BasesAnalysis in SAR Chapter 4 (4.6 B) demonstrates that if operating thermal (th) power is1100 kW, the maximum steady state fuel temperature is less than the safety limit forsteady state operations by a large margin. For normal pool temperature, calculations inChapter 4 demonstrate that the heat flux of the hottest area of the fuel rod generatingthe highest power level in the core during operations is less than the critical heat flux bya large margin up to the maximum permitted cooling temperatures; margin remainseven at temperatures approaching bulk boiling for atmospheric conditions. Therefore,12/2011TS-12 UT TRIGA II TECHNICAL SPECIFICATIONSsteady state operations at a maximum of 1100 kW meet requirements for safeoperation with respect to maximum fuel temperature and thermal hydraulics by a widemargin. Steady state operation of 1100 kW was assumed in analyzing the loss ofcooling and maximum hypothetical accidents. The analysis assumptions are protectedby assuring that the maximum steady state operating power level is 1100 kW.The actual safety system setting will be chosen to ensure that a scram will occur at alevel that does not exceed 1,100 kW.Instrumented fuel element temperatures less than 550°C ensures the SAFETY LIMIT onfuel temperature is met.TS-1312/2011 TECHNICAL SPECIFICATIONS3. LIMITING CONDITIONS FOR OPERATION (LCO)3.1 Core Reactivity3.1.1 ApplicabilityThese specifications are required prior to entering STEADY STATE MODE or PULSINGMODE in OPERATING conditions; reactivity limits on experiments are specified inSection 3.8.3.1.2 ObjectiveThis LCO ensures the reactivity control system is OPERABLE, and that an accidental orinadvertent pulse does not result in exceeding the safety limit.3.1.3 SpecificationAThe maximum available core reactivity (EXCESS REACTIVITY) with all control rodsfully withdrawn does not exceed 4.9% Akk ($7.00) when:1. REFERENCE CORE CONDITIONS exists2. No MOVEABLE EXPERIMENTS with net-negative reactivity worth are inplaceThe reactor is capable of being made subcritical by a SHUTDOWN MARGIN morethan 0.002 Ak/k ($0.29) under REFERENCE CORE CONDITIONS and the followingconditions:B 1. The highest worth control rod is fully withdrawn2. The highest worth NONSECURED EXPERIMENT is in its most positivereactive state, and each SECURED EXPERIMENT is in its most reactivestate.3.1.4 ActionsCONDITION REQUIRED ACTION COMPLETION TIMEA.1 ENSURE REACTOR A.1 IMMEDIATEA. Reactivity with all control SHUTDOWNrods fully withdrawnexceeds 4.9% Ak/k($7.00) A.2 Configure reactor to A.2 Prior to continuedmeet LCO operations12/2011TS-14 UT TRIGA II TECHNICAL SPECIFICATIONSB.1.a ENSURE operable B.1 IMMEDIATEcontrol rods are fullyinsertedANDB.1.b Secure electricalpower to the controlrod circuits (magnetB. The reactor is not or motor power)subcritical by more than B.2 Prior to continued0.002 Ak/k ($0.29) under AND operationsspecified conditionsB.1.c Secure all work onin-core experimentsor installed controlrod drivesANDB.2 Configure reactor tomeet LCO3.1.5 BasesThe stated value for excess reactivity was used in establishing core conditions forcalculations in Chapter 13 (13.4) to demonstrate fuel temperature limits are met duringpotential accident scenarios under extremely conservative conditions of analysis. Sincethe fundamental protection for the UT reactor is the maximum power level and fueltemperature that can be achieved with the available positive core reactivity,experiments with positive reactivity are included in determining excess reactivity. Sinceexperiments with negative reactivity will increase available reactivity if they areremoved during operation, they are not credited in determining excess reactivity.Analysis shows that at the limiting pool water temperature and zero power, fueltemperature approaches 950'C with a reactivity addition of $5.94, and 1050'C with areactivity addition of $5.66, while a $4.00 reactivity addition results in peak fueltemperature of about 770°C. If the pulse occurs with the reactor operating at 880 MW,a $4.00 reactivity insertion results in peak fuel temperature of 930°C; this is only 3%below the safety limit for cladding with temperature greater than 500'C, but is wellbelow the safety limit when cladding temperature is less than 500°C. Since the claddingtemperature is shown to be less than 500°C with the reactor operating in Chapter 4,worst-case steady state operation at 880 kW leads to a maximum fuel temperature wellbelow the safety limit.TS-1512/2011 TECHNICAL SPECIFICATIONSThe limiting SHUTDOWN MARGIN is necessary so that the reactor can be shut downfrom any operating condition, and will remain shutdown after cool down and xenondecay, even if one control rod (including the transient control rod) should remain in thefully withdrawn position. Analysis in Chapter 4 (4.5.1) demonstrates the capability of thecontrol rods to meet this requirement.12/2011TS-16 UT TRIGA II TECHNICAL SPECIFICATIONS3.2 PULSED MODE Operations3.2.1 ApplicabilityThese specifications apply to operation of the reactor in the PULSE MODE.3.2.2 ObjectiveThis Limiting Condition for Operation prevents fuel temperature safety limit from beingexceeded during PULSE MODE operation.3.2.3 SpecificationA hless thansient or equaldrive is positioned6k ($4.00) for reactivity insertion (upon withdrawal)3.2.4 ActionsCONDITION REQUIRED ACTION COMPLETION TIMEA. With all stainless steel A.1 Position the transient rod A.1 IMMEDIATEclad fuel elements, the drive for pulse rod worthworth of the pulse rod in less than or equal to $4.00the transient rod drive ORposition is greater than$4.00 in the PULSE A.2 Place reactor in STEADY A.2 IMMEDIATEMODE STATE MODE3.2.5 BasesThe value for pulsed reactivity with all stainless steel elements in the core was used inestablishing core conditions for calculations in Chapter 13 (13.4) that demonstrate fueltemperature limits are met during potential accident scenarios under extremelyconservative conditions of analysis.TS-1712/2011 TECHNICAL SPECIFICATIONS3.3 MEASURING CHANNELS3.3.1 ApplicabilityThis specification applies to the reactor MEASURING CHANNELS during STEADY STATEMODE and PULSE MODE operations.3.3.2 ObjectiveThe objective is to require that sufficient information is available to the operator toensure safe operation of the reactor3.3.3 SpecificationsA The MEASURING CHANNELS specified in TABLE 1 SHALL be OPERATINGB The neutron count rate on the startup channel is greater 2x10-7 %The particulate continuous air monitor SHALL be operating and capable ofinitiating CONFINEMENT ISOLATIONTABLE 1: MINIMUM MEASURING CHANNEL COMPLEMENTMinimum Number OperableMEASURING CHANNEL STEADY STATEMODE PULSE MODEMODEReactor power level[1] 2 1Primary Pool Water Temperature 1 1Fuel Temperature 1 1Pool area radiation monitor[21 1 1Lower or middle level area monitor[21 1 1Effluent air radiation monitor 1 1Particulate air radiation monitor 1 1NOTE[I]: One "Startup Channel" required to have range that indicates <10 WNOTE[21: High-level alarms audible in the control room may be used3.3.4 ActionsCONDITION REQUIRED ACTION COMPLETION TIMEA.I.1 Restore channel toA.1 Reactor power channels operton2 operation A.1.1 IMMEDIATEnot OPERATING (min 2 Ofor STEADY STATE, 1 A.1.2 ENSURE reactor is A.1.2 IMMEDIATEPULSE MODE) SHUTDOWN12/2011TS-18 UT TRIGA II TECHNICAL SPECIFICATIONSCONDITION REQUIRED ACTION COMPLETION TIMEA.2.1 Establish REACTORA.2 High voltage to reactor SHUTDOWN conditionsafety channel (power A.2. IMMEDIATElevel) detector less than AND80% of requiredoperating value A.2.2 Enter REACTOR SECUREDmodeB.1 Restore channel tooperation B.1 IMMEDIATEORB. Primary water B.2 Monitor pool watertemperature, reactorbay differential pressure temperature B.2 IMMEDIATEANDor fuel temperatureCN noel teoperablre OR At least once per hourCHANNEL not operableB.3 ENSURE reactor is B.3 IMMEDIATESHUTDOWNC.1 Restore MEASURING C.1 IMMEDIATECHANNELORC.2 ENSURE reactor is C.2 IMMEDIATEshutdownC. Pool Area Radiation ORMonitor is not C.3 IMMEDIATEOPERATING C.3 ENSURE personnel are noton the upper levelORC.4 IMMEDIATEC.4 ENSURE personnel onupper level are usingportable survey meters tomonitor dose ratesTS-1912/2011 TECHNICAL SPECIFICATIONSCONDITION REQUIRED ACTION COMPLETION TIMED.1 Restore MEASURING D.1 IMMEDIATECHANNELORD.2 ENSURE reactor is D.2 IMMEDIATEshutdownD. Lower or middle level ORarea monitor is not D.3 ENSURE personnel are not D.IMEATOPERATING in the reactor bayORD.4 IMMEDIATED.4 ENSURE personnelentering reactor bay areusing portable surveymeters to monitor doseratesE.1 Restore MEASURING E.1 IMMEDIATECHANNELORE.2 ENSURE reactor is E.2. IMMEDIATEshutdownE. Continuous particulate.air radiation monitor isnot OPERATING E.3.a ENSURE Argon 41IMMEDIATEmonitor radiationmonitor is OPERATINGANDE.3.b Within 30E.3.b Restore MEASURING working daysCHANNEL12/2011TS-20 UT TRIGA II TECHNICAL SPECIFICATIONSCONDITION REQUIRED ACTION COMPLETION TIMEF.1 Restore MEASURING F.1 IMMEDIATECHANNELORF.2 ENSURE reactor is F.2. IMMEDIATEshutdownF. Argon monitor is not OROPERATING F.3.a. IMMEDIATEF.3.a ENSURE continuous airradiation monitor isOPERATINGANDF.3.b Within 30F.3.b Restore MEASURING working daysCHANNELG.1 Do not perform a reactor G.1 IMMEDIATEG. The neutron count rate startupon the startup channel ORis not greater than G.2 Perform a neutron-source2x10"7 % check on the startup G.2 IMMEDIATEchannel prior to startup3.3.5 BasesMaximum steady state power level is 1100 kW; neutron detectors measure reactorpower level. Chapter 4*and 13 discuss normal and accident heat removal capabilities.Chapter 7 discusses radiation detection and monitoring systems, and neutron andpower level detection systems.According to General Atomics, detector voltages less than 80% of required operatingvalue do not provide reliable, accurate nuclear instrumentation. Therefore, if operatingvoltage falls below the minimum value the power level channel is inoperable.Pool water temperature indication is required to assure water temperature limits aremet, protecting primary cleanup resin integrity. Analysis in Chapter 4 and 13 assume amaximum fuel temperature based on protection of resin integrity. Fuel temperatureindication provides a means of observing that the safety limits are met.The upper and lower level area radiation monitors provide information about radiationhazards in the reactor bay. A loss of reactor pool water (Chapter 13), changes inTS-2112/2011 TECHNICAL SPECIFICATIONSshielding effectiveness (Chapter 11), and releases of radioactive material to therestricted area (Chapter 11) that could cause changes in radiation levels within thereactor bay detectable by these monitors. Portable survey instruments will detectchanges in radiation levels.The air monitors (continuous particulate air- and argon radiation-monitor) provideindication of airborne contaminants in the reactor bay. These channels provideevidence of fuel element failure on independent channels; the particulate air monitorgas has maximum sensitivity to iodine and particulate activity, while the argon channeldetects noble gas.Permitting operation using a single channel of atmospheric monitoring willreduce unnecessary shutdowns while maintaining the ability to detect abnormalconditions as they develop. Relative indications ensure discharges are routine;abnormal indications trigger investigation or action to prevent the release of radioactivematerial to the surrounding environment. Ensuring the alternate airbornecontamination monitor is functioning during outages of one system provides thecontamination monitoring required for detecting abnormal conditions. Limiting theoutage for a single unit to a maximum of 30 days ensures radioactive atmosphericcontaminants are monitored while permitting maintenance and repair outages on theother system.Chapter 13 discusses inventories and releases of radioactive material from fuelelement failure into the reactor bay, and to the environment. Particulate and noble gaschannels monitor more routine discharges. Chapter 11 discusses routine discharges ofradioactive gasses generated from normal operations into the reactor bay and into theenvironment. Chapters 3 and 9 identifies design bases for the confinement andventilation system. Chapter 7 discusses air-monitoring systems. The 30 day interval isselected as adequate to accomplish complex repairs, and limited enough that with onesystem functional there is no significant chance that the system will fail during a periodthat requires detection of airborne radioactivity.Experience has shown that subcritical multiplication with the neutron source used in thereactor does not provide enough neutron flux to correspond to an indicated power levelof 2x10-7 %. Therefore an indicated power of 2x107 %. or more indicates operating in apotential critical condition, and at least one neutron channel is required with sensitivityat a neutron flux level corresponding to reactor power levels less than 2x107 %("Startup Channel"). If the indicated neutron level is less than the minimum sensitivityfor the channel, a neutron source will be used to determine that the channels isresponding to neutrons to ensure that the channel is functioning prior to startup.12/2011TS-22 UT TRIGA II TECHNICAL SPECIFICATIONS3.4 Safety Channel and Control Rod Operability3.4.1 ApplicabilityThis specification applies to the reactor MEASURING Channels during STEADY STATEMODE and PULSE MODE operations.3.4.2 ObjectiveThe objectives are to require the minimum number of REACTOR SAFETY SYSTEMchannels that must be OPERABLE in order to ensure that the fuel temperature safetylimit is not exceeded, and to ensure prompt shutdown in the event of a scram signal.3.4.3 SpecificationsA The SAFETY SYSTEM CHANNELS specified in TABLE 2 are OPERABLECONTROL RODS (STANDARD) are capable of full insertion from the fullyB withdrawn position in less than 1 sec.TABLE 2: REQUIRED SAFETY SYSTEM CHANNELSMinimum Function Required OPERATINGSafety System Number ModeChannel or Operable STEADY PULSEInterlock STATE MODEMODEReactor power 2 Scram YES NAlevelManual scram bar 1 Scram YES YESCONTROL ROD Prevent withdrawal of(STANDARD) 1 standard rods in the NA YESposition interlock PULSE MODEPulse rod Prevent inadvertentinterlockill 1 pulsing while in YES NASTEADY STATE MODENOTE [1]: The pulse rod interlock prevents air from being applied to the pulse rod unlessthe transient rod is fully inserted except during pulse mode or square waveoperations.TS-2312/2011 TECHNICAL SPECIFICATIONS3.4.4 ActionsCONDITION REQUIRED ACTION COMPLETION TIMEA.1 Restore channel or Al. IMMEDIATEA. Any required SAFETY interlock to operationSYSTEM CHANNEL or ORinterlock function is not A2. IMMEDIATEOPERABLE A.2 ENSURE reactor isSHUTDOWN3.4.5 BasesThe power level scram is provided to ensure that reactor operation stays within thelicensed limits of 1,100 kW, preventing abnormally high fuel temperature. The powerlevel scram is not credited in analysis, but provides defense in depth to assure that thereactor is not operated in conditions beyond the assumptions used in analysis (Chapter4 and 13).The manual scram allows the operator to shut down the system if an unsafe orabnormal condition occurs.The CONTROL ROD (STANDARD) interlock function is to prevent withdrawing controlrods (other than the pulse rod) when the reactor is in the PULSE MODE. This will ensurethe reactivity addition rate during a pulse is limited to the reactivity added by the pulserod.The pulse rod interlock function prevents air from being applied to the transient roddrive when it is withdrawn while disconnected from the control rod to preventinadvertent pulses during STEADY STATE MODE operations. The control rod interlockprevents inadvertent pulses which would be likely to exceed the maximum range of thepower level instruments configured for steady state operations.12/2011TS-24 UT TRIGA II TECHNICAL SPECIFICATIONS3.5 Gaseous Effluent Control3.5.1 ApplicabilityThis specification applies to gaseous effluent in STEADY STATE MODE and PULSE MODE.3.5.2 ObjectiveThe objective is to ensure that exposures to the public resulting from gaseous effluentsreleased during normal operations and accident conditions are within limits and ALARA.3.5.3 SpecificationThe reactor bay HVAC confinement system SHALL provide ventilation to theA reactor bay when particulate continuous air monitor indicates less than 10,000cpmThe reactor bay confinement system will enter CONFINEMENT ISOLATION if theB particulate continuous air monitor is in-service and indicates greater than 10,000cpmC Auxiliary purge system SHALL exhaust from reactor bay pool and in-useexperiment areasReleases of Ar-41 from the reactor bay to an unrestricted environment SHALLNOT exceed 100 Ci per year.TS-2512/2011 TECHNICAL SPECIFICATIONS3.5.4 ActionsCONDITION REQUIRED ACTION COMPLETION TIMEA.1 ENSURE reactor is A.1 IMMEDIATESHUTDOWNORA.2.1 ENSURE auxiliary air A.2.1 IMMEDIATEpurge system isOPEATINGA. The reactor bay HVACconfinement ventilation ANDsystem is not OPERABLE A.3.a IMMEDIATEA.3.b SECURE EXPERIMENToperations if failure couldresult in significant A.3.b IMMEDIATErelease of rad. gases oraerosols.A.3.c IMMEDIATEA.3.c ENSURE no irradiatedfuel handingB.1 ENSURE reactor is B.1 IMMEDIATEB The particulate SHUTDOWNcontinuous air monitor ANDis in service and B.1 IMMEDIATEindicates greater than B.2 SECURE reactor bay10,000 cpm, and thereactor bay confinementsystem is not inCONFINEMENT ANDISOLATION B.3 SECURE the fume/sortinghood12/2011TS-26 UT TRIGA II TECHNICAL SPECIFICATIONSCONDITION REQUIRED ACTION COMPLETION TIMEC.1 ENSURE reactor bay C.1 IMMEDIATEHVAC confinementventilation system isOPERATINGORC.2.a ENSURE reactor is C.2.a IMMEDIATESHUTDOWNC. The auxiliary purgesystem is not OPERABLE C.2.b Secure EXPERIMENT C.2.b IMMEDIATEoperations forEXPERIMENT with failuremodes that could result inthe release of radioactivegases or aerosolsC.2.c ENSURE no irradiated fuel C.2.c IMMEDIATEhandingD Calculated releases of Ar-41 from the reactor bay D. Do not operate. D. IMMEDIATEexhaust plenum exceed100 Ci per year.3.5.5 BasesThe confinement and ventilation system is described in Chapter 9. Routine operationsproduce radioactive gas, principally Argon 41, in the reactor bay. If the confinementsystem is not functioning and the purge system is not operating, radioactive gasses willbuildup in the reactor bay. During this interval, experiment activities that might causeairborne radionuclide levels to be elevated are prohibited.Chapter 13 addresses the maximum hypothetical fission product inventory release.Using unrealistically conservative assumptions, concentrations for a few nuclides ofiodine would be in excess of occupational derived air concentrations for a matter ofhours or days. 90Sr activity available for release from fuel rods previously used at otherfacilities is estimated to be at most about 4 times the ALl. In either case (radio-iodine or-Sr), there is no credible scenario for accidental inhalation or ingestion of the undilutednuclides that might be released from a damaged fuel element. Finally, fuel elementfailure during a fuel handling accident is likely to be observed and mitigatedimmediately.TS-2712/2011 TECHNICAL SPECIFICATIONSThe CAP-88 (Clean Air Act Assessment Package-1988) computer model is a set ofcomputer programs, databases and associated utility programs for estimation of doseand risk from radionuclide emissions to air. CAP-88 is composed of modified versions ofAIRDOS-EPA (Mo79) and DARTAB (ORNL5692). CAP-88 was used to analyze argon 41effluents from the UT TRIGA reactor. Analysis shows 100 Ci per year results in amaximum does to individuals in the effluent plume of 0.142 mrem in a year, well withinthe 10CFR20 limit of 10 mrem/year for stack effluents.12/2011TS-28 UT TRIGA II TECHNICAL SPECIFICATIONS3.6 Limitations on Experiments3.6.1 ApplicabilityThis specification applies to operations in STEADY STATE MODE and PULSE MODE.3.6.2 ObjectivesThese Limiting Conditions for Operation prevent reactivity excursions that might causethe fuel temperature to exceed the safety limit (with possible resultant damage to thereactor), and the excessive release of radioactive materials in the event of anEXPERIMENT failure3.6.3 SpecificationsThe reactivity worth of any individual MOVEABLE EXPERIMENT SHALL NOTexceed $1.00The reactivity worth of any individual SECURED EXPERIMENT SHALL NOT exceed$2.50C The total reactivity worth of all EXPERIMENTS shall not exceed $3.003.6.4 ActionsCONDITION REQUIRED ACTION COMPLETION TIMEA.1 ENSURE the reactor is A.1 IMMEDIATESHUTDOWNA. MOVEABLE EXPERIMENTworth is greater than AND$1.00A.2 Remove the experiment A.2 Prior to continuedoperationsB.1 ENSURE the reactor is B.1 IMMEDIATESHUTDOWNB. SECURED EXPERIMENTworth is greater than AND$2.50B.2 Remove the experiment B.2 Prior to continuedoperationsTS-2912/2011 TECHNICAL SPECIFICATIONSC.1 ENSURE the reactor is C.1 IMMEDIATESHUTDOWNC. Total EXPERIMENT worth ANDis greater than $3.00C.2 Remove the experiment C.2 Prior to continuedoperations3.6.5 BasesChapter 13 demonstrates that pulsed reactivity worth less than 2.8% Ak/k ($4.00) willnot challenge fuel integrity. These limits provide assurance that experiments do notexceed the reactivity analyzed; experiment limits are established lower than analysislimits is used to assure margin for experimental error.12/2011TS-30 UT TRIGA II TECHNICAL SPECIFICATIONS3.7 Fuel Integrity3.7.1 ApplicabilityThis specification applies to operations in STEADY STATE MODE and PULSE MODE.3.7.2 ObjectiveThe objective is to prevent the use of damaged fuel in the UT TRIGA reactor.3.7.3 SpecificationsA Fuel elements in the reactor core SHALL NOT be (1) elongated more than 1/10 in.over manufactured length OR (2) laterally bent more than 1/16 in.B Fuel elements SHALL NOT have visual indications of cladding integrity failure.C Fuel elements in the core SHALL NOT release fission products.3.7.4 ActionsCONDITION REQUIRED ACTION COMPLETION TIMEA. Any fuel element iselongated greater than Do not re-insert the fuel1/10in.over element into the upper core IMMEDIATEmanufactured length, or grid plate.bent laterally greaterthan 1/16 in.B. Fuel elements have Do insert or not re-insert thevisual indication of fuel element into the upper IMMEDIATEcladding integrity failure core grid plate.C.1 SECURE PULSE MODE C.1 IMMEDIATEoperationsC. Fission products are C.2.a Operate in STEADY STATE C.2.a IMMEDIATEdetermined to beMODE only to identifyleaking from fuel the failed elementelements in the core AND C.2.b When theC.2.b Remove the failed element is identifiedelement from serviceTS-3112/2011 TECHNICAL SPECIFICATIONS3.7.5 BasesThe above limits on the allowable distortion of a fuel element have been shown tocorrespond to strains that are considerably lower than the strain expected to causerupture of a fuel element and have been successfully applied at TRIGA installations. Fuelcladding integrity is important since it represents the only process barrier for fissionproduct release from the TRIGA reactor.Lateral bend less than 1/16 in. in adjacent fuel elements assures that there is adequateclearance to prevent element contact during operation.Limitingthe use of fuel elements where cladding has been challenged as specified limitsrelease of fission products to the minimum required for assessing fuel elements.12/2011TS-32 UT TRIGA II TECHNICAL SPECIFICATIONS3.8 Reactor Pool Water3.8.1 ApplicabilityThis specification applies to operations in STEADY STATE MODE, PULSE MODE, andSECURED MODE.3.8.2 ObjectiveThe objective is to set acceptable limits on the water quality, temperature, conductivity,and level in the reactor pool.3.8.3 SpecificationsA Water temperature at the exit of the reactor pool SHALL NOT exceed 110'F(48.9-C)B Water conductivity SHALL be less than or equal to 5 Iimho/cm averaged over 1monthC Water level above the core SHALL be at least 6.5 m from bottom of the poolThe pressure difference between chilled water outlet from the pool heatexchanger and pool water inlet SHALL not exceed 7 kPa (1 psig)3.8.4 ActionsCONDITION REQUIRED ACTION COMPLETION TIMEA.1 ENSURE the reactor is A.1 IMMEDIATESHUTDOWNANDA. Water temperature at A.2 Secure flow through the A.2 IMMEDIATEthe exit of the reactor demineralizerpool exceeds 110°F(48.9°C) ANDA.3 Initiate action to reduce A.3 IMMEDIATEwater temperature to lessthan 110°FTS-3312/2011 TECHNICAL SPECIFICATIONSCONDITION REQUIRED ACTION COMPLETION TIMEB.1 ENSURE the reactor is B.1 IMMEDIATESHUTDOWNB. Water conductivity isgreater than 5 pImho/cmB.2 Restore conductivity to less B.2 Within 1 monththan 5 Vmho/cmC. Water level above the C.1 ENSURE the reactor is C.1 IMMEDIATEcore SHALL be at least SHUTDOWN6.5 m from the bottom ANDof the pool for alloperating conditions C.2 Restore water level C.2 IMMEDIATED.1 ENSURE the reactor is D.1 IMMEDIATESHUTDOWNORD.2 Verify TRUE VALUE is less D.2 IMMEDIATED. The pressure difference thn7ka(psgbetween chilled wateroutlet from the pool ORheat exchanger andpool water inletexed 7ater (1 psi D.3 RESTORE pressure D.3 IMMEDIATEexceeds 7 kPa (1 psig) difference to less than 7 kPa(1 psig)ORD.4 Isolate chill water D.4 IMMEDIATE3.8.5 BasesThe resin used in the mixed bed deionizer limits the water temperature of the reactorpool. Resin in use (as described in Section 5.4) maintains mechanical and chemicalintegrity at temperatures below 110'F (48.9°C). Therefore, thermal hydraulic analysiswas conducted to a maximum pool temperature of 48.9°C, and limiting pooltemperature ensures analysis conditions are met.Maintaining low water conductivity over a prolonged period prevents possiblecorrosion, deionizer degradation, or slow leakage of fission products from degradedcladding. Although fuel degradation does not occur over short time intervals, long-term12/2011TS-34 UT TRIGA II TECHNICAL SPECIFICATIONSintegrity of the fuel is important, and a 4-week interval was selected as an appropriatemaximum time for averaging conductivity values.For normal pool temperature, calculations in Chapter 4 assuming 8.1 and 6.5 m abovethe bottom of the pool demonstrate that the heat flux of the hottest area of the fuel rodgenerating the highest power level in the core during operations is less than the criticalheat flux by a large margin up to the maximum permitted cooling temperatures; marginremains even at temperatures approaching bulk boiling for atmospheric conditions.Therefore, pool levels greater than 6.5 m above the pool floor meet requirements forsafe operation with respect to maximum fuel temperature and thermal hydraulics by awide margin.The principle contributor to radiation dose rates at the pool surface is Nitrogen 16generated in the reactor core and dispersed in the pool. Pool surface radiation doserates from Nitrogen 16 with 6.5 m of water above the core are acceptable.Therefore, a minimum pool level of 6.5 feet above the core is adequate to support thecore cooling and provide shielding.The specified pressure difference assures that any postulated heat exchanger leakagewill not release potentially contaminated water to the chill water system.TS-3512/2011 TECHNICAL SPECIFICATIONS3.9 Retest Requirements3.9.1 ApplicabilityThis specification applies to operations in STEADY STATE MODE and PULSE MODE.3.9.2 ObjectiveThe objective is to ensure Technical Specification requirements are met followingmaintenance or operational activities that occur within surveillance test intervals.3.9.3 SpecificationsMaintenance or operational activities SHALL NOT change, defeat or alter equipment orsystems in a way that prevents the systems or equipment from being OPERABLE orotherwise prevent the systems or equipment from fulfilling the safety basis3.9.4 ActionsCONDITION REQUIRED ACTION COMPLETION TIMEMaintenance or anoperational activity isperformed that has the Perform surveillance Prior to continued,potential to change a normal operation insetpoint, calibration, flow OR STEADY STATE MODErate, or other parameter or PULSE MODEthat is measured or verified Operate only to perform retestin meeting a surveillance oroperability requirement3.9.5 BasesOperation of the UT TRIGA reactor will comply with the requirements of TechnicalSpecifications. This specification ensures that if maintenance might challenge aTechnical Specifications requirement, the requirement is verified prior to resumption ofnormal operations.12/2011TS-36 UT TRIGA II TECHNICAL SPECIFICATIONS4. Surveillance Requirements4.1 Core Reactivity4.1.1 ObjectiveThis surveillance ensures that the minimum SHUTDOWN MARGIN requirements andmaximum excess reactivity limits of section 3.1 are met.4.1.2 SpecificationSURVIELLANCE REQUIREMENTSSURVEILLANCE FREQUENCYSHUTDOWN MARGIN Determination ANNUALANNUALFollowing Insertion ofEXCESS REACTIVITY Determination experiments withmeasurable positivereactivityControl Rod Reactivity Worth determination BIENNIAL4.1.3 BasisExperience has shown verification of the minimum allowed SHUTDOWN MARGIN at thespecified frequency is adequate to assure that the limiting safety system setting is metWhen core reactivity parameters are affected by operations or maintenance, additionalactivity is required to ensure changes are incorporated in reactivity evaluations.Reactivity limits are verified by comparing critical control rod positions to referencevalues. The reference values change with burnup and core configuration. Biennialevaluation of control rod position is adequate, although other activities may result incontrol rod worth determination through retest requirements.TS-3712/2011 TECHNICAL SPECIFICATIONS4.2 PULSE MODE4.2.1 ObjectivesThe verification that the pulse rod position does not exceed a reactivity valuecorresponding to $4.00 assures that the limiting condition for operation is met.4.2.2 SpecificationSURVIELLANCE REQUIREMENTSSURVEILLANCE FREQUENCYENSURE Transient Pulse Rod position corresponds to Prior to pulsingreactivity not greater than $4.00 operations4.2.3 BasisVerifying pulse rod position corresponds to less than or equal to $4.00 ensures that themaximum pulsed reactivity meets the limiting condition for operation.12/2011TS-38 UT TRIGA II TECHNICAL SPECIFICATIONS4.3 MEASURING CHANNELS4.3.1 ObjectivesSurveillances on MEASURING CHANNELS at specified frequencies ensure instrumentproblems are identified and corrected before they can affect operations.4.3.2 SpecificationSURVIELLANCE REQUIREMENTSSURVEILLANCE FREQUENCYReactor power level MEASURING CHANNELCHANNEL TEST DAILYCalorimetric calibration ANNUALCHANNEL CHECK loss of high voltage to required power level DAILYinstrumentsCALIBRATION high voltage to required power level ANNUALinstrumentsPrimary pool water temperature CHANNEL CALIBRATION ANNUALFuel temperature CHANNEL CALIBRATION ANNUALUpper level Area radiation monitorCHANNEL CHECK MONTHLYCHANNEL CALIBRATION ANNUALLower or middle level Area Radiation MonitorCHANNEL CHECK MONTHLYCHANNEL CALIBRATION ANNUALContinuous Air Radiation MonitorCHANNEL CHECK DAILYCHANNEL CALIBRATION ANNUALArgon MonitorCHANNEL CHECK DAILYCHANNEL CALIBRATION (Electronic) SEMIANNUALContinuous Particulate Air MonitorCHANNEL CHECK MONTHLYCHANNEL CALIBRATION ANNUALTS-3912/2011 TECHNICAL SPECIFICATIONSSURVIELLANCE REQUIREMENTSSURVEILLANCE FREQUENCYStartup Count Rate DAILY4.3.3 BasisThe DAILY CHANNEL CHECKS will ensure that the SAFETY SYSTEM and MEASURINGCHANNELS are operable. The required periodic calibrations and verifications will permitany long-term drift of the channels to be corrected.12/2011TS-40 UT TRIGA II TECHNICAL SPECIFICATIONS4.4 Safety Channel and Control Rod Operability4.4.1 ObjectiveThe objectives of these surveillance requirements are to ensure the REACTOR SAFETYSYSTEM will function as required. Surveillances related to safety system MEASURINGCHANNELS ensure appropriate signals are reliably transmitted to the shutdown system;the surveillances in this section ensure the control rod system is capable of providingthe necessary actions to respond to these signals.4.4.2 SpecificationsSURVIELLANCE REQUIREMENTSSURVEILLANCE FREQUENCYManual scram SHALL be tested by releasing partially DAILYwithdrawn CONTROL RODS (STANDARD)CONTROL ROD (STANDARD) drop times SHALL be measuredto have a drop time from the fully withdrawn position of ANNUALless than 1 sec.The control rods SHALL be visually inspected for corrosion BIENNIALand mechanical damage at intervalsCONTROL ROD (STANDARD) position interlock functional SEMIANNUALtestPulse rod interlock functional test SEMIANNUALThe CONTROL ROD (TRANSIENT) rod drive cylinder and theassociated air supply system SHALL be inspected, cleaned, ANNUALand lubricated, as necessary.4.4.3 BasisManual and automatic scrams are not credited in accident analysis, although thesystems function to assure long-term safe shutdown conditions. The manual scram andcontrol rod drop timing surveillances are intended to monitor for potential degradationthat might interfere with the operation of the control rod systems. The functional test ofloss of high voltage to the power level monitoring channels assures that the safetychannels will function on demand.The control rod inspections (visual inspections and transient drive system inspections)are similarly intended to identify potential degradation that lead to control roddegradation or inoperability.TS-4112/2011 TECHNICAL SPECIFICATIONSA test of the interlock that prevents the pulse rod from coupling to the drive in the statemode unless the drive is fully down or square wave mode is being used assures thatpulses will not unintentionally occur. In particular, instrumentation alignment for thepulsing mode causes safety channels to be capable of monitoring pulse power; if pulsingoccurs while the instruments are set to normal, steady state operations, they will not becapable of monitoring peak power.A test of the interlock that prevents standard control rod motion while in the pulsemode assures that the interlock will function as required.The functional checks of the control rod drive system assure the control rod drivesystem operates as intended for any pulsing operations. The inspection of the pulse rodmechanism will assure degradation of the pulse rod drive will be detected prior tomalfunctions.12/2011TS-42 UT TRIGA II TECHNICAL SPECIFICATIONS4.5 Gaseous Effluent Control4.5.1 ObjectivesThese surveillances ensure that routine releases are normal, and (in conjunction withMEASURING CHANNEL surveillances) that instruments will alert the facility if conditionsindicate abnormal releases.4.5.2 SpecificationSURVIELLANCE REQUIREMENTSSURVEILLANCE FREQUENCYCONFINEMENT ISOLATION functional test MONTHLYCONFINEMENT ISOLATION damper inspection ANNUALLYCalculate Ar4l discharge SEMIANNUALLY4.5.3 BasisConfinement isolation functional test frequency is adequate to ensure potential failuresare detected prior to system demand.The annual test is adequate to detect degradation of sealing surfaces.Semiannual calculation of Argon 41 is adequate to ensure that discharge limits are met.TS-4312/2011 TECHNICAL SPECIFICATIONS4.6 Limitations on Experiments4.6.1 ObjectivesThis surveillance ensures that experiments do not have significant negative impact onsafety of the public, personnel or the facility.4.6.2 SpecificationSURVIELLANCE REQUIREMENTSSURVEILLANCE FREQUENCYPrior to inserting a newExperiments SHALL be evaluated and approved prior to experiment for purposesimplementation. other than determinationof reactivity worthInitial insertion of a newMeasure and record experiment worth of the EXPERIMENT experiment where(where the absolute value of the estimated worth is greater absolute value of thethan $0.50). estimated worth isgreater than $0.504.6.3 BasisThese surveillances support determination that the limits of 3.6 are met.Experiments with an absolute value of the estimated significant reactivity worth (greaterthan $0.40) will be measured to assure that maximum experiment reactivity worths aremet. If an absolute value of the estimate indicates less than $0.50 reactivity worth, anyerror less than 100% will result in actual reactivity less than the assumptions used inanalysis for inadvertent pulsing at low power operations in the Safety Analysis Report(13.2.3, Case I).12/2011TS-44 UT TRIGA II TECHNICAL SPECIFICATIONS4.7 Fuel Integrity4.7.1 ObjectiveThe objective is to ensure that the dimensions of the fuel elements remain withinacceptable limits.4.7.2 ApplicabilityThis specification applies to the surveillance requirements for the fuel elements in thereactor core.4.7.3 SpecificationSURVIELLANCE REQUIREMENTSSURVEILLANCEFREQUENCYThe STANDARD FUEL ELEMENTS SHALL be visually inspectedfor corrosion and mechanical damage, and measured forlength and bend500 pulses of magnitudeequal to or greater thana pulse insertion of $3.00ANDFollowing the exceedingof a limited safety systemset point with potentialfor causing degradationApproximately 1/4 of the core SHALL be visually inspected BIENNIALannually for corrosion and mechanical damage such thatFull core inspection complete 4, not to exceed 5, years'4.7.4 BasisThe most severe stresses induced in the fuel elements result from pulse operation ofthe reactor, when fuel to cladding differential expansion occurs and gas pressure. Themagnitude of $3.00 pulses warrants inspection following a sufficient number of cycles.Visual inspection of fuel elements at the specified intervals combined withmeasurements at intervals determined by pulsing as described is considered adequateto identify potential degradation of fuel prior to catastrophic fuel element failure.TS-4512/2011 TECHNICAL SPECIFICATIONS4.8 Reactor Pool WaterThis specification applies to the water contained in the UT TRIGA reactor pool.4.8.1 ObjectiveThe objective is to provide surveillance of reactor primary coolant water quality, poollevel, temperature and (in conjunction with MEASURING CHANNEL surveillances), andconductivity.4.8.2 SpecificationSURVIELLANCE REQUIREMENTSSURVEILLANCE FREQUENCYVerify reactor pool water level above the inlet line vacuum DAILYbreakerVerify reactor pool water temperature channel operable DAILYWEEKLYMeasure reactor Pool water conductivityAt least every 30 days4.9.3 BasesSurveillance of the reactor pool will ensure that the water level is adequate beforereactor operation. Evaporation occurs over longer periods of time, and daily checks areadequate to identify the need for water replacement. Pool water level status (not high,not low) is indicated on the control console.Water temperature must be monitored to ensure that the temperature limit related toresin will not be exceeded, and that the conditions for analysis are maintained., A dailycheck on the instrument prior to reactor operation is adequate to ensure the instrumentis operable when it will be needed.Water conductivity must be checked to ensure that the pool cleanup system isperforming properly and to detect any increase in water impurities. A weekly check isadequate to verify water quality is appropriate and also to provide data useful in trendanalysis. If the reactor is not operated for long periods of time, the requirement forchecks at least every 30 days ensures water quality is maintained in a manner that doesnot permit fuel degradation.12/2011TS-46 UT TRIGA II TECHNICAL SPECIFICATIONS4.9 Retest Requirements4.9.1 ObjectiveThe objective is to ensure that a system is OPERABLE within specified limits before beingused after maintenance or operational activities has been performed.4.9.2 SpecificationSURVIELLANCE REQUIREMENTSSURVEILLANCE FREQUENCYEvaluate potential for maintenance or operational activities Following maintenanceor operational activitiesto affect operability and function of equipment required by or systeof equipmenTechnical Specifications; for standard procedures, this for systems of equipmentevaluation is incorporated in instructions. required by TechnicalSpecificationsPerform surveillance to assure affected function meets Prior to resumption ofrequirements normal operations4.9.3 BasesThis specification ensures that work on systems or components has been properlycarried out and that the system or component has been properly reinstalled orreconnected before reliance for safety is placed on it.TS-4712/2011 TECHNICAL SPECIFICATIONS5. Design Features5.1 Reactor Fuel5.1.1 ApplicabilityThis specification applies to the fuel elements used in the reactor core.5.1.2 ObjectiveThe objective is to ensure that the fuel elements are of such a design and fabricated insuch a manner as to permit their use with a high degree of reliability with respect totheir mechanical integrity.5.1.3 Specification(I) The high-hydride fuel element shall contain uranium-zirconium hydride, clad in0.020 in. of 304 stainless steel. It shall contain nominally 8.5 weight percenturanium which has a maximum nominal enrichment of 20%. There shall be 1.55to 1.80 hydrogen atoms to 1.0 zirconium atom.(2) For the fuel loading process, elements shall be loaded in a close packed arrayexcept for experimental facilities or for single positions occupied by control rodsand a neutron startup source.5.1.4 BasesThese types of fuel elements have a long history of successful use in TRIGA reactors.5.2 Reactor Fuel and Fueled Devices in Storage5.2.1 ApplicabilityThis specification applies to reactor fuel elements in storage5.2.2 ObjectiveThe objective is to ensure fuel elements or fueled devices in storage are maintainedSubcritical in a safe condition.12/2011TS-48 UT TRIGA II TECHNICAL SPECIFICATIONS5.2.3 Specification(1) All fuel elements or fueled devices shall be in a safe, stable geometry;(2) The keff of all fuel elements or fueled devices in storage is less than 0.9;(3) The keff of fuel elements or fueled devices in an approved shippingcontainer will meet the applicable Certificate of Compliancespecifications for keff;(4) Irradiated fuel elements or fueled devices will be stored in an array whichwill permit sufficient natural convection cooling by air or water such thatthe fuel element or fueled device will not exceed design values.5.2.4 BasesThis specification is based on American Nuclear Society standard 15.1, section 5.4.5.3 Reactor Building5.3.1 ApplicabilityThis specification applies to the building that houses the TRIGA reactor facility.5.3.2 ObjectiveThe objective is to ensure that provisions are made to restrict the amount of release ofradioactivity into the environment.5.3.3 Specification(I) The reactor shall be housed in a closed room designed to restrict leakage whenthe reactor is in operation, with HVAC system designed to maintain negativedifferential pressure with respect to adjacent spaces and the environment.(2) The minimum free volume of the reactor room shall be approximately 4120 m3.(3) The reactor bay HVAC confinement ventilation system is capable of exhaustingair or other gases from the reactor room at a minimum of 60 ft. above groundlevel.TS-4912/2011 TECHNICAL SPECIFICATIONS(4) Reactor bay HVAC confinement ventilation system operation is designed toprovide a minimum of 2 changes of reactor bay air per hour.5.3.4 BasesTo control the escape of gaseous effluent, the reactor room contains no windows thatcan be opened. The room air is exhausted through an independent exhaust system, anddischarged above the roof to provide dilution.5.4 Experiments5.4.1 ApplicabilityThis specification applies to the design of experiments.5.4.2 ObjectiveThe objective is to ensure that experiments are designed to meet criteria.5.4.3 Specifications(1) EXPERIMENTS with design reactivity worth greater than $1.00 SHALL besecurely fastened (as defined in Section I, Secured Experiment).(2) Design shall ensure that failure of an EXPERIMENT SHALL NOT lead to a directfailure of a fuel element or of other experiments that could result in ameasurable increase in reactivity or a measurable release of radioactivity dueto the associated failure.(3) EXPERIMENTS SHALL be designed so that they do not cause bulk boiling ofcore water(4) EXPERIMENT design SHALL ensure no interference with control rods orshadowing of reactor control instrumentation.(5) EXPERIMENT design shall minimize the potential for industrial hazards, suchas fire or the release of hazardous and toxic materials.(6) Where the possibility exists that the failure of an EXPERIMENT (except fueledEXPERIMENTS) could release radioactive gases or aerosols to the reactor bayor atmosphere, the quantity and type of material shall be limited such that12/2011TS-50 UT TRIGA II TECHNICAL SPECIFICATIONSthe airborne concentration of radioactivity is less than 1,000 times theDerived Air Concentration.For in-core samples a decay time of five minutes following irradiation may beused in radioactive inventory calculations to account for processing prior topotential exposure.(7) Each fueled experiment shall be limited such that the total inventory of (1)radioactive iodine isotopes 131 through 135 in the experiment is not greaterthan 9.32E5 IVCi, and (2) radioactive strontium is not greater than 9.35E4 VCi.Alternate calculations may be accomplished to demonstrate equivalent timesfor protective actions based on DAC limits for specific experiments, if desired.These limits do not apply to TRIGA fuel elements used in experiments asmaximum hypothetical accident analysis applies. For in-core samples a decaytime of five minutes following irradiation to account may be used incalculations.(8) The following assumptions shall be used in experiment design:a. If effluents from an experimental facility exhaust through a hold-up tankwhich closes automatically at a high radiation level, at least 10% of thegaseous activity or aerosols produced will escape.b. If effluents from an experimental facility exhaust through a filterinstallation designed for greater than 99% efficiency for 0.3 micronparticles, at least 10% of the aerosols produced will escape.c. For materials whose boiling point is above 1302F and where vaporsformed by boiling this material could escape only through an undisturbedcolumn of water above the core, at least 10% of these vapors will escape.(9) Use of explosive solid or liquid material with a National Fire ProtectionAssociation Reactivity (Stability) index of 2, 3, or 4 in the reactor pool orbiological shielding SHALL NOT exceed the equivalent of 25 milligrams of TNTwithout prior NRC approval.5.4.4 BasisDesigning the experiment to reactivity and thermal-hydraulic conditions ensures thatthe experiment is not capable of breaching fission product barriers or interfering withthe control systems (interferences from other -than reactivity -effects with the controland safety systems are also prohibited). Design constraints on industrial hazardsTS-5112/2011 TECHNICAL SPECIFICATIONSensure personnel safety and continuity of operations. Design constraints limiting therelease of radioactive gasses prevent unacceptable personnel exposure during off-normal experiment conditions.A Derived Air Concentration assumes a 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> per year exposure; if exposure iscontrolled to a specific time limit, such as time required for recognizing the situationand evacuating, limiting values for an experiment can be higher than a DAC.Limits on radioiodine and radioactive strontium in fueled experiments permits a 1 hourevacuation time for releases of radioiodine and a 2-hour evacuation time for releases ofradioactive strontium based on a TRIGA fuel distribution of the radioisotopes fromfission of 235U.12/2011TS-52 UT TRIGA II TECHNICAL SPECIFICATIONS6. Administrative Controls6.1 Organization and Responsibilities of PersonnelThis chapter describes and discusses the Conduct of Operations at the University ofTexas TRIGA. The Conduct of Operations involves the administrative aspects of facilityoperations, the facility emergency plan, the security plan, the Reactor Operatorselection and requalification plan, and environmental reports. License is used inChapter 12 in reference to reactor operators and senior reactors subject to 10CFR50.55requirements.a) StructureUniversity AdministrationFig. 1 illustrates the organizational structure that is applied to the management andoperation of the University of Texas and the reactor facility. Responsibility for the safeoperation of the reactor facility is a function of the management structure of Fig. 11.These responsibilities include safeguarding the public and staff from undue radiationexposures and adherence to license or other operation constraints. Functionalorganization separates the responsibilities of academic functions and businessfunctions. The office of the President administers these activities and other activitiesthrough several vice presidents.Office of the PresidentThe University of Texas at AuxtinEExecu-tive Vice President Vice Presiden; foran~d Provost U niv ersft. Ope rat iontsAssociate V~ie Presid.en,Heal~eh y sandevuiRadiet.ion Sa f,; frce .......... Radiation Sa fe y CormmtteeAssociate Directorof N RTLReactor S rpemio Health PhysicistFigure 6.1, Organzational StructureI"Standard for Administrative Controls" ANSI/ANS -15.18 1979TS-5312/2011 TECHNICAL SPECIFICATIONSNETL Facility AdministrationThe facility administrative structure is shown in Fig. 2. Facility operation staff is anorganization of a director and at least four full time equivalent persons. This staff of fourprovides for basic operation requirements. Four typical staff positions consist of anassociate director, a reactor supervisor, a reactor operator, and a health physicist. Oneor more of the listed positions may also include duties typical of a research scientist. Thereactor supervisor, health physicist, and one other position are to be full time. One fulltime equivalent position may consist of several part-time persons such as assistants,technicians and secretaries. Faculty, students, and researchers supplement theorganization. Titles for staff positions are descriptive and may vary from actualdesignations. Descriptions of key components of the organization follow.NETL DirectorNETL.Associate DirectorReactor Supervisor NEIL Lab Manager Health Physicist ----------RaorOperators Tehncal Suppo7rt Lab Assisstants Rad-Con TechnclarsFigure 2, NETL Facility Administrationb) Functional ResponsibilityVice President and ProvostResearch and academic educational programs are administered through the Office ofthe Executive Vice President and Provost. Separate officers assist with theadministration of research activities and academic affairs with functions delegated tothe Dean of the College of Engineering and Chairman of the Mechanical EngineeringDepartment.Vice President for University OperationsUniversity operations activities are administered through the Office of the VicePresident for Operations. This office is responsible for multiple operational functions ofthe University including university support programs, human resources, campus safetyand security, campus real estate, and campus planning and facilities management.12/2011TS-54 UT TRIGA II TECHNICAL SPECIFICATIONSAssociate Vice President Campus Safety and SecurityThe associate vice president for campus safety and security oversees multiple aspects ofsafety and security on campus including environmental health and safety, campuspolice, parking and transportation, fire prevention, and emergency preparedness.Director of Nuclear Engineering Teaching LaboratoryNuclear Engineering Teaching Laboratory programs are directed by a senior classifiedstaff member or faculty member. The director oversees strategic guidance of theNuclear Engineering Teaching Laboratory including aspects of facility operations,research, and service work. The director must interact with senior University of Texas atAustin management regarding issues related to the Nuclear Engineering TeachingLaboratory.Associate Director of Nuclear Engineering LaboratoryThe Associate Director performs the day to day duties of directing the activities of thefacility. The Associate Director is knowledgeable of regulatory requirements, licenseconditions, and standard operating practices. The associate director will also beinvolved in soliciting and carrying out research utilizing the reactor and other specializedequipment at the Nuclear Engineering Teaching Laboratory.Reactor Oversight CommitteeThe Reactor Oversight Committee is established through the Office of the Dean of theCollege of Engineering of The University of Texas at Austin. Broad responsibilities of thecommittee include the evaluation, review, and approval of facility standards for safeoperation.The Dean shall appoint at least three members to the Committee that represent a broadspectrum of expertise appropriate to reactor technology. The committee will meet atleast twice each calendar year or more frequently as circumstances warrant. TheReactor Oversight Committee shall be consulted by the Nuclear Engineering TeachingLaboratory concerning unusual or exceptional actions that affect administration of thereactor program.Radiation Safety OfficerA Radiation Safety Officer acts as the delegated authority of the Radiation SafetyCommittee in the daily implementation of policies and practices regarding the safe useof radioisotopes and sources of radiation as determined by the Radiation SafetyCommittee. The Radiation Safety Program is administered through the UniversityTS-5512/2011 TECHNICAL SPECIFICATIONSEnvironmental Health and Safety division. The responsibilities of the Radiation SafetyOfficer are outlined in The University of Texas at Austin Manual of Radiation Safety.Radiation Safety CommitteeThe Radiation Safety Committee is established through the Office of the President ofThe University of Texas at Austin. Responsibilities of the committee are broad andinclude all policies and practices regarding the license, purchase, shipment, use,monitoring, disposal, and transfer of radioisotopes or sources of ionizing radiation atThe University of Texas at Austin.The President shall appoint at least three members to the Committee and appoint oneas Chairperson. The Committee will meet at least once each year on a called basis or asrequired to approve formally applications to use radioactive materials. The RadiationSafety Committee shall be consulted by the University Safety Office concerning anyunusual or exceptional action that affects the administration of the Radiation SafetyProgram.Reactor SupervisorReactor operation at the Nuclear Engineering Teaching Laboratory is directed by aReactor Supervisor. Responsibilities of the Reactor Supervisor include control of licensedocumentation, reactor operation, equipment maintenance, experiment operation, andinstruction of persons with access to laboratory areas.Activities of reactor operators with USNRC licenses will be subject to the direction of aperson with a USNRC senior operator license. The Reactor Supervisor shall be qualifiedas a senior operator. This person is to be knowledgeable of regulatory requirements,license conditions, and standard operating practices.Health PhysicistRadiological safety of the Nuclear Engineering Teaching Laboratory is monitored by ahealth physicist, who will be knowledgeable of the facility radiological hazards.Responsibilities of the health physicist will include calibration of radiation detectioninstruments, measurements of radiation levels, control of radioactive contamination,maintenance of radiation records, and assistance with other facility monitoringactivities.Activities of the health physicist will depend on two conditions. One condition will bethe normal operation responsibilities determined by the director of the facility. Asecond condition will be communications specified by the radiation safety officer. Thiscombination of responsibility and communication provides for safety programimplementation by the director, but establishes independent review. The health12/2011TS-56 UT TRIGA II TECHNICAL SPECIFICATIONSphysicist's activities will meet the requirements of the director and the policies of anindependent university safety organization.Laboratory ManagerLaboratory operations and research support is provided by a designated LaboratoryManager. The function is typically combined with the Health Physicist position.Reactor OperatorsReactor operators (and senior reactor operators) are licensed by the USNRC to operatethe UT TREIGA II nuclear research reactor. University staff and/or students may beemployed as reactor operators.Technical SupportStaff positions supporting various aspects of facility operations are assigned as required.Radiological Controls TechniciansRadiological Controls Technicians are supervised by the Health Physicist to performradiological controls and monitoring functions. Radiological Controls Technicians aregenerally supported as Undergraduate Research Assistant positions.Laboratory AssistantsLaboratory Assistants are supervised by the Laboratory Manager to perform laboratoryoperations and analysis. Laboratory Assistants are generally supported asUndergraduate Research Assistant positions.c) StaffingOperation of the reactor and activities associated with the reactor, control system,instrument system, radiation monitoring system, and engineered safety features will bethe function of staff personnel with the appropriate training and certification2.Whenever the reactor is not secured, the reactor shall be under the direction of a(USNRC licensed) Senior Operator who is designated as Reactor Supervisor. TheSupervisor may be on call if capable of arriving at the facility within thirty minutes andcognizant of reactor operations. The Reactor Supervisor shall directly supervise:2 "Selection and Training of Personnel for Research Reactors", ANSI/ANS -15.4 -1970 (N380)TS-5712/2011 TECHNICAL SPECIFICATIONSa. All fuel element or control rod relocations or installations within the reactor coreregion, and subsequent INITIAL STARTUP and approach to power.b. Relocation or installation of any experiment in the core region with a reactivityworth of greater than one dollar, and subsequent INITIAL STARTUP and approachto power.c. Recovery from an unscheduled shutdown or significant power reductions,d. Any INITIAL STARTUP and approach to power following modifications to reactorsafety or control rod drive systems.Whenever the reactor is not secured, a (USNRC licensed) Reactor Operator (or SeniorReactor Operator) who meets requirements of the Operator Requalification Programshall be at the reactor control console, and directly responsible for controlmanipulations. All activities that require the presence of licensed operators will alsorequire the presence in the facility complex of a second person capable of performingprescribed written instructions.Only the Reactor Operator at the controls or personnel authorized by, and under directsupervision of, the Reactor Operator at the controls shall manipulate the controls.Whenever the reactor is not secured, operation of equipment that has the potential toaffect reactivity or power level shall be manipulated only with the knowledge andconsent of the Reactor Operator at the controls. The Reactor Operator at the controlsmay authorize persons to manipulate reactivity controls who are training either as (1) astudent enrolled in academic or industry course making use of the reactor, (2) to qualifyfor an operator license, or (3) in accordance the approved Reactor Operatorrequalification program.Whenever the reactor is not secured, a second person (i.e., in addition to the reactoroperator at the control console) capable of initiating the Reactor Emergency Plan will bepresent in the NETL building. Unexpected absence of this second person for greaterthan two hours will be acceptable if immediate action is taken to obtain a replacement.Staffing required for performing experiments with the reactor will be determined by aclassification system specified for the experiments. Requirements will range from thepresence of a certified operator for some routine experiments to the presence of asenior operator and the experimenter for other less routine experiments.6.2 Review and AuditThe review and audit process is the responsibility of the Reactor Oversight Committee(ROC).12/2011TS-58 UT TRIGA II TECHNICAL SPECIFICATIONSComposition and QualificationsThe ROC shall consist of at least three (3) members appointed by the Dean of theCollege of Engineering that are knowledgeable in fields which relate to nuclear safety.The university radiological safety officer shall be a member or an ex-officio member. Thecommittee will perform the functions of review and audit or designate a knowledgeableperson for audit functions.Charter and RulesThe operations of the ROC shall be in accordance with an established charter, includingprovisions for:a. Meeting frequency (at least twice each year, with approximately 4-8 monthfrequency).b. Quorums (not less than one-half the membership where the operating staff doesnot contribute a majority).c. Dissemination, review, and approval of minutes.d. Use of subgroups.Review FunctionThe responsibilities of the Reactor Safeguards Committee to shall include but are notlimited to review of the following:a. All new procedures (and major revisions of procedures) with safety significanceb. Proposed changes or modifications to reactor facility equipment, or systemshaving safety significancec. Proposed new (or revised) experiments, or classes of experiments, that couldaffect reactivity or result in the release of radioactivityd. Determination of whether items a) through c) involve unreviewed safetyquestions, changes in the facility as designed, or changes in TechnicalSpecifications.e. Violations of Technical Specifications or the facility operating licenseef. Violations of internal procedures or instruction having safety significanceg. Reportable occurrencesh. Audit reportsMinor changes to procedures and experiments that do not change the intent and do notsignificantly increase the potential consequences may be accomplished following reviewand approval by a senior reactor operator and independently by one of the ReactorSupervisor, Associate Director or Director. These changes should be reviewed at thenext scheduled meeting of the Reactor Oversight Committee.TS-5912/2011 TECHNICAL SPECIFICATIONSAudit FunctionThe audit function shall be a selected examination of operating records, logs, or otherdocuments. Audits will be by a Reactor Oversight Committee member or by anindividual appointed by the committee to perform the audit. The audit should be by anyindividual not directly responsible for the records and may include discussions withcognizant personnel or observation of operations. The following items shall be auditedand a report made within 3 months to the Director and Reactor Committee:a. Conformance of facility operations with license and technical specifications atleast once each calendar year.b. Results of actions to correct deficiencies that may occur in reactor facilityequipment, structures, systems, or methods of operation that affect safety atleast once per calendar year.c. Function of the retraining and requalification program for reactor operators atleast once every other calendar year.d. The reactor facility emergency plan and physical security plan, and implementingprocedures at least once every other year.6.3 ProceduresWritten procedures shall govern many of the activities associated with reactoroperation. Activities subject to written procedures will include:a. Startup, operation, and shutdown of the reactorb. Fuel loading, unloading, and movement within the reactor.c. Control rod removal or replacement.d. Routine maintenance, testing, and calibration of control rod drives and othersystems that could have an effect on reactor safety.e. Administrative controls for operations, maintenance, conduct of experiments,and conduct of tours of the Reactor Facility.f. Implementing procedures for the Emergency Plan or Physical Security Plan.Written procedures shall also govern:a. Personnel radiation protection, in accordance with the Radiation ProtectionProgram as indicated in Chapter 11b. Administrative controls for operations and maintenancec. Administrative controls for the conduct of irradiations and experiments thatcould affect core safety or reactivity12/2011TS-60 UT TRIGA II TECHNICAL SPECIFICATIONSA master Procedure Control procedure specifies the process for creating, changing,editing, and distributing procedures. Preparation of the procedures and minormodifications of the procedures will be by certified operators. Substantive changes ormajor modifications to procedures, and new prepared procedures will be submitted tothe Reactor Oversight Committee for review and approval. Temporary deviations fromthe procedures may be made by the reactor supervisor or designated senior operatorprovided changes of substance are reported for review and approval.Proposed experiments will be submitted to the reactor oversight committee for reviewand approval of the experiment and its safety analysis3, as indicated in Chapter 10.Substantive changes to approved experiments will require re-approval whileinsignificant changes that do not alter experiment safety may be approved by a senioroperator and independently one of the following, Reactor Supervisor, AssociateDirector, or Director. Experiments will be approved first as proposed experiments forone time application, and subsequently, as approved experiments for repeatedapplications following a review of the results and experience of the initial experimentimplementation.6.4 Review of Proposals for Experimentsa ) All proposals for new experiments involving the reactor shall be reviewed withrespect to safety in accordance with the procedures in (b) below and on thebasis of criteria in (c) below.b) Procedures:1. Proposed reactor operations by an experimenter are reviewed by theReactor Supervisor, who may determine that the operation is described by apreviously approved EXPERIMENT or procedure. If the Reactor Supervisordetermines that the proposed operation has not been approved by theReactor Oversight Committee, the experimenter shall describe the proposedEXPERIMENT in written form in sufficient detail for consideration of safetyaspects. If potentially hazardous operations are involved, proposedprocedures and safety measures including protective and monitoringequipment shall be described.2. The scope of the EXPERIMENT and the procedures and safety measures asdescribed in the approved proposal, Including any amendments orconditions added by those reviewing and approving it, shall be binding onthe experimenter and the OPERATING personnel. Minor deviations shall beallowed only in the manner described in Section 6 above. Recordedaffirmative votes on proposed new or revised experiments or procedures3 ANSI/ANS 15.6, op. cit.TS-6112/2011 TECHNICAL SPECIFICATIONSmust indicated that the Committee determines that the proposed actionsdo not involve changes in the facility as designed, changes in TechnicalSpecifications, changes that under the guidance of 10 CFR 50.59 requireprior approval of the NRC, and could be taken without endangering thehealth and safety of workers or the public or constituting a significanthazard to the integrity of the reactor core.3. Transmission to the Reactor Supervisor for scheduling.c) Criteria that shall be met before approval can be granted shall include:1. The EXPERIMENT must meet the applicable Limiting Conditions for Operationand Design Description specifications.2. It must not involve violation of any condition of the facility license or ofFederal, State, University, or Facility regulations and procedures.3. The conduct of tests or experiments not described in the safety analysisreport (as updated) must be evaluated in accordance with 10 CFR 50.59 todetermine if the test or experiment can be accomplished without obtainingprior NRC approval via license amendment pursuant to 10 CFR Sec. 50.90.4. In the safety review the basic criterion is that there shall be no hazard to thereactor, personnel or public. The review SHALL determine that there isreasonable assurance that the experiment can be performed with nosignificant risk to the safety of the reactor, personnel or the public.6.5 Operator RequalificationAn NRC approved UT TRIGA Requalification Plan is in place to maintain training andqualification of reactor operators and senior reactor operators. License qualification bywritten and operating test, and license issuance or removal, are the responsibility of theU.S. Nuclear Regulatory Commission. No rights of the license may be assigned orotherwise transferred and the licensee is subject to and shall observe all rules,regulations and orders of the Commission. Requalification training maintains the skillsand knowledge of operators and senior operators during the period of the license.Training also provides for the initial license qualification.6.6 Emergency Plan and ProceduresAn NRC approved Emergency Plan following the general guidance set forth in ANSI/ANS15.16, Emergency Planning for Research Reactors is in place. The plan specifies twoaction levels, the first level being a locally defined Non-Reactor Specific Event, and thesecond level being the lowest level FEMA classification, a Notification of Unusual Event.12/2011TS -62 UT TRIGA II TECHNICAL SPECIFICATIONSProcedures reviewed and approved by the Reactor Oversight Committee areestablished to manage implementation of emergency response.6.7 Physical Security PlanAn NRC approved Security Plan Security Plan is in place. The plan incorporatescompensatory measures implemented following security posture changes initiated post9/11. The Plan and portions of the procedures are classified as Safeguards Information.Security procedures implementing the plan, approved by the Reactor OversightCommittee, are established.6.8 Action To Be Taken In The Event A Safety Limit Is ExceededIn the event that a Safety Limit is not met,a. The reactor shall be shutdown and secured.b. The Reactor Supervisor, Associate Director, and Director shall be notifiedc. The safety limit violation shall be reported to the Nuclear RegulatoryCommission within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone, confirmed via written statement byemail, fax or telegraphd. A safety limit violation report shall be prepared within 14 days of the event todescribe:1. Applicable circumstances leading to the violation including (where known)cause and contributing factors2. Effect of the violation on reactor facility components, systems, andstructures3. Effect of the violation on the health and safety of the personnel and thepublic4. Corrective action taken to prevent recurrencee. The Reactor Oversight Committee shall review the report and any followupreportsf. The report and any followup reports shall be submitted to the NuclearRegulatory Commission.g. Operations shall not resume until the USNRC approves resumption.6.9 Action To Be Taken In The Event Of A Reportable Occurrencea ) A reportable occurrence is any of the following conditions:1. Any actual safety system setting less conservative than specified in Section2.2, Limiting Safety System Settings;2. VIOLATION OF SL, LSSS OR LCO;TS-6312/2011 TECHNICAL SPECIFICATIONSNOTESViolation of an LSSS or LCO occurs through failure to comply with an "Action"statement when "Specification"is not met; failure to comply with the"Specification" is not by itself a violation.Surveillance Requirements must be met for allequipment/components/conditions to be considered operable.Failure to perform a surveillance within the required time interval or failure ofa surveillance test shall result in the /component/condition being inoperable3. Incidents or conditions that prevented or could have prevented theperformance of the intended safety functions of an engineered safetyfeature or the REACTOR SAFETY SYSTEM;4. Release of fission products from the fuel that cause airborne contaminationlevels in the reactor bay to exceed 10CFR20 limits for releases to unrestrictedareas;5. An uncontrolled or unanticipated change in reactivity greater than $1.00;6. An observed inadequacy in the implementation of either administrative orprocedural controls, such that the inadequacy has caused the existence ordevelopment of an unsafe condition in connection with the operation of thereactor;b) In the event of a reportable occurrence, as defined in the TechnicalSpecifications, and in addition to the reporting requirements,1. The Reactor Supervisor, the Associate Director and the Director shall benotified2. If a reactor shutdown is required, resumption of normal operations shall beauthorized by the Associate Director or Director3. The event shall be reviewed by the Reactor Oversight Committee during anormally scheduled meeting6.10 Plant Operating RecordsRecords of the following activities shall be maintained and retained for the periodsspecified below4.The records may be in the form of logs, data sheets, electronic files, or4 "Records and Reports for Research Reactors", ANSI/ANS -15.3-1974 (N399).12/2011TS-64 UT TRIGA II TECHNICAL SPECIFICATIONSother suitable forms. The required information may be contained in single or multiplerecords, or a combination thereof.Lifetime RecordsLifetime records are records to be retained for the lifetime of the reactor facility. (Note:Applicable annual reports, if they contain all of the required information, may be usedas records in this section.)a. Gaseous and liquid radioactive effluents released to the environs.b. Offsite environmental monitoring surveys required by Technical Specifications.c. Events that impact or effect decommissioning of the facility.d. Radiation exposure for all personnel monitored.e. Updated drawings of the reactor facility.Five Year PeriodRecords to be retained for a period of at least five years or for the life of the componentinvolved whichever is shorter.a. Normal reactor facility operation (supporting documents such as checklists, logsheets, etc. shall be maintained for a period of at least one year).b. Principal maintenance operations.c. Reportable occurrences.d. Surveillance activities required by technical specifications.e. Reactor facility radiation and contamination surveys where required byapplicable regulations.f. Experiments performed with the reactor.q. Fuel inventories, receipts, and shipments.h. Approved changes in operating procedures.+/-. Records of meeting and audit reports of the review and audit group.TS-6512/2011 TECHNICAL SPECIFICATIONSOne Training CycleTraining records to be retained for. at least one license cycle are the requalificationrecords of licensed operations personnel. Records of the most recent complete cycleshall be maintained at all times the individual is employed.6.11 Reporting RequirementsThis section describes the reports required to NRC, including report content, timing ofreports, and report format. Refer to section 12.4 above for the reporting requirementsfor safety limit violations, radioactivity releases above allowable limits, and reportableoccurrences. All written reports shall be sent within prescribed intervals to the UnitedStates Nuclear Regulatory Commission, Washington, D.C., 20555, Attn: DocumentControl Desk.Operating ReportsRoutine annual reports covering the activities of the reactor facility during the previouscalendar year shall be submitted to licensing authorities within three months followingthe end of each prescribed year. Each annual operating report shall include thefollowing information:a. A narrative summary of reactor operating experience including the energyproduced by the reactor or the hours the reactor was critical, or both.b. The unscheduled shutdowns including, where applicable, corrective action takento preclude recurrence.c. Tabulation of major preventive and corrective maintenance operations havingsafety significance.d. Tabulation of major changes in the reactor facility and procedures, andtabulation of new tests or experiments, or both, that are significantly differentfrom those performed previously, including conclusions that no new orunanalyzed safety questions were identified.e. A summary of the nature and amount of radioactive effluents released ordischarged to the environs beyond the effective control of the owner-operatoras determined at or before the point of such release or discharge. The summaryshall include, to the extent practicable, an estimate of individual radionuclidespresent in the effluent. If the estimated average release after dilution ordiffusion is less than 25% of the concentration allowed or recommended, astatement to this effect is sufficient.12/2011TS-66 UT TRIGA II TECHNICAL SPECIFICATIONSf. A summarized result of environmental surveys performed outside the facility.g. A summary of exposures received by facility personnel and visitors where suchexposures are greater than 25% of that allowed or recommended.Other or Special ReportsThere shall be a report not later than the following working day by telephone andconfirmed in writing by facsimile or similar conveyance of any reportable occurrenceidentified in 6.9.There shall be a written report describing the circumstances of any reportableoccurrence identified in 6.9 within 14 days of occurrence.There shall be a written report within 30 days of:a. Permanent changes in the facility organization involving Director or Supervisor.b. Significant changes in the transient or accident analysis as described in the SafetyAnalysis Report.TS-6712/2011