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| number = ML12156A196
| number = ML12156A196
| issue date = 01/17/2012
| issue date = 01/17/2012
| title = University of Texas at Austin - Redacted Supplement to Application Containing Table of Contents, Chapter 4 and Chapter 9 of the Safety Analysis Report (TAC No. ME7694)
| title = University of Texas at Austin - Redacted Supplement to Application Containing Table of Contents, Chapter 4 and Chapter 9 of the Safety Analysis Report
| author name = Biegalski S
| author name = Biegalski S
| author affiliation = Univ of Texas - Austin
| author affiliation = Univ of Texas - Austin
| addressee name = Lising A J
| addressee name = Lising A
| addressee affiliation = NRC/Document Control Desk, NRC/NRR/DPR
| addressee affiliation = NRC/Document Control Desk, NRC/NRR/DPR
| docket = 05000602
| docket = 05000602
| license number = R-129
| license number = R-129
| contact person = Lising A J
| contact person = Lising A
| case reference number = TAC ME7694
| case reference number = TAC ME7694
| document type = Letter, License-Application for (Amend/Renewal/New) for DKT 30, 40, 70
| document type = Letter, License-Application for (Amend/Renewal/New) for DKT 30, 40, 70
| page count = 97
| page count = 97
| project = TAC:ME7694
| stage = Supplement
}}
}}
=Text=
{{#Wiki_filter:UNIVERSITY OF TEXAS AT AUSTIN RESEARCH REACTOR LICENSE NO. R-129 DOCKET NO. 50-602 UNIVERSITY OF TEXAS AT AUSTIN LICENSE RENEWAL APPLICATION JANUARY 17, 2012 REDACTED VERSION*
SECURITY-RELATED INFORMATION REMOVED
*REDACTED TEXT AND FIGURES BLACKED OUT OR DENOTED BY BRACKETS
DepanmL  aoto i""hanical Engineering THE UNIVERSITY OF TEKAS AT AUSTIN Nuclearf.'5i*eerinig    waaing taboratory AtArin, "7Txas78758 5.I 2-232-5370 -FAX 512-471- -589- htp,'/wu'A me.nrexas.edul/.-nel/
ATTN: Document Control Desk, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001 Allan Jason Lising Project Manager Division of Policy and Rulemaking Research and Test Reactors Licensing Branch January17, 2012
==SUBJECT:==
Docket No. 50-602, Information Supplementing Request for Renewal of Facility Operating License R-129 (TAC ME 7694)
==REFERENCE:==
(1) ML110040316 (2) Letter, 12/12/2011 Docket No. 50-602, Request for Renewal of FacilityOperating License R-129 Sir:
In accordance with USNRC direction (ADAMS ML110040316), a request for renewal of Facility Operating License R-129 (Docket 50-602) was submitted on 12/12/2011. The attached material provides minor editorial corrections and clarification of three chapters previously submitted of the Safety Analysis Report and the Technical Specifications. An additional item is included to support review of the proposed Technical Specifications. In summary:
* Table of Contents (reflecting updates)
* Chapter 4, additional figures are provided to better describe the control rod drive mechanisms, and the section on thermal hydraulic analysis was substantially augmented.
* Chapter 9, operation of the auxiliary purge system and the confinement isolation was revised.
* Chapter 12, the responsibilities of the Senior Reactor Operator was rewritten to emphasize the role of the Supervisor in reactor operations.
* Technical Specifications, editorial changes and various improvements were made.
* Technical Specifications review material: a tabulation of the current Technical Specifications was prepared with a comparison to the proposed, new Technical Specifications.
Your attention in this matter is greatly appreciated, I declare under penalty of perjury that the foregoing is true and correct.
0I THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR              0*:.
000            l        01/2012 SAFETY ANALYSIS REPORT Table of Contents Section                                                                                Page
: 1. THE FACILITY                                                                        1-1 1.1 Introduction                                                                        1-1 1.2 Summary and conclusions on principle safety considerations                          1-2 1.3 General description of the facility                                                1-3 A. Site                                                              1-3 B. Building                                                          1-3 C. Reactor                                                          1-3 C.1 Reactor Core.                                    1-5 C.2 Reactor Reflector.                              1-5 D. Reactor Control.                                                  1-6 E. Experiment Facilities.                                            1-6 E.1 Upper Grid Plate 7L and 3L Facilities            1-6 E.2 Central Thimble                                  1-6 E.3 Rotary Specimen Rack (RSR)                      1-6 E.4 Pneumatic Tubes                                  1-7 E.5 Beam Port Facilities                            1-7 E.5 (1) Beam Port 1 (BP1)                            1-7 E.5 (2) Beam Port 2 (BP2)                            1-8 E.5 (3) Beam Port 3 (BP3)                            1-9 E.5 (4) Beam Port 4 (BP4)                          1-10 E.5 (5) Beam Port 5 (BPS)                          1-10 F Other Experiment and Research Facilities                          1-10 1.4 Overview of shared facilities and equipment                                        1-10 1.4.3 Reference the other facilities operating history, safety and reliability        1-10 1.5 Summary of operations                                                              1-12 1.6 Compliance with NWPA of 1982                                                      1-12 1.7 Facility history & modifications                                                  1-13 2.0 SITE DESCRIPTION                                                                    2-1 2.1 GENERAL LOCATION AND AREA                                                          2-1 2.2 POPULATION AND EMPLOYMENT                                                          2-7 2.3 CLIMATOLOGY                                                                        2-11 2.4 GEOLOGY                                                                            2-14 2.5 SEISMOLOGY                                                                        2-22 2.6 HYDROLOGY                                                                          2-22 2.7 HISTORICAL                                                                        2-27 3.0 DESIGN OF SYSTEMS, STRUCTURES AND COMPONENTS                                        3-1 3.1 Design Criteria for Structures, Systems and Components for Safe Reactor Operation  3-2 3.1.1 Fuel Moderator Elements                                        3-2
SAFETY ANALYSIS REPORT, TABLE OF CONTENTS                                  01/2012 Table of Contents Section                                                                    Page 3.1.2 Control Rods                                        3-3 3.1.3 Core and structural Support                          3-4 3.1.4 Pool and Pool Support Systems                        3-4 3.1.5 Biological Shielding                                3-4 3.1.6 NETL Building/Reactor Bay                            3-5 A. Building                              3-6 B. Reactor Bay                          3-7 3.1.7 Ventilation Systems                                  3-7 3.1.8 Instruments and Controls                            3-8 3.1.9 Sumps and Drains                                    3-8 3.2 Meteorological Damage                                                  3-9 3.3 Water Damage                                                            3-9 3.4 Seismic Damage                                                        3-10 A. Core and structural Support                            3-10 B. Pool and pool cooling                                  3-10 C. Building                                              3-10 4.0 Reactor                                                                4-1 4.1 Summary description                                                    4-1 4.2 Reactor Core                                                            4-1 4.2.1 Reactor Fuel                                        4-2 A. Fuel matrix                            4-2 (1) Fabrication                        4-3 (2) Physical Properties                4-4 (3) Operational Properties            4-7 (4) Neutronic Properties              4-7 (5) Fuel Morphology & Outgassing      4-8 (6) Zr water reaction                  4-9 (7) Mechanical Effects                4-9 (8) Fission Product Release          4-10 B. Cladding                              4-10 4.2.2 Control Rods and Drive Mechanisms                  4-11 A. Control Rods                          4-13 B. Standard Control Rod Drives          4-16 C. Transient Control Rod Drive          4-17 D. Control Functions                    4-19 E. Evaluation of the Control Rod System  4-20 4.2.3 Neutron Moderator and Reflector (Core Structure)    4-20 A. Upper grid plate                      4-20 B. Reflector                            4-23 ii
THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT
                                                                *Ilk, I 01/2012 1/01 Table of Contents Section                                                                          Page (1) Radial Reflector                          4-23 (2) Graphite Rods.                            4-24 (3) Axial Reflector                          4-24 C. Lower grid plate                            4-24 4.2.4 Neutron Startup Source                                    4-26 4.2.5 Core support structure                                    4-26 A. Core Support Platform                        4-26 B. Safety plate                                4-27 4.3 Reactor Pool                                                                  4-28 4.4 Biological Shield                                                            4-30 4.5 Nuclear Design                                                                4-32 4.5.1 Normal Operating Conditions                              4-32 4.5.2 Nominal Reactivity Worth Values                          4-33 4.5.3 Reactor Core Physics                                      4-32 A. Reference Calculations                      4-34 B. Prompt Negative Temperature Coefficient      4-35 4.5.4 Operating Limits                                          4-39 A. Core Peaking Factors                        4-39 B. Power distribution within a Fuel Element. 4-40 C. Power per rod                                4-41 4.6 Core Reactivity                                                              4-45 4.7 Thermal Hydraulic Design                                                    4-47 4.7.1 Heat Transfer Model                                      4-48 4.7.2 Results                                                  4-51 Appendix 4.1, PULSING THERMAL RESPONSE                                          4.1-1 5.0 REACTOR COOLANT SYSTEMS                                                      5-1 5.1 Summary Description                                                          5-1 5.2 Reactor Pool                                                                  5-1 5.2.1 Heat Load                                                  5-2 5.2.2 Pool Fabrication                                          5-3 5.2.3 Beam Ports                                                5-3 5.3 Pool Cooling System                                                          5-4 5.3.1 Reactor Pool                                              5-4 5.3.2 Pool Heat Exchanger                                        5-5 5.3.3 Secondary Cooling                                        5-10 5.3.4 Control System                                            5-10 5.4 Primary Cleanup System                                                      5-11 5.5 Makeup Water System                                                          5-12 5.6 Cooling System Instruments and Controls                                      5-13 iii
SAFETY ANALYSIS REPORT, TABLE OF CONTENTS                                          01/2012 Table of Contents Section                                                                            Page 6.0 ENGINEERED SAFEGUARD FEATURES                                                    6-1 6.1 References                                                                      6-1 7.0 INSTRUMENTATION AND CONTROL SYSTEM                                              7-1 7.1 DESIGN BASES                                                                    7-1 7.1.1. NM-1000 Neutron Channel                                    7-3 7.1.2. NP-1000 Power Safety Channel                                7-5 7.1.3. Reactor Control Console                                    7-6 7.1.4. Reactor Operating Modes                                    7-7 7.1.5. Reactor Scram and Shutdown System                          7-11 7.1.6. Logic Functions                                            7-12 7.1.7 Mechanical Hardware                                        7-13 7.2 DESIGN EVALUATION                                                              7-14 8.0 ELECTRIC POWER SYSTEMS                                                          8-1 9.0 AUXILIARY SYSTEMS                                                                9-1 9.1 Confinement System                                            ......            9-1 9.2 HVAC (Normal Operations)                                                        9-1 9.2.1 Design basis                                                9-2 9.2.2 System description                                          9-3 9.2.3 Operational analysis and safety function                    9-4 9.2.4 Instruments and Controls                                    9-6 9.2.5 Technical Specifications, bases, testing and surveillances  9-8 9.3 Auxiliary Purge System                                                          9-8 9.3.1 Design basis                                                9-8 9.3.2 System description                                          9-8 9.3.3 Operational Analysis and Safety Function                    9-9 9.3.4 Instruments and controls                                    9-9 9.3.5 Technical Specifications, bases, testing and surveillances  9-10 9.4 Fuel storage and handling                                                      9-10 9.4.1 Design basis                                                9-10 9.4.2 System description                                          9-10 9.4.3 Operational analysis and safety function                    9-12 9.4.4 Instruments and controls                                    9-12 9.4.5 Technical Specifications, bases, testing and surveillances  9-12 9.5 Fire protection systems                                                        9-13 9.5.1 Design basis                                                9-13 9.5.2 System description                                          9-13 9.5.3 Operational analysis and safety function                    9-14 iv
THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR                o              I  01/2012 0401/                  2012 SAFETY ANALYSIS REPORT Table of Contents Section                                                                              Page 9.5.4 Instruments and controls                                    9-15 9.5.5 Technical Specifications, bases, testing and surveillances  9-15 9.5 Communications systems                                                          9-15 9.5.1 Design basis                                                9-15 9.5.2 System description                                          9-15 9.5.4 Instruments and controls                                    9-16 9.5.5 Technical Specifications, bases, testing and surveillances  9-16 9.6 Control, storage, use of byproduct material (including labs)                    9-16 9.6.1 Design basis                                                  9-16 9.6.2 System description (drawings, tables)                        9-16 9.6.3 Operational analysis and safety function                    9-17 9.6.4 Instruments and controls                                    9-17 9.6.5 Technical Specifications, bases, testing and surveillances  9-17 9.7 Control and storage of reusable components                                      9-17 9.7.1 Design basis                                                9-17 9.7.2 System description                                          9-17 9.7.3 Operational analysis and safety function                    9-17 9.7.4 Instruments and controls                                    9-17 9.7.5 Technical Specifications, bases, testing and surveillances  9-18 9.8 Compressed gas systems                                                          9-18 9.8.1 Design basis                                                9-18 9.8.2 System description                                          9-18 9.8.3 Operational analysis and safety function                    9-18 9.8.4 Instruments and controls                                    9-19 9.8.5 Technical Specifications, bases, testing and surveillances  9-19 10.0 EXPERIMENTAL FACILTIES AND UTILIZATION                                          10-1 10.1 Summary Description                                                            10-1 10.2 In-Core Facilities                                                              10-3 10.2.1 Central Thimb;e (In-Core Facility)                          10-4 A. DESCRIPTION                                    10-4 B. DESIGN & SPECIFICATIONS                        10-5 C. REACTIVITY                                      10-6 D. RADIOLOGICAL ASSESSMENT                        10-6 E. INSTRUMENTATION                                10-7 F. PHYSICAL RESTRAINTS, SHIELDS, OR BEAM CATCHERS  10-7 G. OPERATING CHARACTERISTICS                      10-7 H. SAFETY ASSESSMENT                              10-8 10.2.2 Fuel Element Positions (In-Core Facilities)                10-8 V
SAFETY ANALYSIS REPORT, TABLE OF CONTENTS                                        01/2012 Table of Contents Section                                                                          Page 10.2.2.1 Pneumatic Sample Transit System                            10-8 A. DESCRIPTION.                                    10-8 B. DESIGN & SPECIFICATIONS.                        10-9 C. REACTIVITY                                      10-10 D. RADIOLOGICAL ASSEMENT                          10-11 E. INSTRUMENTATION                                10-11 F. PHYSICAL RETRAINTS, SHIELDS, OR BEAM CATCHERS  10-12 G. OPERATING CHARACTERISTICS                      10-12 H. SAFETY ASSESSMENT                              10-12 10.2.2.2 Three Element Irradiator                                  10-13 A. DESCRIPTION.                                    10-13 B. DESIGN & SPECIFICATIONS.                        10-13 B (1) Upper and Lower Grid Plate Modifications. 10-13 B (2) Alignment Frame.                            10-14 B (3) Three Element Facility Canister.            10-14 C. REACTIVITY                                      10-16 C (1) Reactivity Calculation                      10-17 C (2) Reactivity Measurements                      10-18 D. RADIOLOGICAL ASSESSMENT                        10-18 E. INSTRUMENTATION                                10-19 F. PHYSICAL RESTRAINTS, SHIELDS, or BEAM CATCHERS  10-19 G. OPERATING CHARACTERISTICS                      10-19 H. SAFETY ASSESSMENT                              10-19 H (1) Cooling                                      10-19 H (2) Temperature                                  10-20 H (3) Pressure                                    10-21 H (4) LOCA potential                              10-22 10.2.2.3 6/7 Element Irradiator                                    10-22 A. DESCRIPTION                                    10-22 B. DESIGN AND SPESIFICATIONS,                      10-22 C. REACTIVITY.                                    10-23 D. RADIOLOGICAL ASSESSMENT                        10-23 E. INSTRUMENTATION                                10-23 F. PHYSICAL RESTRAINTS, SHIELDS OR BEAM CATCHERS  10-24 G. OPERATING CHARACTERISTICS                      10-24 H. SAFETY ASSESSMENT                              10-24 H (1) Temperature (Fuel)                          10-24 H (2) Temperature (Lead)                          10-24 H (3) Pressure (irradiation Can)                  10-24 H (4) Pressure (Lead Sleeve)                      10-25 vi
THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR            0#-0,                01/2012 SAFETY ANALYSIS REPORT Table of Contents Section                                                                            Page H (5) Mass                                        10-25 H (6) Structural                                  10-25 10.2.3 Rotary Specimen Rack                                        10-26 A. DESCPIPTION                                    10-26 B. DESIGN SPEC!,FiC1,1A*ATiN5                    10-26 C. REACTIVITY                                    10-28 D. RADIOLOGICAL ASSESSMENT                        10-28 E. INSTRUMENTATION                                10-29 F. PHYSICAL RESTRAINTS, SHIELDS OR BEAM CATCHERS  10-29 G. OPERATING CHARACTERISTICS                      10-29 H. SAFETY ASSESMENT                              10-29 10.3 Beam Ports                                                                    10-29 A. DESCRIPTION                                                    10-29 B. DESIGN AND SPECIFICATIONS                                      10-30 C. REACTIVITY                                                      10-31 D. RADIOLOGICAL ASSESSMENT                                        10-31 E. INSTRUMENTATION                                                10-31 F. PHYSICAL RESTRAINTS, SHIELDS, OR BEAM CATCHERS                  10-31 G. OPERATING CHARACTERISTICS                                      10-33 H. SAFETY ASSESSMENT                                              10-33 10.4 Cold Neutron Source                                                            10-34 A. DESCRIPTION                                                    10-34 B. DESIGN AND SPECIFICATIONS                                      10-34 C. REACTIVITY                                                      10-37 D. RADIOLOGICAL                                                    10-37 E. INSTRUMENTATION                                                10-37 F. PHYSICAL RESTRAINTS, SHIELDS, OR BEAM CATCHERS                  10-39 G. OPERATING CHARACTERISTICS                                      10-39 H. SAFETY ANALYSIS                                                10-40 10.5 Non-reactor experiment facilities                                            10-41 10.5.1 Neutron generator room                                      10-41 10.5.2 Subcritical assembly                                        10-42 10.5.3 Laboratories                                                10-42 10.5.3.1 Radiochemistry laboratory                10-42 10.5.3.2 Neuron Activation Analysis Laboratory    10-43 10.5.3.3 Radiation detection laboratory          10-43 10.5.3.4 Sample preparation laboratory            10-43 10.5.3.5 General purpose laboratory              10-43 10.6 Experiment Review                                                            10-43 vii
SAFETY ANALYSIS REPORT, TABLE OF CONTENTS                                                01/2012 Table of Contents Section                                                                                  Page 11 Radiation Protection and Waste Management                                              11-1 11.1 Radiation Protection                                                                11-1 11.1.1 Radiation Sources                                                11-1 11.1.1.1 Airborne Radiation Sources                      11-1 11.1.1.1.1 Production of Ar-41 in the Reactor Room      11-1 11.1.1.1.2 Radiological Impact of Ar-41 Outside the      11-2 Operations Boundary 11.1.1.2 Liquid Radioactive Sources                      11-3 11.1.1.2.1 Radioactivity in the Primary Coolant          11-3 11.1.1.2.2 N-16 Radiation Dose Rates from Primary        11-4 Coolant 11.1.1.3 Solid Radioactive Sources                      11-4 11.1.1.3.1 Shielding Logic                              11-6 11.1.2 Radiation Protection Program                                      11-6 11.1.2.1 Management and Administration                  11-7 11.1.2.1.1 Level 1 Personnel                            11-7 11.1.2.1.2 Level 2 Personnel                            11-7 11.1.2.1.3 Level 3 Personnel                            11-9 11.1.2.1.4 Level 4 Personnel                            11-10 11.1.2.1.5 Other Facility Staff                        11-11 11.1.2.2 Health Physic Procedures and Document Control  11-11 11.1.2.3 Radiation Protection Training                  11-11 11.1.2.4 Audits of the Radiation Protection Program    11-13 11.1.2.5 Health Physics Records and Record Keeping      11-13 11.1.3 ALARA Program                                                    11-13 11.1.4 Radiation Monitoring and Surveying                              11-14 11.1.4.1 Monitoring for Radiation Levels and            11-14 Contamination 11.1.4.2 Radiation Monitoring Equipment                11-15 11.1.4.3 Instrument Calibration                        11-15 11.1.5 Radiation Exposure Control and Dosimetry                        11-16 11.1.5.1 Shielding                                      11-16 11.1.5.2 Containment                                    11-16 11.1.5.3 Entry Control                                  11-17 11.1.5.4 Personal Protective Equipment                  11-17 11.1.5.5 Representative Annual Radiation Doses          11-17 11.1.5.6 Personnel Dosimetry Devices                    11-18 11.1.6 Contamination Control                                            11-18 11.1.7 Environmental Monitoring                                        11-19 11.2 Radioactive Waste Management                                                        11-19 11.2.1 Radioactive Waste Management Program                            11-20 viii
THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR                                          01/2012 SAFETY ANALYSIS REPORT                                                Ru01/201t Table of Contents Section                                                                                    Page 11.2.2 Radioactive Waste Controls                                        11-20 11.2.2.1 Gaseous Waste                                  11-20 11.2.2.2 Liquid Waste                                    11-21 112.2.3 Solid Waste                                      11-21 11 2.2.4 Mixed Waste                                    11-21 11.2.2.5 Decommissioning Waste                          11-21 11.2.3 Release of Radioactive Waste                                      11-22 12 Conduct of Operations                                                                    12-1 12.1 Organization                                                                          12-1 12.1.1 Structure                                                          12-1 12.1.1.1 University Administration                        12-1 12.1.1.2 NETL Facility Administration                    12-1 12.1.2 Responsibility                                                      12-3 12.1.2.1 Executive Vice President and Provost            12-3 12.1.2.2 Vice President for University Operation          12-3 12.1.2.3 Associate Vice President of Campus Safety And  12-3 Security 12.1.2.4 Director of Nuc!ear Engineering Teaching        12-3 Laboratory 12.1,2.5 Associate Director of Nuclear Engineering      12-3 Teaching Labor3tory 12.1.2.6 Reactor Oversight Committee                      12-4 12.1.2.7 Radiation Safety Officer                        12-4 12.1.2.8 Radiation Safety Committee                      12-4 12.1.2.9 Reactor Supervisor                              12-4 12.1.2.10 Health Physicist                                12-6 12.1.2.11 Laboratcry Manager.                            12-6 12.1.2.12 Reactor Operators                              12-6 12.1.2.13 Technical Support                              12-6 12.1.2.14 Radiological Controls Technicians              12-6 12.1.2.15 Laboratory Assistants                          12-7 12.1.3 Staffing                                                            12-7 12.1.4 Selection and Training of Personnel                                12-8 12.1.4.1 Qualifications                                  12-8 12.1.4.2 Job Descriptions                                12-8 12.1.4.2.1 Facility Director                              12-8 12.1.4.2.2 Associate Director                            12-9 12.1.4.2.3 Reactor Supervisor                            12-9 12.1.4.2.4 Health Physicist                              12-9 ix
SAFETY ANALYSIS REPORT, TABLE OF CONTENTS                                              01/2012 Table of Contents Section                                                                                Page 12.1.4.2.5 Laboratory Manager                        12-9 12.1.4.2.6 Reactor Operators                        12-9 12.1.4.2.7 Technical Support                        12-9 12.1.4.2.8 Radiological Controls Technician          12-10 12.1.4.2.9 Laboratory Assistants                    12-10 12.1.5 Radiation Safety                                              12-10 12.2 Review and Audit Activities                                                      12-10 12.2.1 Composition and Qualifications                                12-10 12.2.2 Charter and Rules                                            12-11 12.2.3 Review Function                                              12-11 12.2.4 Audit Function                                                12-12 12.3 Procedures                                                                        12-12 12.4 Required Actions                                                                  12-13 12.4.1 Safety Limit Violation                                        12-13 12.4.2 Release of Radioactivity                                      12-14 12.4.3 Other Reportable Occurrences                                  12-14 12.5 Reports                                                                          12-14 12.5.1 Operating Reports                                            12-15 12.5.2 Other or Special Reports                                      12-15 12.6 Records                                                                          12-16 12.6.1 Lifetime Records                                              12-16 12.6.2 Five Year Period                                              12-16 12.6.3 One Training Cycle                                            12-17 12.7 Emergency Planning                                                                12-17 12.8 Security Planning                                                                12-17 12.9 Quality Assurance                                                                12-17 12.10 Operator Requalification                                                        12-18 12.11 Startup Program                                                                  12-19 12.12 Environmental Report                                                            12-19 13.0 ACCIDENT ANALYSIS                                                                  13-1 13.1 Notation and Fuel Properties                                                      13-1 13.2 Accident Initiating Events and Scenarios                                          13-2 13.3 Maximum Hypothetical Accidents, Single Element Failure in Air                      13-5 13.3.1 Assumptions                                                    13-5 13.3.2 Analysis                                                      13-6 A. Radionuclide Inventory Buildup and Decay, Theory  13-7 B. Fission Product Inventory Calculations            13-7 C. Fission Product release                          13-10 D. ALl Consequence Analysis                          13-11 x
THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR                :, -01/2012 SAFETY ANALYSIS REPORTI0/22 Table of Contents Section                                                                            Page E. DAC Consequence Analysis                    13-14 F. Effluent release Ccnsequence Analysis      13-17 F (1) Atmospheric Dispersion                  13-18 F (2) CASE I                                  13-19 F (3) CASE II                                  13-20 F (3) Source Term Release Rate                13-22 13.3.3 Results and Conclusions                                  13-24 13.4 Insertion of Excess Reactivity                                                13-25 13.4.1 Initial Conditions, Assumptions, and Approximations      13-25 13.4.2 Computational Model for Power Excursions                  13-26 13.4.3 Results and Conclusions                                  13-30 13.5 Loss of Reactor Coolant Accident                                              13-32 13.5.1 Initial Conditions, Assumptions, and Approximations      13-34 13.5.2 Heat Transfer to Air                                      13-34 A. Buoyancy Forces                            13-35 B. Friction Losses                            13-35 C. Losses from Flow Restrictions              13-35 13.5.7 Radiation Levels from the Unccvered Core                  13-39 13.5.8 Results and Conclusions                                  13-43 13.6 Loss of Coolant Flow                                                          13-43 13.6.1 Initialing Events and Scenarios                          13-43 13.6.2 Accident Analysis and Determination of Consequences      13-43 13.7 Mishandling or Malfunction of Fuel                                            13-44 13.7.1 Initiating Events and Scenarios                          13-44 13.7.2 Analysis                                                  13-44 13.8 Experiment Malfunction                                                        13-44 13.8.1 Accident Initiating Events and Scenarios                  13-44 13.8.2. Analysis and Determination of Consequences              13-45 A. Administrative Controls                    13-45 B. Reactivity Considerations                  13-45 C. Fueled Experiment Fission Product Inventory 13-46 D. Explosives                                  13-47 13.9 Loss of Normal Electric Power                                                13-49 13.9.1 Initiating Events and Scenarios                          13-49 13.9.2 Accident Analysis and Determination of Consequences      13-49 13.10 External Events                                                              13-49 13.10.1 Accident Initiating Events and Scenarios                13-49 13.10.2 Accident Analysis and Determination of Consequences      13-50 13.11 Experiment Mishandling or Malfunction                                        13-50 13.11.1 Initiating Events and Scenarios                          13-50 xi
SAFETY ANALYSIS REPORT, TABLE OF CONTENTS                                    01/2012 Table of Contents Section                                                                      Page 13.11.2 Accident Analysis and Determination of Consequences  13-50 Appendix 13.1, T-6 DEPLETION ANALYSIS INPUT FILE FOR SCALE CALCULATION      13.1-1 Appendix 13.2, ORIGEN ARP INPUT                                              13.2-1 Appendix 13.3, MCNP INPUT FOR LOCA DOSES                                    13.3-1 15.0 FINANCIAL QUALIFICATIONS                                                15-1 15.1 Financial Ability to Operate a Nuclear Research Reactor                  15-1 15.2 Financial Ability to Decommission the Facility                          15-1 15.3 Bibliography                                                            15-1 Appendix 15.1, STATUTES AND EXCERPTS REGARDING UT                            15.1-1 Appendix 15.2, FIVE-YEAR OPERATING COST ESTIMATE                            15.2-1 Appendix 15.3, Letter of Intent, Ultimate Decommissioning                    15.3-1 Appendix 15.4, DECOMMISSIONING COST ESTIMATE                                15.4-1 APPENDIX 15.5, FUELS ASSISTANCE CONTRACT                                    15.5-1 xii
THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR              0*"                01/2012 SAFETY ANALYSIS REPORT                                      ?** REM LIST OF FIGURES                                                                  Page Figure 1.1, UT TRIGA Mark II Nuclear Research Reactor                              1-4 Figure 1.2, Core and Support Structure Details                                      1-5 Figure 1.3, Beam Ports                                                              1-8 Figure 1.4A, Days of Operation per Year                                            1-12 Figure 1.4B, Burnup per Year                                                      1-12 Figure 2.1, STATE OF TEXAS COUNTIES                                                2-2 Figure 2.2, TRAVIS COUNTY                                                          2-3 Figure 2.3, CITY OF AUSTIN                                                        2-4 Figure 2.4, JJ PICKLE RESEARCH CAMPUS                                              2-5 Figure 2.5, LAND USAGE AROUND JJ PICKLE RESEARCH CAMPUS, 2007                      2-6 Figure 2.6, 2009 ZIP CODE BOUNDARIES                                              2-10 Figure 2.7, AUSTIN CLIMATOLOGY DATA                                                2-11 Figure 2.8, AUSTIN WIND ROSE DATA                                                  2-12 Figure 2.9, TROPICAL STORM PATHS WITHIN 50 NAUTICAL MILES OF AUSTIN, TEXAS (ALL    2-21 RECORDED HURRICANES RATED H1 AND UP)
Figure 2.10, TROPICAL STORM PATHS WITHIN 50 NAUTICAL MILES OF AUSTIN, TEXAS (ALL  2-21 RECORDED STORMS RATED TROP OR SUBTROP)
Figure 2.11, BALCONES FAULT ZONE                                                  2-23 Figure 2.12, TEXAS EARTHQUAKE DATA                                                2-24 Figure 2.13, TEXAS EARTHQUAKE DATA                                                2-25 Figure 2.14, LOCAL WATER AQUIFERS                                                  2-26 Figure 2.15, RESEARCH CAMPUS AREA 1940                                            2-27 Figure 2.16, PICLKE RESEARCH CAMPUS 1960                                          2-28 Figure 2.17, BALCONES RESEARCH CENTER 1990                                        2-29 Figure 4.1: H/Zr Phase Diagram                                                      4-6 Figure 4.2A, Zr-H Transport Cross Section & TRIGA Thermal Neutron Spectra          4-7 Figure 4.2B, Fuel Temperature Coefficient of Reactivity                            4-7 Figure 4.3, Thermal Pressurization in Fuel and Hydriding Ratios                    4-9 Figure 4.4A, Temperature and Cladding Strength for 0.2% Yield                      4-11 Figure 4.4B, Temperature, Cladding Strength, and Stress                            4-12 Figure 4.5, Lower Gird Plate Control Rod Positions                                4-14 Figure 4.6, Standard Control Rod Configuration                                    4-15 Figure 4.7, Standard/Stepper Motor Control Rod Drive                              4-16 Figure 4.8, Transient Rod Drive                                                    4-18 Figure 4.9a, UT TRIGA Core                                                        4-21 Figure 4.9b, Core Top View                                                        4-21 Figure 4.10a, 6/7-Element Facility Grid                                            4-22 Figure 4.10b, Upper Grid Plate Cut-out for 6/7-Element Grid                        4-22 Figure 4.1la, Reflector Top Assembly                                              4-23 Figure 4.11b, Reflector Bottom Assembly                                            4-23 xiii
SAFETY ANALYSIS REPORT, TABLE OF CONTENTS                                                01/2012 LIST OF FIGURES                                                                          Page Figure 4.12b, Graphite Reflector Through port Detail                                    4-23 Figure 4.12c, Graphite Reflector, Radial & Piercing-Beam Ports                          4-23 Figure 4.13a, Tangential Beam Port Insert                                                4-23 Figure 4.13b, Radial Beam Port insert                                                    4-23 Figure 4.13c, Inner Shroud Surface                                                      4-24 Figure 4.14, Reflector Component and Assembly Views                                      4-25 Figure 4.15, Fuel Element Adapter                                                        4-26 Figure 4.16, Core Support Views                                                          4-27 Figure 4.17, Core and Support Structure Views                                            4-27 Figure 4.18, Safety Plate                                                                4-28 Figure 4.19a, Pool                                                                      4-29 Figure 4.19b, Side View                                                                  4-29 Figure 4.19c, Top View                                                                  4-29 Figure 4.20, Biological Shielding, Base Dimensions                                      4-31 Figure 4.21, Reactivity Loss with Power                                                  4-34 Figure 4.22, Radial Variation of Power Within a TRIGA Fuel Rod. (Data Points from Monte  4-41 Carlo Calculations [Ahrens 1999a])
Figure 4.23, Critical Heat Flux Ratio (Bernath and Biasi Correlations)                  4-44 Figure 4.24, Core Power, 45 kW Hot Element                                              4-45 Figure 4.25, Power Coefficient of Reactivity                                            4-46 Figure 4.25: Unit Cell Fuel Element Model                                                4-50 Figure 4.26a, Unit Cell Temperature Distribution (10.5 kW)                              4-55 Figure 4.26b, Unit Cell Temperature Distribution (22.5 kW)                              4-56 Figure 4.27, Single Rod Flow Cooling Flow Rate versus Power Level 49°C 6.5 Pool,        4-56 Figure 4.28, Comparison of Calculated and Observed Fuel Temperatures                    4-58 Figure 5.1A, Pool Fabrication                                                            5-4 Figure 5.1B, Cross Section                                                                5-4 Figure  5.C, Beam Orientation                                                            5-4 Figure  5.2, Pool Cooling System                                                          5-4 Figure  5.3, Pool Cleanup System                                                        5-11 Figure  5.4, Cooling and Cleanup Instrumentation                                        5-13 FIGURE 7.1, CONTROL SYSTEM BLOCK DIAGRAM                                                  7-3 Figure 7.2, NEUTRON CHANNEL OPERATING RANGES                                              7-4 Figure 7.3, Auxiliary Display Panel                                                      7-5 Figure 7.3, LAYOUT OF THE REACTOR CONTROL CONSOLE                                        7-6 Figure  7.4, CONSOLE CONTROL PANELS                                                      7-8 Figure  7.5, TYPICAL VDEO DISPLAY DATA                                                    7-9 Figure  7.6, ROD CONTROL PANEL                                                            7-9 Figure  7.7, LOGIC DIAGRAM FOR CONTROL SYSTEM                                            7-13 Figure 9.1, Conceptual Diagram of the Reactor Bay HVAC System                            9-2 Figure 9.2A, Main Reactor Bay HVAC System                                                9-3 Figure 9.2B, Main Reactor Bay HVAC Control System Control                                9-4 xiv
THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR          *a**
00F            01/2012 SAFETY ANALYSIS REPORT LIST OF FIGURES                                                              Page Figure 9.3, Confinement System Ventilation Contrcos                          9-7 Figure 9.4A, Purge Air System                                                9-8 Figure 9.4B, Purge Air Controls                                              9-8 Figure 9.5A, Storage Well                                                    9-11 Figure 9.5b, Fuel Storage Closure                                            9-11 Figure 10.1, Core Grid Plate Design and Dimensions                            10-3 Figure 10.2, Reactor Core Diagram                                            10-4 Figure 10.3, Central Thimble Union Assembly                                  10-5 Figure 10.4, Three Element Irradiator                                        10-16 Figure 10.5, Rotary Specimen Rack Diagram                                    10-28 Figure 10.6, Rotary Specimen Rack Raceway Geometry                          10-28 Figure 10.7, Rotary Specimen Rack Rotation Control Box                      10-28 Figure 10.8, Beam Port Layout                                                10-30 Figure 10.9, A1230 Cryomech Cryorefrigerator and Cold Head                  10-35 Figure 10.10, Cryomech Cold-Head and Vacuum Box                              10-36 Figure 10.11, TCNS Vacuum Jacket and Other Instruments (units in cm)        10-36 Figure 10.12, Silicone Diode and Heater Relative to Cold-Head                10-37 Figure 10.13, Neon and Mesitylene Handling System with Pressure Transducers  10-38 Figure 10.14, Shielding around TCNS Facility                                10-40 Figure 10.15, Thermo MP 320 Neutron Generator at NETL                        10-41 Figure 10.16, Subcritical Assemblies                                        10-42 Figure 12.1, University Administration                                        12-2 Figure 12.2, NETL Facility Administration                                    12-2 Figure 13.1, Ratio of Radionuclide Inventory to ALl                          13-13 Figure 13.2, Ratio of Radionuclide Concentration to 10CFR 20 DAC Values      13-14 Figure 13.3, FUEL Temperature and Pulsed Reactivity                          13-35 Figure 13.4A, Pulse Measurements                                            13-31 Figure 13.4B, Fuel Temperature and Peak Pulse Power                          13-31 Figure 13.5A, Cooling Time                                                  13-37 Figure 13.5B, Cooling Time and Power Density                                13-38 Figure 13.6, Core Model                                                      13-41 Figure 13.7A, Bay Model Top View                                            13-41 Figure 13.7B, Bay Model Cross Section                                        13-41 Figure 13.8A, Building Model                                                13-42 Figure 13.8B, MCNP Side View                                                13-42 Figure 13.8C, Top View                                                      13-42 xv
SAFETY ANALYSIS REPORT, TABLE OF CONTENTS                                        01/2012 Table                                                                            Page Table 1.1, SHUTDOWN OR DECOMMISSIONED U.S. TRIGA REACTORS                        1-10 Table 1.2, U.S. OPERATING RESEARCH REACTORS USING TRIGA FUEL                      1-10 Table 2.1, AUSTIN AND TRAVIS COUNTY POPULATION TRENDS                            2-8 Table 2.2, TRAVIS COUNTY 2009 AUSTIN POPULATION DENSITY DISTRIBUTION BY ZIP CODE  2-9 Table 2.3, 1982 METEOROLOGICAL DATA FOR AUSTIN TEXAS                            2-15 Table 2.4, HISTORICAL METEOROLOGICAL DATA FOR AUSTIN TEXAS                      2-16 Table 2.5, HISTORICAL METEOROLOGICAL DATA FOR AUSTIN TEXAS                      2-17 Table 2.6, HISTORICAL METEOROLOGICAL DATA FOR AUSTIN TEXAS                      2-18 Table 2.7, HISTORICAL METEOROLOGICAL DATA FOR AUSTIN TEXAS                      2-19 Table 2.8, TRAVIS COUNTY TORNADO FREQUENCIES                                    2-20 Table 2.9 GROUND WATER ACTIVITY                                                  2-26 Table 3.1, SSC Vulnerability                                                      3-2 Table 4.1, TRIGA Fuel Properties                                                  4-3 Table 4.2, Physical Properties of High-Hydrogen U-ZrH                            4-4 Table 4.3, U-ZrH Volumetric Specific Heat Capacity (Cp)                          4-6 Table 4.4, Summary of Control Rod Design Parameters                              4-13 Table 4.5, Control Rod Information                                              4-15 Table 4.6, Summary of Reactor SCRAMs                                            4-19 Table 4.7, Summary of Control Rod Interlocks                                    4-19 Table 4.8, Upper Grid Plate Penetrations                                        4-21 Table 4.9, Displaced Fuel Spaces                                                4-22 Table 4.10, Lower Grid Plate Penetrations                                        4-25 Table 4.11, Reactor Coolant System Design Summary                                4-28 Table 4.12, Significant Shielding and Pool Levels                                4-32 Table 4.13, Control Rod Worth                                                    4-33 Table 4.14, Reactivity Values                                                    4-33 Table 4.15, GA-4361 Calculation Model                                            4-35 Table 4.16, Selected TRIGA II Nuclear Properties                                4-35 Table 4.17, UT TRIGA Data                                                        4-36 Table 4.18, Critical Heat Flux ratio, Bernath Correlation                        4-43 Table 4.19, Core Power, 45 kW Hot Element                                        4-44 Table 4.20, Reactivity Limits                                                    4-46 Table 4.21, Limiting Core reactivity                                            4-47 Table 4.22, Thermodynamic Values                                                4-49 Table 4.23, Hydrostatic Pressure                                                4-51 Table 4.24, Coolant Temperature for 49°C 6.5 m Pool                              4-51 Table 4.25a, Outer Cladding Temperature (°C) for 49°C and 6.5 m Pool            4-52 Table 4.25b, Inner Cladding Temperature (°C) for 49"C and 6.5 m Pool            4-53 Table 4.26a, Heat Flux (Nodes 1-9) 49°C 6.5 Pool,                                4-53 Table 4.26b, Heat Flux (Nodes 10-15) 49°C 6.5 Pool                              4-54 Table 4.27, Peak Fuel Centerline Line Temperature (K)49°C 6.5 Pool,              4-54 Table 4-28, Coolant Flow for 1100 kW Operation                                  4-57 Table 4-29, Observed Fuel Temperatures                                          4-57 Table 4-30, Fuel Temperature Comparison                                          4-58 Table 5.1, Reactor Coolant System design Summary                                  5-2 Table 5.2, Heat Exchanger, Heat Transfer and Hydraulic Parameters                5-9 xvi
THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR                0            j          01/2012 Table 9.1, Typical Confinement Vent & Purge Parameters                                    9-4 Table 9.2, Reactor Ventilation System Modes                                              9-5 Table 10.1: Composition of Al 6061                                                        1.0-6 Table 10.2: Activation Products in Central Thimble 6061 Aluminum Alloy after 60 Year      10-7 Irradiation Table 10.3 Characteristic Dimension of UT-TRIGA PTS                                      10-10 Table 10.4: Activation of Pneumatic Transit System Cadmium Liner                        10-11 Table 10.5: Flux Measurements in Pneumatic Transit System zt 100 kW                      10-12 Table 10.6: Activity of Three Element Irradiator Cd Liner                                10-19 Table 10.7: Rotary Specimen Rack Gears                                                  10-27 Table 10.8: Items to be Addressed in Safety Analysis for Experiments                    10-44 Table 11.1, Representative Solid Radioactive Sources                                      11-5 Table 11.2, Representative Radiation Detection Instrumentation                          11-15 Table 11.3, Representative Occupational Exposures                                        11-17 Table 13.1. Neutronic Properties of TRIGA Mkll ZrH1.6 Fue! Elements.                      13-1 Table 13.2, Dimensions of TRIGA Mkll ZrH1.6 Fuel Elementsl                                13-1 Table 13.3, Thermal and Mechanical Properties of TRIGA Mkll ZrH1.6 Fuel Elements and      13-2 Type 304 Stainless Steel Cladding Table 13.4, UT TRIGA Core-Conditions Basis for Calculations                              13-2 Table 13.5, Relevant IOCFR20 Appendix B Values                                            13-5 Table 13.6, SCALE T-6 Sequence Continuous Burnup Parameters                              13-8 Table 13.7A, 1 MTU Gaseous Fission Product Inventory for 3.5 kW Case (Ci)                13-8 Table 13.7B, 1 MTU Particulate Fission Product Inventory (Ci)                            13-9 Table 13.8A. Gaseous Fission product Release from Single Element (lVCi)                  13-10 Table 13.8B. Particulate Fission Product Release from Single Element                    13-11 Table 13.9A, Fraction of Gaseous Fission Product Inventory to 10CFR20 ALl                13-12 Table 13.9B, Fraction of Particulate Fission Product Inventory to IOCFR20 ALl            13-12 Table 13.10A, Fraction of Instantaneous Gaseous Fission Product Inventory to 10CFR20    13-14 DAC[1]
Table 13.10B, Fraction of Instantaneous Particulate Fission Product Inventory to 10CFR20 13-15 DAC [1]
Table 13.11, DAC Ratios for All Cases                                                    13-16 Table 13.12, Reactor Bay Atmosphere Following MHA Compared to Effluent Limit            13-17 Table 13.13: BRIGGS URBAN DISPERSION PARAMETERS                                          13-18 Table 13.14, Calculated ?/Q Values                                                      13-21 Table 13.15, Reactor Bay Atmosphere Following MHA Compared to Effluent Limit            13-21 Table 13.16, Calculated Plume Meander Factor (M) for < 6 m s-1 Winds                    13-21 Table 13.17, Minimum Dispersion Parameters by Stability Class                            13-22 Table 13.18, Minimum ?/Q by Stability Class                                              13-22 Table 13.19, Effluent Limit Ratio to Release Concentrations                              13-23 Table 13.20, Low Power Pulsed Reactivity Response                                        13-28 Table 13.21, Initial Power 880 kW Pulsed Reactivity Response                            13-30 Table 13.22, Gamma Source Term                                                          13-39 Table 13.23, Height/Thickness Dimensions of Unit Cell                                    13-40 Table 13.24, Unit Cell Areas                                                            13-40 Table 13.25, Material Characterization                                                  13-40 Table 13.26, Post LOCA Doses                                                            13-42 xvii
SAFETY ANALYSIS REPORT, TABLE OF CONTENTS                          01/2012 Table 13.27, Calculations Supporting Limits on Fueled Experiments 13-46 Table 13.28, Material Strengths                                  13-48 Table 13.29, Container Diameter to Thickness Ratio                13-49 xviii
THE UNIVERSITY OF TEXAS TRIGA 11RESEARCH REACTOR                                        01/2012 SAFETY ANALYSIS REPORT, CHAPTER 4 i    ET0/
4.0      Reactor This chapter will discuss the reactor core (fuel, control rods, reflector and core support, neutron source, core structure), reactor pool, biological shielding, nuclear design (normal operating conditions, and operating limits), and thermal hydraulic design.
4.1      Summary description The University of Texas Nuclear Engineering Teaching Laboratory (NETL) is home to a General Atomics' TRIGA Mark II research reactor. This installation follows 25 years (1963-1988) of successful operation of a TRIGA reactor at Taylor Hall on the main campus.
The basic TRIGA design uses U-ZrH 1.6 fuel clad with stainless steel in natural water convection cooling mode during operation, with a maximum decay heat that can be removed by natural convection of either water or air. The reactor is located in an open pool of purified, light water that serves as a heat sink during operations at power. Nuclear properties and characteristics control heat generation; thermodynamic characteristics of the fuel and the coolant control heat removal and temperature response. Maximum fuel temperature is the principle design constraint. 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 evolve into space around the fuel matrix; the hydrogen at elevated temperature can generate pressure inside the cladding. Temperature that produces stress greater than the yield strength for the stainless 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 mass exceeding critical loading (i.e., excess reactivity) is required to compensate for the negative fuel temperature coefficient, as well as potential experiments, fission product poisons, and fuel burnup. There are several major experiment facilities that could affect core reactivity, as described in Chapter 10. Experiment program requirements vary widely; limits are imposed on the reactivity effects of experiments.          The amount of excess reactivity determines the maximum possible power, and therefore the maximum possible fuel temperature.
4.2      Reactor Core The University of Texas at Austin TRIGA II reactor core is configured in a hexagonal prism volume bounded by aluminum plates at the upper and lower surfaces (grid plates), and surrounded by a cylinder of graphite (aluminum clad) acting as a neutron reflector. Sections of the reflector are cut away to support experimental facilities, including beam ports and a rotating specimen rack. The core assembly is supported by structural aluminum, and includes an aluminum plate that serves to limit downward travel of control elements.
Page 4-1
CHAPTER 4: REACTOR                                                                        01/2012 4.2.1    Reactor Fuel The TRIGA fuel system was'developed around the concept of inherent safety, with fuel and cladding designed to withstand all credible envikonmental and radiation conditions during its lifetime at the reactor site. A TRIGA fuel element consists of (A) a central fueled region containing fuel matrix, bounded by an axial reflector and (B) stainless steel end caps at the top and 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 cladding temperature to prevent cladding rupture. The fuel lattice structure that comprises the NETL TRIGA 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 is adequate to limit maximum steady state operating temperature. ; The TRIGA fuel matrix exhibits a large prompt negative temperature coefficient of&#xfd;,reactivity. The maximum fuel temperature resulting from sudden insertion of all available excess reactivity would cause power excursion to terminate before any core damage is possible. Limits on core lattice excess reactivity and individual fuel element temperature therefore are interrelated. The maximum possible TRIGA fuel fission product inventory is limited by fissionable material loading. The maximum TRIGA fuel decay heat produced by fission product inventory can be removed by natural convection in air or water.
Handling, transport, and storage of TRIGA fuel elements at the NETL, fresh and irradiated, are described in Chapter 9, Auxiliary Systems.
A.      Fuel matrix A TRIGA fuel element consists of a central fueled region containing fuel matrix, bounded by an axial reflector (with a molybdenum disk as a protective interface between the fuel and the lower graphite/axial reflector,- and stainless steel end caps-at the top and bottom with a stainless steel cladding.
The basic safety limit for the TRIGA reactor system is the fuel temperature; this applies for both the steady-state and pulse mode of operation. Twe, limiting temperatures are of interest, depending on the type of, TRIGA fuel used. The TRIGA fuel which is considered low hydride, that with an H/Zr ratio of less than ..5, has a lower temperature !imit than fuel with a higher H/Zr ratio. Fig. 4.1. indicates that the higher hydride compositions are single phase and
.are not subject to the large volume changes associated with the phase transformations at approximately 530'C in the lower hydrides. Also, it has been noted' that the higher hydrides lack any significant thermal diffusion of hydrogen. These two facts preclude concomitant volume changes. The important properties of delta phasetU-ZrH are given in Table 4.1.
'GA-3618, Thermal Migration of Hydrogen in Uranium-Zirconium Alloys, Marten U. et. Al., General Dynamics, General Atomics Division (1962)
Page 4-2
THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 4 R~eETI I      01/2012 Graphite dummy elements may be used to fill grid positions in the core. The dummy elements are of the same general dimensions and construction as the fuel-moderator elements. They are clad in aluminum and have a graphite length of Table 4.1, TRIGA Fuel Properties Property                                                Mark III Dimensions Outside diameter, Do = 2ro Inside diameter, Dj= 2ri Overall length Length of fuel zone, L Length of graphite axial reflectors End fixtures and cladding Cladding thickness Burnable poisons Uranium content Weight    percent U 23 5 U enrichment percent 235U  content Physicalpropertiesof fuel excluding cladding H/Zr atomic ratio Thermal conductivity (W cm-' K-1 )
Heat capacity [T >0 0 C] (J cm 3 K')
Mechanicalpropertiesof delta phase U-ZrI-I0 Elastic modulus at 20'C Elastic modulus at 650'C Ultimate tensile strength (to 650'C)
Compressive strength (20'C)
Compressive yield (20'C)
(1) Fabrication A uranium loaded zirconium hydride was found to produce desired moderating characteristics and acceptably low parasitic neutron absorption with strong temperature feedback and high heat capacity. Feedstock of between (or recycled material) are cast in controlled atmosphere, high-temperature induction furnace. 2 Fuel element castings are machined to cylinders of approximately 5 inches in length. A center hole is drilled the length of the cylinder. Additional machining is required for fuel meat to be 2 TRIGA International: A New TRIGA Fuel Fabrication Facility at CERCA - Gerard Ilarbormier, Jean-Claude Ottone, CFIRCA, Proceedings of the 1997 TRTR Annual meeting Page 4-3
CHAPTER 4: REACTOR                                                                                01/2012 fabricated into instrumented fuel assemblies (IFEs, described below) and fuel element followers. The cylinders are heated in a high temperature electric furnace with a hydrogen atmosphere. The exterior and center surface: exposed to hydrogen induces the cylindrical fuel meat to hydride, with a target Zr:H ratio of 1'.6; :.A pure zirconium filler rod is placed in the center hole to maintain nearly uniform thermo-hydraulic properties. Each TRIGA fuel element contains 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 seal in:the upper end fixture, and a leadout tube provides a watertight conduit carrying the leadout wires 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 the transient rod (air filled follower) and the standard rods (fuel fo!lowers) respectively.
(2) Physical Properties The zirconium-hydrogen system is essentially a simple eutectoid, with at least four separate hydride phases. The delta and epsilon phases are respectively face-centered cubic and face-centered tetragonal hydride phases. The two phase delta + epsilon region exists between ZrHI.64 and ZrH1 .7 4 at room temperature, and closes at ZrH 1.7 at 455&deg;C. From 455&deg;C to about 10500 C, the delta phase is supported by a broadening range of H/Zr ratios. Other important properties observed for the delta phase U-ZrH are listed in Table 4.2.
The ratio of Zr-H plays a significant role ir determining physicai properties. The H:ZR material has a cubic structure in the delta-phase at ratios greater than 1.4. in lower H:Zr ratios (< 1.5) a phase change occurs at about 955&deg;F (535&deg;C) with large dens:ity differences between the phases leading to potential for deformation (swelling, and cracking). For hydrogen to zirconium atom ratios greater. than 1.5, the matrix is single phase (delta or epsilon) and does not exhibit phase separation with ,thermal cycling; Thermal diffusion of hydrogen is rrtinimal in higher ratios as well, minimizing potential for dceformaticrn from evolutiion of hydrogen gas. Any hydrogen gas is in equilibrium with: the matrix, substantially retained by the cladding, Losses through the cladding from hydrogen migration are about 1%for cladding temperature about 93 0 0F (500 0C).
Table 4.2, Physical Properties of High-Hydrogen U-ZrH Property                      Temperature              Value      Units Thermal Conductivity,                    93&deg;C - 650&deg;C              0.22    W cm 6      psi 20&deg;C 0                9.1x10 6      psi Elastic Modulus                            650 C                6.0x10 Ultimate Tensile Strength                  20&deg;C                2.4 x10 4      psi Compressive Strength                        20&deg;C                6.0 x10 4      psi Compressive Yield                            20'C                3.5 x10 4      psi Heat of Formation                          298&deg;C                37.75    kcal g-molr Page 4-4
00o&deg;NlTi            01/2012 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 4                                  000o    nE  t At ratios greater than 1.6 there can be a shift to higher density tetragonal. Higher hydride compositions are single phase. and are not subject to the large volume changes associated with the phase transformations at approximately 530TC as in :the: lower hydrides. The stability extends from the minimum on the scale (OTC).to the maximum onwthe:scale (950TC), indicating no volume changes from morphology which might stress cladding occur for a target ratio of 1.6 other than thermal expansion. Significantly, zirconium hydrides at these ratios lack any significant thermal diffusion of hydrogen under isothermal conditions' Undernon-isothermal conditions, hydrogen migrates from high temperature regions to low temperature regions, with equilibrium disassociation pressures lower after redistribution. Hydrogen' dissociates slightly from the fuel matrix at high temperatures, and is re-absorbed :into the matrix at lower temperatures, with the equilibrium hydrogen dissociation pressure a function of both the composition and temperature. The equilibrium hydrogen dissociation pressure is governed by the composition and temperature. For ZrH 1 .6, the equilibrium hydrogen pressure is one atmosphere at about 760 0 C. Hydrogen dissociation pressures of hydrides are similar in alloys up to about 75 weight per cent uranium. For the delta and epsilon phases, dimensional changes from hydrogen migration are not significant.            In the delta .phase, equilibrium disassociation pressures are related by:
K log pK      +
With:
P pressure (atm)
T= temperature (K)
K1= -3.8415 + 38.6433-X - 34.2639.X + 9.282122X3 K2= -31.2981 + 23.5741,X -. 6.0280.X2 X= hydrogen to zirconium atom ratio At a ratio of 1.7 the equilibrium disassociation pressure corresponds to a temperature of about 1400F (300&deg;C). The density of ZrH. decreases as hydrogen ratio increases; from low ratios to the delta phase (H:Zr of 1.5) the density change is high' with little cha&ngefor further increases.
Massively. hydrided bulk density' is-reported to be about 2% Ylower than x~ray diffraction analysis. For TRIGA-fuel with aZr:H ratio of 1:1.6, the uranium density,.volumefractioh, and weight fraction are related by:
WU p(A)=
0.177-0.125. WU and WU=    0.177-pu (A) 1 + 0. 125.-pu (A)
Page 4-5
CHAPTER 4: REACTOR CHPE,:    ECO I  01/2012 0121 po, (A) = 19.07. V)J(A) where pu(A)= Uranium density WU - Uraniuri Weight fraction V1  = volume'fraaction of uranium in the U-ZrH1 .6 alloy Thermal conductivity ha*Ibeen determined from short-pulse heating techniques. Using thermal diffusivity values, density, and specific heat the thermal conductivity of uranium zirconium with a Zr:H ratio of 1:1.6 is 0.042 +/-+0.002 ca[-1 s-5 cm 'C -.
Volumetric specific heat is a function of temperature and composition. Table 4.3 lists values for variations in Zr:H and w% U based on a O&deg;C reference, showing variation less than 10%.
Table 4.3, U-ZrH Volumetric Specific Heat Capacity (Cp)
ZrH                  W% U              Value                    Units U-ZrH 1.6            8.5              2.04 + 4.17x10          W s .cmr3 3
U-ZrH. 7          20                2.17 + 4.36x10          W s -cr 850 w  600 P
450 I
I I                h 0F      02 I      I 04
                                        @4.8 I
06 I
08 I
to I ---
12    0      16 L.
I LB m
I1 Zw HYDROGEN CONTENT (Di Zra Figure 4.1: H/Zr Phase Diagram Page 4-6:
01/2012 THE UNIVERSITY OF TEXAS TRIGA 11RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 4 (3) Operational Properties The neutronic properties of ZrH are the primary motivation for incorporation in TRIGA fuel development. The morphology of ZrH, in particular hydrogen diffusion in the material, imposes limits during operation. Ultimately, personnel exposure related to TRIGA fuel is limited during normal operations and abnormal events by retaining fission products in the fuel elements. It is well known that zirconium can undergo a reaction with water that releases hydrogen, with subsequent potential for a mixture that can be detonated. Such a reaction has the potential-,to release a large fraction of fission product inventory of affected fuel elements, but is not likely given characteristics of operation and properties of the fuel matrix. Fuel element changes occur during operation from thermal stress, which can affect fuel performance. Fuel cladding prevents migration of fission products for the fuel element, but in the absence of cladding it is not likely that all fission products will escape the fuel meat. Finally, thermal effects related to fuel matrix from steady state and pulsing operations are considered.
(4) Neutronic Properties A large fraction of neutron moderation occurs through interactions with hydrogen in the fuel matrix. The zirconium hydride structure has a profound effect on neutron scattering at low energies because of zirconium-hydrogen binding, with distinct lattice energy levels of 0.13 eV and about 0.25 eV found in scattering experiments. Thermal neutrons that interact wi~th.hydrogen in the lattice (where neutron energy is below the lattice energies) therefore have potential to gain energy. Because the fission cross section has 1/v dependence in the thermal range, increasing thermal neutron energy decreases fission probability. If fuel temperature increases, thermal excitation creates more of these relatively high-efnergy lattice centers as indicated in Fig. 4.2a. When the rate of fission is high enough to create elevated fuel temperatures, the elevated fuel temperatures decrease the rate of fission. This phenomenon is responsible for an extremely, high feedback of negative reactivity from fuel temperature illustrated in Fig. 4.2b. Maximum possible fuel temperature and maximum theoretical power level are therefore a function of the amount of fuel in the reactor.
I00 USTAINLESS                    STEEL CLAD
                                                                          -12              8.5 WT-% U-ZrHj.6O CORE 00        "                                400-C
                              -a                                          a 260 401 ba 20' 0 L 0.01                  0.... ........... . l.0 NEUTRON ENERGY (eVI 0  200  400    600  800    1000  /POO
                                                                                      'TEMPERATURE tIC)
Figure 4.2A, Zr-H Transport Cross Section &                  TRIGA  Figure 4.2B, Fuel Temperature Coefficient of Thermal Neutron Spectra                                                  Reactivity Page 4-7
CHAPTER 4: REACTOR                                                                          1      01/2012 (5) Fuel Morphology & Outgassing As noted previously, during fuel fabrication the ratio of hydrogen to zirconium is enhanced by thermally induced diffusion in an atmosphere of pressurized hydrogen. During reactor operation, temperature gradients influence hydrogen diffusivity to promote outgassing, bounded by temperature induced pressurization of the hydrogen in free volume of the cladding. Pressure inside the fuel element does not intrinsically pose a challenge to fuel element integrity, and will be considered as part of cladding performance in a later section. At a given temperature, higher H:Zr ratios (in the absence of phase 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 of free hydrogen gas at higher temperatures also produces more partial gas pressure in the free volume of the element. Calculations performed with a higher mass fraction of uranium result in an increase in the partial pressure of hydrogen by as much as a factor of four. 3 The fuel rod diameter is on the order of the path length of neutron from generation to absorption, 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 outer surface of the fuel. Fuel rod temperature gradient during normal, steady-state operations is monotonically decreasing from a peak at the center of the fuel rod. Routine power changes occur at a rate that allows quasi-steady state thermal equilibrium, but pulsing operations do not. As a consequence, power distribution and development of temperature gradients in steady-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 than during steady state. Diffusion rates are finite, and the diffusion coefficient for thermal diffusion of hydrogen in zirconium 4 (ranging from 4x10 5 to 2x10 8 cm 2 s-1, and requiring days to equilibrate) lags the time cons-cant for the temperature changes. The temperature gradient during the transient peaks near the surface of the fuel rod rather than the center, and rapidly vanishes as the system comes to equilibrium. Therefore thermal gradients in pulsing bias hydrogen diffusion towards the center of the fuel rod with only a small region near the surface having a gradient that promctes outgassing. Surface cooling from endothermic gas emission lowers the surface temperature and therefore tends to iower the diffusion constant at the fuel rod surfaces.        Re-absorption occurs where hydride surfaces are at relatively lower temperatLres. There is evidence that low permeability oxide films on fuel surfaces retard mass transfer. Local heat transfer effects cause the surface temperature to be lower than that which would occur during adiabatic conditions.
3  Journal of NUCLEAR SCIENCE and TECHNOLOGY, Vol. 37, No. 10, p. 887-892 (October 2000); Estimation of Hydrogen Redistributionin Zirconium Hydride under Temperature Gradient 4  Congreso Internacional de Metalugia y Materiales, Primeras Jornadas Internacionales de Materiales Nucleares (19 al 23 de Octubre de 2009, Buenos Airesm Argentina; Some Peculiarities of Hydrogen Behavior and Delayed Hydride Cracking in Zirconium Based Reactor Alloys, Shmakov, R.N. Singh Page 4-8
THE UNIVERSITY OF TEXAS TRIGA ImRESEARCH REACTOR                                    000            01/2012 SAFETY ANALYSIS REPORT, CHAPTER 4                                                        0O
                                    %03 1027 5
Z
                                            /1,1/                o,,o~Ao DATA FROM GA-8129 AND NAA-SR-9374 600    700    80o  900  ICo      I100  200 1300 TEMPERATURE( C)
Figure 4.3, Thermal Pressurization in Fuel and Hydriding Ratios Long term operations with steady state fuel temperatures exceeding 750'C (1023*K) may have time- and temperata ure-dependent fuel growth.5 Mechanisms contributing to the growth are identified as fission recoils and gaseous fission products, strongly influenced by thermal gradients. Analysis of steady state operating fuel,temperatures is provided in section 4.6, with pulsing operations fuel temperatures in Appendix 4.1.
(6) Zr water reaction Among th.e.chemical properties of U-ZrH: and ZrH, :the. reaction rate of the hydride with water is of particular interest. Since the hydriding reaction &#xfd;is exothermic, water will react more readily with zirconium than with zirconium hydride systems. Zirconium is frequently used in contact with water in reactors, and the zirconium-waterreaction is not a safety hazard.
Experiments carried out at GA Technologies show that.the zirconium hydride systems have a relatively low chemical reactivity with respect to water and air . These tests have involved the, quenching with water of both powders and solid specimens of.U-ZrH after~heating to as high as; 850&deg;C, and of solid U-Zr alloy after, heating to as high as 1200*C. Tests have also been made to' determine the extent to which fission.products are removed from the surfaces of thefuel' elements at room temperature. Results prove that, .because of the high resistance to leaching, a large fraction of the fission products is retained in even completely unclad U-ZrH fuel..:
(7) Mechanical Effects At room temperature the hydride is like ceramic and shows little ductility. However, at the elevated temperatures of interest for pulsing, the material is found to be more ductile. The effect of very large thermal stress on hydride fuel bodies has been observed in hot cell observations to cause relatively widely spaced cracks which tend to be 5General Atomics Technical Report E-117-833 6
NUREG/CR-2387 Credible Accidents for TRIGA and TRIGA Fueled Reactors, S. C. Hawley,S. C. and Kathren, R. L.,
PNL-4208 (1982)
Page 4-9 I
CHAPTER 4: REACTOR                                                                  A    01/2012 either radial or normal to the central axis and do not interfere with radial heat flow. Since the segments tend to be orthogonal, their relative positions appear to be quite stable. During fabrication, a molybdenum disk is placed between the lowest fuel mass and the lower axial-graphite reflector, minimizing potential for interaction that might affect the graphite and cause position changes in fuel meat that has developed surface imperfections. Anticipated mechanical effects from operation of the reactor are not expected to create conditions that challenge fuel performance.
7 (8) Fission Product Release Early in development of U-ZrHx fuel, experiments were performed 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 test conducted, with water flowing across the surface of the foil to remove fission products for analysis. The test was performed for 1 day and for 8 days with the total fractional fission product loss calculated to be between 10-7 and 10-s from preferential leaching of radionuclides, with gasses evolving from depths of 2.6 plm in the foil, and particulate from 22 A. Acceptable 8 upper values for release fraction are 1.0 x 10-4 for noble gases and iodine contained within the fuel, 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 noble gases, which is in SARs for other reactor facilities [NUREG-1390, 1990].
B.      Cladding The 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 to release of those fission products that migrate to escape the fuel matrix surface. As noted previously, the free hydrogen in the space within the fuel element pressurizes the interior of the fuel element when fuel temperature is elevated during reactor operations. Power levels are acceptable if they do not result in temperatures that produce stress from the gas pressure that challenges the integrity of the cladding. A cylinder is considered a thin shell if wall thickness is less than about 10% of the radius and the classic equation for hoop stress created by internal pressure is:
o= P.r/t where:
oe is the hoop stress P is internal pressure r is inside radius t is the wall thickness 7  General Atomic report GA-655, Uranium-Zirconium-Hydride Fuel Elements, Merten, Stone, Wallace (1959) 8  NUREG/CR-2387, op. cit.
Page 4-10
THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 4                                            R~eETL I  01/2012 For                                                stress is      times the internal pressure. Fig.
4.4A provides temperature dependent ultimate strength and the 0.2% yield, and Fig. 4.4B shows where the hoop stress induced by the internal pressure intersects with ultimate strength. This intersection corresponds to a fuel temperature of 950'C for cladding temperatures greater than 500'C.
cc 103 400 500        600        700        800        900        1000  i'300 TEMPERATURE ( 0 C)
Figure 4.4A, Temperature and Cladding Strength for 0.2% Yield Page ,4-11
CHAPTER 4: REACTOR I 01/2012 Therefore, if fuel and cladding temperature remains below 950&deg;C with cladding temperatures greater than 500&deg;C, the stainless- steel cladding will not fail from overpressure. For cladding temperatures less than 500TC, hydrogen pressure from peak fuel temperature of 1150TC would not produce a stress in the clad in excess of its ultimate strength. The limiting fuel temperature and pressure is therefore the design basis for the UT TRIGA fuel. TRIGA fuel with a hydrogen to zirconium ratio of at least 1.65 has been pulsed to temperatures of about 1150TC without damage to the clad 9 .
1OS.
ULTIM~ATE 5TRENGTH 304. SS 1:  S T R E NZ rH 1 .
65 o*,
1-io2.
10 2_-;
500      600          700        800          900        1000        1100 TEMPERATURE (*C)
Figure 4.4B, Temperature, Cladding Strength, and Stress 9 "Annual Core Pulse Reactor," General Dynamics, General Atomics Division report GACD 6977 (Supplement 2),
Dee. J. B., et. A].
Page 4-12
THE UNIVERSITY OF TEXAS TRIGA IIRESEARCH REACTOR                                N0ETL        01/2012 0.00  _
SAFETY ANALYSIS REPORT, CHAPTER 4 There are several reasons why the gas pressure should be less for the transient conditions than the equilibrium condition values would predict. For example, the gas diffusion rates are finite; surface cooling is believed to be caused by endothermic gas emission which tends to lower the diffusion constant at the surface. Reabsbr ption takes placelconcurrently on the cooler hydride surfaces away from the hot spot. There is evidence for a low permeability oxide film on the fuel surface. Some iocal, heat transfer does take place during the pulse time to cause a less than adiabatic true surface temperature.
4.2.2    Control Rods and Drive Mechanisms The control rods and drive mechanisms consist of (A) control rods, (B) standard, (or stepper) control rod drives, (C)transient rod drives, (D)control functions, and (E) system operation. The UT TRIGA reactor was installed with 4 control rods, three standard rods magnetically coupled to the control rod drive, and one pulse rod pneumatically coupled to the control rod drive. One of the standard rods, the regulating rod, is capable of being either automatically controlled with instrumentation and control systems described in Chapter 7 or manually from the reactor control console. The other control rods are manually shimmed. Principle design parameters for the control rods are provided in Table 4.4.
A.      Control Rods The standard/stepper control rods (regulating and shim) are sealed 304 stainless steel tubes approximately 43 in. (109 cm) long by 1.35 in. (3.43 cm) in diameter in which the uppermost 6.5 in. (16. 5 cm) section is an air void, followed by 15 in. (38.1 cm) of a neutron absorber, solid boron carbide. Standard control rods have a fuel follower attached so that as the control rod is withdrawn from the core the water channel is filled with a fuel element as illustrated in Fig. 4.6.
The fuel follower, 15 in. (0.381 cm) of U-ZrH1 .6 fuel, is immediately below the neutron absorber of the standard control rods. The bottom 6.5 in. (16.5 cm) of the standard control rod is an air void. Table 4.4 summarizes control rod design parameters.
Table 4.4, Summary of Control Rod Design Parameters Cladding Material          Aluminum                        SS 304 OD                1.25 in.        3.18 cm        1.35 in.      3.43 cm Length            36.75 in..      93.35 cm      43.13 in.      109.5 cm ...
Wall thickness    0.028 in.        0.071 cm      0.02 in.      0.051 cm Poison Section Material          Boron Carbide OD                1.19 in.        3.02 cm        1.31 in.      3.32 cm Length            15 in.          38.1 cm        14.25 in.      36.20 cm Follower Section Material          Air                            U-ZrH.1 6 OD                1.25 in.        3.18 cm        1.31 in        3.34 cm Length            20.88 in.        53.02 cm Page 4-13
CHAPTER 4: REACTOR                                                                  1      01/2012 The transient (also called safety-transient or pulse) rod is a sealed, 36.75 in. (93.35 cm) long by 1.25 in. (3.18 cm) diameter tube containing boron in graphite as a neutron absorber. Below the absorber is an air filled follower section. The absorber section is 15 in. (38.1 cm) long and the follower is 20.88 in. (53.02 cm) long. The transient rod passes through the core in a perforated aluminum guide tube. The tube receives its support from the safety plate and its lateral positioning from both grid plates. It extends approximately 10 in. (25.4 cm) above the top grid plate. Water passage through the tube is provided by a large number of holes distributed evenly over its length. A locking device is built into the lower end of the assembly.
Control rods are withdrawn out of the core through the upper grid plate; when fully inserted the followers extend down through the lower grid plate. All fuel element position penetrations in the upper grid plate are identical; the lower grid plate (an excerpt in Fig. 4.5, fully described later in Chapter 4) has a set of 11 penetrations in the C and D rings (shaded in gray and black in Fig. 4.5, black representing the current configuration) with the same diameter as the upper grid plate. One of these penetrations in reserved for the central thimble (position Al) while the others are available for use as control rod positions. A safety plate is mounted below the lower grid plate as shown in Fig. 4.6, so that the control rod cannot exit the core region in the downward direction.
Figure 4.5, Lower Gird Plate Control Rod Positions Control rod worth is principally a function of control rod dimensions and location, experiment facilities in the core, with lessor influence by fuel and control rod burnup. Estimated control rod from the 1991 preliminary safety analysis report is provide in Table 4.5, along with the worth of each control rod as measured in June 2011. Sections of the control rod are separated and secured by 1-inch magneform fittings.
Page 4-14
THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 4 NETI I    01/2012 Table 4.5, Control Rod Informaton Rod      Location          Diameter                  Estimated,(1991). Current (2011)
In.      cm.      %Ak/k                            $S Transient Rod    C Ring          1.25      3.18      2.1            3.00          3.10 Regulating Rod    C ring        1.35      3.43      2.6            3.71          2.82 Shim 1            D ring        1.35      3.43      2.0            2.86          2.52 Shim 2            D ring        1.35      3.43      2.0            2.86          3.07
                -U&deg; Air                  16.5 cm T.I J B4C F          38.1 cm LN U-ZrH                    38.1 cm Air K,
6.5cmc..
Figure 4.6, Standard Control Rod ConfiguratioIn A threaded fitting at the end of each control rod connects to a series of shafts that connect to control rod drive mechanisms mounted' on a bridge that spans the reactor pool. The top section of the connecting shafts for standard rods passes through a hole in the bottom of a tube supported by the control rod drive housing. The tube is designed with slots that provide a hydraulic cushion for the rod during a scram, and also prevent the bottom of the control, rod from impacting the safety plate.
The shaft is secured to a cylinder that rests on the bottom of the housing when the rod is fully inserted. The top of the cylinder is secured to an iron core, engaged by an electromagnet for fail-safe control. The electromagnet is at the bottom of a small shaft controlled by the control rod drive mechanism. When the electromagnet is energized, the iron core is coupled to the drive unit.
Page 4m15
CHAPTER 4: REACTOR CHPE :RACO I 01/2012 121 The top section of the transient rod is connected to a single acting pneumatic cylinder which operates on a fixed piston, that couples the connecting rods to the drive. The transient rod drive is mounted on a steel frame that. bolts to the bridge. Any value from zero to a maximum of 15 in. (38.1. cm,):. of rod may:ibe withdrawn from the core; rod travel is limited by administrative control-not to exceed to the maximum licensed step insertion of reactivity.
B.      Standard Control Rod Drives The rod drive mechanism for the standard rod drives is an electric stepping-motor-actuated linear drive equipped with a magnetic coupler and a positive feedback potentiometer. A stepping motor drives a pinion gear and a 10-turn potentiometer via a chain and pulley gear mechanism. The potentiometer is used to provide rod position information.
                                                                    ~MAGNET WIRE CONDUIT
                                                                    -MAGNET DOWN ADJUSTMENT SCREW' MAGNET DRAW TUBE                                -MOTOR BIAS ADJUSTMENT T    CENTER SWITCH ROD DOWN LIMIT SWITCH-'..                Y_-,        IAOUNTING PLATE WIRE CONDUIT PULL-ROD ROD MOTOR PULL-ROD DRAW TUBE
                          *PULL LOCK
                                                                  .CONNECTING ROD Figure 4.7, Standard/Stepper Motor Control Rod Drive Page 4-16
THE UNIVERSITY OF TEXAS TRIGA 11RESEARCH REACTOR                      000 o      f        01/2012 SAFETY ANALYSIS REPORT, CHAPTER 4                                      00_              I The pinion gear engages a rack attached to the magnet draw tube. An electromagnet, attached to the lower end of the draw tube, engages an iron 3rm-nature. The armature is screwed and pinned into the upper end of a connecting rod that terminates at its lower end in the control rod. When the stepping motor is energized (via the rod' control UP/DOWN switch on the reactor control console), the pinion gear shaft rotates, thus raising the magnet draw tube, the armature and the connecting rod will raise with the draw tube so that the control rod is withdrawn from the reactor core. In the event of a reactor scram, the magnet' is de-energized and the armature will be released. The connecting rod, the piston, and the control rod will then drop, thus reinserting the control rod.
Stepping motors operate on phase-switched direct current power. The motor shaft advances 200 steps per revolution (1.8 degrees per' step). Since current is maintained -on the'motor windings when the motor is not being stepped, a high holding torque is maintained. The torque versus speed characteristic of a stepping motor is greatly dependent on the drive circuit used to step the motor. To optimize the torque characteristic for the motor frame size, a Translator Module was selected to drive the stepping motor. This combination of stepping motor and translator module produces the optimum torque at the operating speeds of the control rod drives. 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 Drive The safety transient control rod*drive is operated with a pneumatics rod drive. Operation of the transient rod drive is controlled from'the reactor control console. The transient rod is a scrammable rod operated in both pulse and steady-state modes of reactor operation. During steady state operation, the transient rod will function as an alternate safety rod with air continuously supplied to the rod. Rod position is thus controlled by'Operation of an electric motor that positions the air drive cylinder. The position of the transient control rod and its associated reactivity worth will generally dictate removal of the rod as the first step of a startup for steady-state operation. Rod withdrawal speed is about 28 ipm (1.E9cm/s).
The transient rod drive is a single-acting pneumatic cylinder with its piston attached to the transient rod through a connecting rod assembly., The piston rod passes through an air seal at the lower end of the cylinder. Compressed air is supplied to the lower end of the cylinder from an accumulator tank when a three -way solenoid valve located in the piping between the accumulator and cylinder is energized. The compressed air. drives the piston upward in the cylinder 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 the end of its travel, the piston strikes the anvil of an oil filled hydraulic shock absorber, which has a spring return, and which decelerates the piston at a controlled rate over its last 2 in. (5 cm.) of travel. When the solenoid is de-energized, a solenoid valve cuts off the compressed air supply and exhausts the pressure in the cylinder, thus allowing the piston to drop by gravity to its original position and restore the transient rod to a position fully inserted in the reactor core.
Page 4-17
CHAPTER 4: REACTOR I    01/2012 V EN.T VALVE Figure 4.8, Transient Rod Drive The extent of transient rod withdrawal from the core during a pulse is determined by raising or lowering the de~coupled cylinder, thereby controlling the distance the piston travels when air is applied. The cylinder has external threads running most of its length, which engage a series of ball bearings contained in a ball-nut mounted in the drive housing. As the ball-nut is rotated by a worm gear, the cylinder moves up or down depending onr!the direction of worm gear rotation.
A ten-turn Potentiometer driven by the worm shaft provides a signal indicating the position of the cylinder and the distance the transient rod will be ejected from the core. Motor 'operation for pneumatic cylinder positioning is controlled by a switch on the reactor control console. The magnet power key switch on the control console power supply prevents unauthorized firing of the transient rod drive.
Attached to and extending downward from the transient rod drive housing is the rod guide support, which serves several purposes. The air inlet connection near the bottom of the cylinder 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 rod assembly that moves the transient rod. The flanged connector stops the downward movement of the transient rod when the connector strikes the damp pad at the bottom of the rod guide support. A microswitch is mounted on the outside of the guide tube with its actuating lever Page 4-18
THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 4
* eNETI I    01/2012 extending 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 the instrument 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 withdrawn position and the rod drops into the core from the fu!! out position in less than 1 second.
D.        Control Functions Instrumentation and controls provide protective actions through the control rod system, as described in Table 4.6. A trip signal from the reactor protection system or the reactor control systems will deenergize the electro magnets and the pulse rod air solenoid valve previously described which allows gravity to insert the control rods.
Table 4.6, Summary of Reactor SCRAMs Limiting Trip Setpoint Measuring Channel                Steady          Pulse                        Actual Setpoint State SS - 1050 (NPP/NP) 1080 NM 1100 kW        2000 MW                        - 1 Pl      910 NP Maximum thermal power                                                          Pulse - 1910 NPP Power Channel High power          110%            110%
Detector High Voltage              80%            8C%
High Fuel Temperature                      550&deg;C Magnet current loss Manual Scram DAC and CSC watchdog timers In addition, the reactor control system (described in Chapter 7) has interlocks to prevent various conditions from developing. Table 4.7 is a summary of the functions.
Table 4.7, Summary of Control Rod Interlocks INTERLOCK                  SETPOINT                              FUNCTION/PURPOSE Inhibit standard rod motion if nuclear instrument S2        startup channel reading is less than instrument sensitivity/ensure nuclear instrument startup channel is operating Pulse Rod Interlock      Pulse rod inserted            Prevent applying power to pulse rod unless rod inserted/prevent inadvertent pulse Prevent withdrawal of more than 1 rod/Limit Muti      dWithdrawal              s    , me        maximum reactivity addition rate (does not apply in automatic flux control)
Prevent withdrawing standard control rods in pulse Pulse Mode Interlock    Mode switch in Hi Pulse      mode
  .Pulse-Power Interlock    10 kW                        Prevent pulsing if power level is greater than 10 kW These safety settings are conservative in the sense that if they are adhered to, the consequence of normal or abnormal operation would be fuel and clad temperatures well below the safety limits indicated in the reactor design bases. Because of the conservatism in these safety Page 4-19
CHAPTER 4: REACTOR                                                                        01/2012 settings, it is reasonable that at some later date less restrictive safety system settings could be assigned in conjunction -with upgrading of the reactor to operate at higher steady-state power levels or in the pulsing mode while using the same fuel and core configuration.
Administrative limitations are imposed for the excess reactivity, transient conditions and coolant water temperature as follows:
: 1) Maximum core excess reactivity of 4.9% Ak/k ($7.00) with a shutdown margin of at least 0.2% 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, and
: 3) Core inlet water temperature of 48.9&deg;C.
E.      Evaluation of the Control Rod System The reactivity worth and speed of travel for the control rods are adequate to allow complete control of the reactor system during operation from a shutdown condition to full power. The TRIGA system does not rely on speed of control for reactor safety; scram times for the rods are measured periodically to monitor potential degradation of the control rod system. The inherent shutdown mechanism (temperature feedback) of the TRIGA prevents unsafe excursions and the control system is used only for the planned shutdown of the reactor and to control the power level in steady state operation. A scram. does not challenge the control integrity or operation, or affect the integrity or operation of other~reactor systems.
4.2.3    Neutron Moderator and Reflector (Core Structure)
The UT TRIGA core is supported within a reflector assembFy. The reflector assembly supports (A) an upper grid plate, (B) core barrel and reflector, and (C) lower grid plate, shown in Fig.
4.9a/b. The upper and lower grid plates provide alignment and support for the fuel elements.
A.      Upper grid plate The upper grid plate provides alignment for fuel elements and control rods, and (in conjunction with the top fuel assembly) space for cooling flow. The.top grid. plate is fabricated from a circular aluminum plate 5/8 inches (1.59 cm.) thick and 21..6 in. (55.245 cm) diameter, anodized to resist wear and corrosion. The top of the upper grid plate is 59 in. (150 cm.) above the bottom of the pool.                                                                    diameter are established on a triangular pitch of 1.714 in. (4.35 cm), separated by radial fuel arrays integrated 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 and guide tubes, the pneumatic transfer tube, and the central thimble. Small 0.203 in. (8 mm) holes at various positions in the top grid plate permit insertion of wires or foils into the core to obtain Page 4-20
THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 0    0  No NET_
T          01/2012 SAFETY ANALYSIS REPORT, CHAPTER 4 flux data. The flux probe holes are counter sunk/chamfered to (820) to 0.31 in. (11 mm). The center fuel element position is reserved as an experimental facility. The outermost fuel positions in the radial arrays are not fabricated for fuel insertion. Upper grid plate penetrations are summarized in Table 4.8.
01<
Ito.                        2W
                                                                            '                            /
Figure.4.9a, UT TRIGA Core                                    Figure 4.9b, Core Top View The 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 grid plate using % in. (0.953 cm) holes centered along each of the hexagonal faces of the G ring fuel positions.
Table 4.8, Upper Grid Plate Penetrations
                    -Penetration Function..:              Size              "
Fuel Elements    -,                  1.505 in. (3.8227 cm), diameter 3Telement              -  .1.2          in. (3.048 cm). radius !
6/7-Element                          2.2 in. (5.588 cm) radius Upper grid plate alignment            3/8 in. (0.9525 cm) diameter Flux probes                          0.203 in. (0.5156 cm) diameter..
Fuel positions are :indexed by !etters denoting a "ring" where elements are colfinear with respect to the adjacent radial array fuei positions; A is the centralring position and G is furthest from the center. One radial array is used as a reference position, and the fuel positions range from 1 at the index to the maximum value for the ring, except for the G ring. Since the vertices of 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 facility encompassing the, central thimble since the central thimble does not contain fuel).
Page 4-21
CHAPTER 4: REACTOR I    01/2012 The inserts mesh in slots milled in the circular grid plate cutouts; 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 element facilities permit specimen as large as 2.4 in. (6.1 cm).
In addition to the experiment facilities that replace fuel positions, the current core configuration 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 or potentially displaced by core equipment. For control rods, only currently used positions are identified; there are alternate positions useable for control rods.
Table 4.9, Displaced Fuel Spaces Facility                      Core Location    .
Page 4-22
0 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 4 0    4 nETL I    01/2012 B.      Reflector The 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 fuel region.
(1) Radial Reflector. The radial reflector is a 10.2 in. (25.91 cm) graphite ring with an inner diameter 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 surface of the top section (Fig. 4.9a), with flat welded plates tying the top and bottom sections to the lift points. Angle plate structures are welded on the outer perimeter as points to secure the power level detectors. A 3 inch (7.62 cm.) wide well is fabricated in the top section (Fig. 4.11b),
and blocks with threaded penetrations are welded at the inner perimeter of the well to allow securing the rotary specimen rack (an experimental assembly) in the well.
Figure 4.11a, Reflector Top Assembly            Figure 4.11b, Reflector Bottom Assembly The 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 of the reactor core, with one "through port" (Fig. 4.12a) and a 10 in. (25.3 cm.) cylinder cut from the inner surface to allow a 3 inch wide experimental facility surrounding the core.
1 Figure 4.12a, Graphite Reflector, Through Port Figure 4.12b, Graphite Reflector Through port Detail Page 4-23
CHAPTER 4: REACTOR                                                                                        01/2012 I    01/2012 CHAPTER 4: REACTOR                                                                    ------      a--
Figure 4.12c, Graphite Reflector, Radial & Piercing-Beam Ports The through port has a rectangular water-filled cut-out between the core shroud and the beam port penetration (Fig. 4.12b). Aluminum canisters that mate with the beam ports are nested in the reflector in two of the beam ports, one radial and one tangential (Fig. 4.12c, Fig. 4.13a/b).
The third beam port (radial) penetrates the core shroud (Fig. 4.13c).
Figure 4.13a, Tangential Beam Fort Insert                        Figure 4.13b, Radial Beam Port inert Figure 4.13c, Inner Shroud Surface (2) Graphite Rods. Graphite dummy elements may be uFAd to fP!. grid positions not filled by the fuel-moderator elemerts or other core compound, ,. They are of the same general dimensions and construction a&#xfd;. the fuel-moderator elements, !&#xfd;ut are f;lled entirely with graphite and are clad witi        !,'.hr*ir, :m.
(3) Axial Reflector. Graphite cylinders are placed above and below the fuel in the fuel elements. Fuel element construction was previously discussed.
C.      Lower grid plate The lower grid plate (Fig. 4.14) provides alignment for fuel elements and control rods, and (in conjunction with the top fuel assembly) space for cooling flow. The lower (or bottom) grid Page 4-24
THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 4 0E~
I  01/2012 plate is fabricated from a circular aluminuLr. p*2t.e &#xfd;..75 inches (3.81 cm.), anodized to resist wear and corrosion. The top of the bottom *id piate is S.9 in. (25.19 cm.) above the bottom of the pool. The bottom grid plate is fabricated with fuel position penetrations and penetrations matching the flux probe holes on toe sarne, center 5s the upper grid plate, but also contains penetrations that support alignment cf the 3, G-,af:d 7 element facilities (Table 4.10). All but 11 fuel penetrations in the lower grid plate are sarnfller than the diameter of the fuel element and chamfered to provide a surface supporting triflutes on the bottom of the fuel element elements.
Table 4.10, Lower Grid Plate Penetrations Penetration Function                      Size Central thimble                                  1.505 Control Rod                                      1.505 Flux Hole Probes                                8 mm 3-Element Alignment                              3/8 in.
Lower grid plate alignment Lower Grid Plate Support                                        Lower Grid Plate 11 Reflector Canister Bottom View                                Grid Plate in Core Shroud Figure 4.14, Reflector Component and Assembly Views Ten lower grid plate penetrations are the same diameter as the penetration in the upper grid plate, providing clearance for the central thimble and control rods. Since only 4 controls rods Page 4-25
CHAPTER 4: REACTOR                                                                        01/2012 are installed, unused control rod positions (i.e., large diameter holes) can be used for fuel with an adapter to support positioning the fuel above the lower grid plate (Fig. 4.15).
Figure 4.15, Fuel Element Adapter 4.2.4  Neutron Startup Source The reactor license permits the use of sealed neutron sources, including a is a standard sealed neutron source, encapsulated in stainless steel. The source is maintained in an aluminum-cylinder source holder of approximately the same dimensions as a fuel element. The source holder is manufactured as upper and lower (threaded) sections. The top of the lower section is at the horizontal centerline of the core. A soft'aluminum ring provides sealing against water leakage into the cavity.
The source holder may be positioned in any one of the fuel positions defined by the upper and lower grid plates. The upper end fixture of the source holder is similar to that of the fuel element; the source holder can be installed or removed with the fuel handling tool. In addition, the upper end fixture has a small hole through which one end of a stainless steel wire may be inserted to facilitate handling operation from the top of the tank.
4.2.5  Core support structure.
The core support structure includes (A)a platform supporting the reflector and core structure, and (B)a "safety plate" thaftprevents the control rods in a failure mode from falling out of the core.
A.      Core Support Platform The reflector assembly rests on a platform (Fig. 4.16) constructed of structural angle 6061-T5 aluminum with a 3 in. x 3 in. x %in. (7.62 cm x 7.62 cm x 0.953 cm) web. Aluminum 6061-T651 plate is used for safety plate support pads (%in., 1.905 tmP), cross braces (% in., 0.953 cm.), and platform support pads (Y/in., 1.27 cm.). Angle aluminumr is inserted 9 iii. (22.86 cm)frorn two edges to support the safety plate, with angle bracing on the edges perpendicular to the safety plate supports.
Page 4-26
THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 4 IMeETI I    01/2012 Core Support Top View                Core Support Side View              Core Support Side View Figure 4.16, Core Support Views The platform top surface is 30 X in. X 30 Y4 in., with the top surface 16 Y4 in. above the pool floor. Surfaces are matte finished for uniform appearance with shot cleaning and peening using glass beads (MIL-STD -852).
                            -h                                                1 Core and Support Structure Assembly                      Core and Support Assembly Isometric Figure 4.17, Core and Support Structure Views B.      Safety plate The safety plate (Fig 4.18) limits the distance that a control rod can fall to less than 17.44 in.
(44.30 cm) below the top surface of the lower grid plate. The safety plate is an aluminum plate Y2 in. (1.27 cm.) thick, 12 in. (30.48 cm) X 13.5 in. (34.29 cm), anodized to resist wear and corrosion (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 previously described, the bottom grid plate has a set of through-penetrations for optional placement of control rods. A special adapter is required to support fuel elements when these locations are used for fuel. The adapters have a central alignment pin that fits within holes in the safety plate, and an offset keeper-pin that prevents the adapter from rotating around the central pin.
Page 4-27
CHAPTER 4: REACTOR                                                                                    01/2012 Figure 4.18, Safety Plate 4.3 Reactor Pool The reactor pool is a 26 foot, 11.5 in. (8.2169 m) tall tank formed by the union of two half-cylinders with a radius of 6 1/2z feet separated by 6 Y feet (1.9812 m). The bottom of the pool is 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 minimum level of 6.5 m (21.35 ft.) required for operations. Volume of water in pool (excluding the reflector, beam tubes and core-metal) is 40.57 mi3 and 32.50 m 3 for the nominal and minimum-required levels. Table 4.11 summarizes reactor coolant system design.
Table 4.11, Reactor Coolant System Design Summary Material                  Aluminum plate (6061)
Reactor Tank                      Thickness                        Y4 in. (0.635 cm)
Volume (maximum)                    11000 gal (41.64 M 3)
Pipes                      .. Aluminum 6061 Coolant Lines                                                Iron-Plastic Liner, 316 SS Ball and Stem Fittings      .            . Aluminum (Victaulic)
Type            .                Centrifugal Coolant Pump                        Material                          Stainless Steel Capacity                        250 gpm (15.8 Ips)
Type                            Shell & Tube Materials (shell) "                    Carbon steel Materials (tubes)                    304 stainless steel HeatExchanger                      Heat Duty Flow Rate (shell),            "
Flow Rate (tubes)      .    -
Tube Inlet            ....          ..
Typical Heat Exchanger            Iu  e uOutlet Operating Parameters Shell Inlet Shell Outlet Page 4-28
THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR                00          T          01/2012 SAFETY ANALYSIS REPORT, CHAPTER 4                              _ 0000 The pool (Figs. 4.19a/b/c) is fabricated from sheets of 0.25 in. (0.635 cm) 6061 aluminum in 4 vertical sections welded to a Y2 in. thick aluminum plate. Full penetration inspection was performed 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 the beam 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 threaded connection to perform a leak test during fabrication. A 2 in. X 2 in. X Y4 in. (square) aluminum channel was rolled and welded to the upper edge of the tank.
Beam port penetrations are fabricated around the core to allow extractior of radiatior beams to support experiments. The beam ports are centered 90.2 cm (35 in.) above the pool floor, 7.2 cm (2.83 in.) below the core centerline. The section of the beam ports that are an integral part of the pool include an in-pool section, interface with the pool wall, and a section.
extending outside of the pool.
In pool sections are 6 in. (15.4 cm) in diameter, with a 0.635 cm (0.25 in.) wall thickness. The in pool section for BP 1 and 5 is 6 in. (15 cm), while the remaining in-pool beam port sections are much longer. Supports (2 in. X 2 in. X Y4 in. aluminum angle bracket) are welded at the bottom of the pool and directly onto. BP 2, 3, and 4 because of the extended lengths. BP 2 and 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 extends approximately 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 (14.8125 in.) from the in-pool end. BP 4 extends 43 cm into the pool (16.94 in.) with the support 7.62 cm (3 in.) from the in pool end. Beam port 1 and 5 are aligned in a single beam line. A flight tube inserted into BP 1/5 extends through the reflector near the core shroud; BP 1 and 5 are equipped with a bellows to seal a neutron flight-tube. Beam ports 2, 3, and 4 are sealed at the in-pool end. BP 2 is tangential to the core shroud, offset 34.29 cm (13 /2 in.) from core center rotated 3 0 0 with respect to BP 3. Beam port 3 is 90' with respect to BP 1/5, aligned to the center of the core. Alignment of BP 4 is through the core center, rotated 60' from BP 3.
The beam port interface with the pool wall includes a reinforcing flange on the inner pool wall. The flange is 3/8 in. thick, 11 in. in diameter. The flange is welded on the outer Page 4-29
CHAPTER 4: REACTOR                                                                    01/2012 diameter 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 pool walls. A stainless steel (304) ring is machined for a slip fit over the extension. The ring is welded to 6 5/8 in. diameter stainless steel pipe (SST 304W/ASTM 312) extending the flight tube 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, the in-pool beam port supports are welded to the pool floor.
Detection of potential pool leakage could occur in a number of ways.
: 1.      Pool water level is maintained approximately 8.1 m above the pool floor, and monitored with an alarm on the control room console. A sudden decrease in pool water will create a condition that alerts the reactor operator at the controls.
: 2.      Losses to evaporation are compensated by makeup water. Makeup water usage is closely monitored, and changes in makeup requirements or increases in makeup water that do not correspond to power level operation are a primary pool-leak indicator.
: 3.      French drains around. the reactor pool shielding ,foundation are collected in a
                .sump, and sampled periodically. Increases in radiation levels from the sump
                ..(particularly tritium) could indicate pool leakage.,
4.4    Biological Shield Pool water system and. shield structure (Fig. 4.20), design combine to control the effective radiation levels from, the, operation of the reactor.: One goal of the design is a radiological exposure constraint of 1 mrem/hour for accessible areas of the pool and shield system. Dose
.levels assume a full power operation level of 1.500 megawatts (thermal). Radiation doses above the pool' and at specific penetrations into or through the shield may exceed the design goal. The reference. case design is a solid structure without any system penetrations.
Tank assembly is by shop fabrication. A protective layer of epoxy paint and bitumen coal tar pitch with paper provides a barrier between the aluminum pool tank and the reactor shield concrete.
Page 4-30
THE UNIVERSITY OF TEXAS TRIGA 11RESEARCH REACTOR              N  0001/2012 0
SAFETY ANALYSIS REPORT, CHAPTER 4 thick foundation pad supports the reactor pool and shield structure.
Standard weight concrete,'                          comprises the foundation pad. High density
: concrete, of Five beam tubes atthe level of the reactor -provide experimental access, to:reactor neutron and gamma radiations. Two of the tubes combine to penetrate the complete reactor pool and shield structure from one side to the 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 leakage when the tubes are not in use. Beam port details are discussed in Chapter 10. A summary of significant component elevations and control functions is provided in Table 4.12.
Page 4-31
CHAPTER 4: REACTOR CHAPTER 4: REACTOR I    01/2012 01/2012 Table 4.12, Significant Shielding and Pool Levels Levei                    Notes Parameter of Interest (meters)
CONCRETE PAD FLOOR SAFETY PLATE GRID PLATE CORE BOTTOM BEAM PORT CL CORE CL CORE TOP                                                                  1 GRID PLATE MAIN LOWER SHIELDING TRANSITIONAL CONCRETE STEP SHIFT TO HIGH DENSITY C()NCRETE MIN CORE LEVEL (TS)
VACUUM-BREAKERS LOW POOL LEVEL SCRAM LOW POOL LEVEL                                                          op LOW POOL LEVEL ALARM NORMAL POOL LEVEL HIGH POOL LEVEL HIGH POOOL LEVEL ALARM'1 TOP OF TOP LEVEL 4.5 Nuclear Design The characteristics and operating parameters of this reactor,&#xfd; have been calculated and extrapolated using experience and data obtained from existing TRIGA reactors as bench marks in evaluating the calculated data. There are several TRIGA systems with essentially the same core and reflector relationship as this TRIGA so the values presented here are felt to be accurate to within 5% but, of course, are influenced by specific core configuration details as well as operational details. An operational core of                              3 fuel followed control rods, and one air followed control rod is to be arranged in 5 rings with a central, water filled hole. Dimension of the active fueled core, approximated as cylinder, 15 in. (
cylinder radius is calculated as the average radius ofa hexagonal fuel array with 4.5.1  Normal Operating Conditions Reactivity worth of core components is generally determined by calculation and/or comparison of the reactivity worth associated with the difference in the reactivity worth of control rod positions in the critical condition, component installed and component removed. The 1992 UT SAR provided data indicating estimated worth of the control rods (Table 4.13). Control rod worth is influenced by core the experiment configuration, with significant impact from the large in core irradiation sites. Table 4.13 provides the worth of the control rods in the current configuration (3 element facility in Eli, F13, and F14). Change in core configuration require validation that control rod worth is not affected by the experiment facility, or re-establishment Page 4-32
THE UNIVERSITY OF TEXAS TRIGA 11RESEARCH REACTOR                              00      A TI            01/2012 SAFETY ANALYSIS REPORT, CHAPTER 4                                        ....
of the control rod worth followed by verification that the limiting conditions for operation are met.
Table 4.13 Control Rod Worth Reference                      Current (2011),
Control Rd        .          Position              Worth            Position  ; Worth.
Transient rod                C ring        2.1% Ak/k    $3.00        C-1        $3.10 Regulating rod              C ring        2.6% Ak/k    $3.71        C-7        $2.82 Shim 1                        D ring .      2.0% Ak/k    $2.86        D-14        $2.52 Shim 2                        D ring        2.0% Ak/k    $2.86        D-6        $3.07 4.5.2 Nominal Reactivity Worth Values Reactivity values for core components based on calculations and observations are provided in Table 4.14, with Technical Specifications values in bold face type. Current values are based on measurements; nominal values are calculations frOm indicated sources.
                                          'Table 4.14, Reactivity Values
                                    $"TS                                                CURRENT    NOMINAL Parameter                            LIMIT      VALUE        VALUE 10 Reactor Reference Data Notebook, Safety Analysis, report Table 4-5; SAR Table 4-6 indicates CT Fuel $0.90, CT Void -$0.15, PNT Void -$0. 10, RSR void -0.20 "3-Element Experiment Authorization 12 Significant deviation from values in 3-Element Experiment Authorization (cf. E-Ring -$0.50 & D-Ring $0.95)
Page 4-33
CHAPTER 4: REACTOR I      01/2012 Table 4.14, Reactivity Values TS      CURRENT  NOMINAL LIMIT      VALUE      VALUE 4.5.3  Reactor Core Physics The 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 step insertions of reactivity is the prompt negative temperature coefficient (Fig. 4.21) associated with the TRIGA fuel and core design. This temperature coefficient allows a greater freedom in steady-state operation as the effect of incidental reactivity changes occurring from the experimental devices in the core is greatly reduced.
44.0 3.0 i-2.0
            .5 t            I            I n            I 0
0          200            100              600          800        1000 POWER  (KW)
Figure 4.21, Reactivity Loss with Power Page 4-34
THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 4 SNETL I 01/2012 A.      Reference Calculations A reference calculation of neutron flux distribution across the core was performed by General Atomics 13 . The calculations were accomplished on an .BM-7090 using General Atomics (diffusion theory based) codes GAMBLE and GAZE, and GAIV1-I. GAM-l is a fast neutron (using P1 treatment), temperature dependent (using methods developed by Nordhiem) cross section calculations for neutrons above 1 eV. GATHER-I was used to calculated cross sections below 1 eV. Homogenization was accomplished by the transport theory code DSN for group-dependent disadvantage factors (a second homogenization was accomplished for inhomogeneities in cells with control rods).            No attempt was made to account for spatial variations in core temperatures. Basic core data for the calculations is provided in Table 4.15, with selected nuclear properties in Table 4.16. The model varies from the UT TRIGA reactor in specification of control rods, with one poison and three aluminum followers, where the UT TRIGA uses one aluminum and three poison followers; since this effects only the homogenization for two discrete cells, the results for core wide parameters is valid. UT TRIGA data is provided in Table 4.17.
Table 4.15, GA-4361 Calculation Model Area    Volume        volume Radius Fraction in.        cm          crn2      cm3 Cell Region U-ZrH 1.7          0.7175.      1.822        10.429    397.34      0.6308 SS Cladding        0.7375        1.873        0.592      22.56      0.0358 Water              0.9032        2.294        5.511    209.98      0.3334 TOTAL                  na          na        16.532    629.88      1.0000 Table 4.16, Selected TRIGA II Nuclear Properties Number of cells                    80                  91 Fuel Temperature                  23&deg;C              200&deg;C 1 eV to 10 MeV 1;a                            0.00660            0.00675 if                            0.00135            0.00135 Flux/watt                      2.46x10 7          2.21x10 7 p'll                            0.9405              0.9481 0 to 1 eV 1"                              0.0873              0.0794 if                              0.0526              0.0472 7 7          1.08x10 Flux/watt                      1.11x10
                          % of fissions                    94.6                94.5 Vave cm/s                      2.73x10 5          2.94x10s Eae eV                        0.0391              0.0455 NOTE 1: Resonance escape probability 13  GA-4361, CalculatedFluxes and Cross Sections for the TRIGA Reactors, G. B. West. August 1963 Page 4-35
CHAPTER 4: REACTOR                                                              1    01/202 Table 4.17, UT TRIGA Data Core Configuration Ref Cold Clean Critical Loading :        64 elements Ref Operational Loading                  90 elements Actulal'Initial Criticality Fuel' elemeht pitch                      0.043536 cm
                    * " Coolantlvolume to cell ratio          .32.86%
                                    ... .Fuel              Elem ents
* Cladding,                              .. SS 304.
Fuel matrix                    .J-ZrH..6 Fuel Mass                                2.5 kg Uranium fraction                          8.5%
Enrichment                                19.5%,
Nuclear Parameters Prompt neutron lifetime,( f)              41 ps Effective delayed neutron.
0 .007 fraction (13)
Prompt negative temperature                        0 coefficient (a)                          1x104 Ak/k C B.      Prompt Negative Temperature Coefficient .
GA Technologies, the designer of the reactor, has developed techniques to calculate the temperature coefficient accurately and therefore predict the transient behavior of the reactor.
This temperature coefficient arises primarily from a change in the disadvantage factor resulting from the heating of the uranium zirconium hydride fuel-moderator elements. The coefficient is prompt because the fuel is intimately mixed with a large portion of the moderator and thus fuel and solid moderator temperatures rise simultaneously. A quantitative calculation of the temperature coefficient requires knowledge of the energy dependent distribution of thermal neutron flux in'the reactor.
The basic physical processes which occur when the fuel-'-moderator elements are heated can be described as follows: the rise in temperature of the hydride increases the probability.that a thermal neutr'on in the fuel element will gain energy from an excited state of an oscillating hydrogen atom in the lattice. As the neutrons gain e.nergy-from the ZrH, their mean free path is increased appreciably. Since the average chord length. in:the fuel element is comparable to a mean free path, the probability of escape from the fuel element before capture.is increased. In the water the neutrons are rapidly thermalized so that the ,capture and escape probabilities are relatively insensitive to the energy with which the neutron enters the water. The*heating of the moderator mixed with the fuel thus causes the spectrurn to h1airden more in the fuel than in the water. As a result, there is a temperature dependent disadvantage factor for the unit cell in the core which decreases the ratio of absorptions in the fuel to total cell absorptions as the fuel element temperature is increased. This brings about a shift in the core neutron balance, giving a loss of reactivity.
Page 4-36
THE UNIVERSITY OF TEXAS TRIGA IIRESEARCH REACTOR                              l'    Tl        01/2012 SAFETY ANALYSIS REPORT, CHAPTER 4                                        &#xf7;    n The temperature coefficient then, depends on spatial variations of the thermal neutron spectrum over distances of the order of a mean free path with large changes of mean free path occurring because of the energy change in a single collision. A quantitative description of these processes requires a knowledge of the differential slow neutron energy transfer cross section in water and zirconium hydride, the energy dependence of the transport cross section of hydrogen as bound in water and zirconium hydride, the energy dependence of the capture and fission cross sections of all relevant materials, and a multigroup transport theory reactor description 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 GA Technologies: GGC-3"4 , GAZE-2 1 s, and GAMBLE-5 16, as well as DTF-IV 17, an Sn multigroup transport code written at Los Alamos. Neutron cross sections for energies above thermal (>1 eV) 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 flux derived by solution of the 81 equations for each discrete reactor region composition. This code and its related cross-section library predict the age of each of the common                  moderating materials to within a few percent of the experimentally determined values and use the resonance integral work of Adler, Hinman, and Nordhein to generate cross sections for resonance materials which are properly averaged over the region spectrum. Thermal cross sections 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 the chemically bound moderator atoms. The bound hydrogen kernels used for hydrogen in the water were generated by the THERMIDOR code 1 8 using thermalization work of Nelkin1 9 . Early thermalization work by McReynolds et a120 on zirconium hydride has been greatly extended at GA Technologies 21, and work by Parks resulted in the SUMMIT t251 code, which was used to generate the kernels for hydrogen as bound in ZrH. These scattering models have been used to predict adequately the water and hydride (temperature dependent) spectra as measured at the GA Technologies linear accelerator as shown in section 4.2.1 (A).
'4 General Atomics Report GA-7157, "Users and Programmer Manual for the GGC-3 Multigroup Cross Section Code," General Dynamics, General Atomic Division (1967) is General Atomics Report GA-3152 "GAZE-2: A One-Dimensional, Multigroup, Neutron Diffusion Theory Code for the IBM-7090," Lenihan, S. R., 'General Dynamics, General Atomic Division (1962) 16 General Atomrics Report GA-818, "GAMBLE A program for the Solution for the Multigroup Neutron-Diffusion Equations in Two Dimensions, with Arbitrary Group Scattering, for the UNIVAC-1108 Computer," 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 Equation with Anisotropic Scatterings, Los Alamos Scientific Laboratory, new Mexico (1965) 18 "THERIMIDOR- A FORTRAN II Code forCalculating the Nelkin Scattering Kernel for Bound Hydrogen (A modification of Robespierre),"Gulf General Atomic, Inc. (unpublished data) Brown, H. D., Jr.
19 "Scattering of Slow Neutrons by Water," Phys. Rev., 11, 741-746, Nelkin, M. S. (1960) 20 "Neutron Thermalization by Chemically-Bound Hydrogen and Carbon," Proc. nd Intl. Conf. Peaceful 2
Used at Energy (A/Conf. 15/F/1540), Geneva, IAEA (1958) 21 General Atomics Report GA-4490 Neutron Interactions in Zirconium Hydride, Whittenmore, W. L.,
General Dynamics, General Atomic Division (1964)
Page 4-37
CHAPTER 4: REACTOR                                                                      01/2012 Qualitatively, the scattering of slow neutrons by zirconium hydride can be described by a model in which the hydrogen atom motion is treated as an isotropic harmonic oscillator with energy transfer quantized -in multiples. of 70..14 eV. More precisely, the SUMMIT model uses a frequency spectrum with two branches, one for the optical modes for energy transfer with the bound proton, and the other for-the acoustical modes for energy transfer with the lattice as a whole. The optical modes are represented as a. broad frequency band centered at 0.14 CV, and whose width is adjusted to fit the cross section data of Woods et al. 1281. The low frequency acoustical modes are assumed to have a Debye spectrum with a cutoff of 0.02 eV and a weight determined by an effective mass of 360.
This structure then allows a neutron to slow down by thetransition in energy units of 0.14 eV as long as its energy is above 0.14 eV. Below 0.14 eV the neutron can still lose energy by the inefficient process of exciting acoustic Debye type modes in which the hydrogen atoms move in phase with the zirconium atoms, which in turn move in phase with one another. These modes therefore, correspond to the motion of a group of atoms whose mass is much greater than that of hydrogen, and indeed even greater than the mass of zirconium. Because of the large effective mass, these modes are very inefficient for thermalizing neutrons, but for neutron energies below 0.14 eV they provide the only mechanism for neutron slowing down within the ZrH. (In a TRIGA core, the water also provides for neutron thermalization below 0.14 eV.) In addition, in the ZrH it is possible for a neutron to gain one or more energy units of -0.14 eV in one or several scatterings, from excited Einstein oscillators. Since the number of excited oscillators present in a ZrH lattice increases with temperature, this process of neutron speeding up is strongly temperature dependent and plays an important role in the behavior of ZrH moderated reactors.
Calculations of the temperature coefficient were done in the following steps:
: a.      Multigroup cross sections were generated by the GGC-3 code for a homogenized unit cell. Separate cross-section sets were generated for each fuel element temperature by use of the temperature dependent hydride kernels and Doppler broadening of the 238 U resonance integral to reflect the proper temperature. Water at room temperature was used for all prompt coefficient calculations.
: b.      A value for. k- was computed for each fuel element temperature by transport cell calculations, using the P1 approximation. Comparisons have shown 54 and S8 results to be nearly identical. Group dependent disadvantage factors defined as (Dg'/ (1gc (region cell) were calculated for each cell region (fuel, clad, and water).
: c.      The thermal group disadvantage factors were used as input for a second GGC-3 calculation where cross sections for a homogenized core were generated which gave the same neutron balance as the thermal group portion of the discrete cell calculation.
Page 4-38
THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR                          *g0 fl0 T I  " 01/2012
                                                                          .---&deg;&deg;O 0
* SAFETY ANALYSIS REPORT, CHAPTER 4
: d.      The cross sections for an equivalent honmogenized core were used in a full reactor calculation to determine the contribution to the temperature coefficient due to the increased leakage of therma! neutrons into <the, reflector with increasing hydride temperature. This calculation requires severa! thermnalgrbups, butiransport effects are no longer of major concern. Thus, reactivity calculati6ns :s a fbiit-tion of fuei element temperature have been done on the entire reactor with the use of diffusion theory codes.
Results from the above calculations indicate that more than 50% of the temperature coefficient for a standard TRIGA core comes from the temperature-dependent disadvantage factor or "cell effect", and ~20% each from .Doppler broadening of the 2 3 8 U resonances and temperature dependent leakage from the core. This produces a temperature coefficient of ~ -0.01%/&deg;C, which is rather constant with temperature.
Because of the prompt negative temperature coefficient a significant amount of reactivity .is needed to overcome temperature and .allow the reactor to operate at the higher power levels in steady-state operation. Fig. A.19 shows. the. relationship of reactor power level and associated reactivity loss to achieve a given power level.
4.5.4    Operating Limits The core-wide operating limits associated with nuclear design are based on spatialdistribution of neutron flux that determines*the local peak power production. Therefore (A)., the peaking factors 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 and capabilities.                                                                      -        .    .
A.      Core Peaking Factors The core.is generally modeled as a&#xfd;rght cylinder. Neutron flux,.varies along the. axis of a cylindrical reactor using periodic Bessel functions. Neutron flux varies: radially in a cylindrical reactor using period sine functions. The product of these two functions. provides a relationship between average core power and the maximum power at a location within the core. Neutron flux and fission rate also varies significantly across the radius of a TRIGA fuel element; the complexities of the system do notlend themselves to reasonable analytic description.
Core Radial Peaking Factor. Classically, the radial hot-channel factor for a cylindrical reactor (using R as the physical radius and Re as the physical radius and the extrapolation distance) is given 2by:
22 Elements of Nuclear Reactor Design, 2nd Edition (1983), J. Weisman, Section 6.3 Page 4-39
CHAPTER 4: REACTOR                                                                          01/2012 1.202 *(R
                                                ./[2.4048"(4      Dl However, TRIGA fuel elements are on the order of a mean free path of thermal neutrons.
Consequently, there is a significant change in thermal neutron flux across a fuel element.
Calculated thermal neutron flux data 23 indicates that the ratio of peak to average neutron flux (peaking factor) for TRIGA 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 actual case is 14% less than power calculated using the assumption.
Core Axial Peaking Factor. The axial distribution of power in the hottest fuel element is sinusoidal, with the peak power a factor of rn/2 times the average, and heat conduction radial only. The axial factor for power produced within a fuel element is given by:
g(z)    1.514"cos(',*z in which &#xa3; = L / 2 and f.., is the extrapolation length in graphite, namely, 0.0275 m. The value used to calculate power in the limiting -location-within the fuel element is therefore 4% higher than power calculated with the actual peaking factor. Actual power produced in the most limiting actual case is 4% less than power calculated using the assumption; therefore calculated temperatures will bound actual temperatures.
Core Local Peaking Factor. The location on the fuel rod producing the most thermal power with thermal power distributed over N fuel rods is therefore:
q  N  D, L P
B. Powerdistribution:within a Fuel Element.                  -
The radial and axial distribution~of the power'within a.fuel.element is given by q.'(r,z)= qfL(r)g(z) in which r is measured from the vertical axis of the fuel element and z is measured along the axis, from the center of the fuel element. The axial peaking factor follows from the previous assumption of the core axial peaking factor, but (since there is a significant flux depression across a TRIGA fuel element) distribution of power produced across the radius of the fuel the radial peaking factor requires a different approach than the previous radial peaking factor for the core. The radial factor within a fuel element is given by:
23 GA-4361, Calculated Fluxes and Cross Sections for TRIGA Reactors (8/14/1963), G.' B. West Page 4-40
THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 4                                            000 000IIETI I  01/2012 er2 f  a+cri+
I + br + dr 2 in which the parameters of the rational polynomial approximation Iare derived from flux-depression calculations for the TRIGA fuel2 4. Values for the coefficients are: a = 0:824461 b =
-0.26315, c = -0.21869, d = -0.01726, and e = +0.04679. The fit is IElustrated anFig. 4.20.
1.3 1.2 1.
L    1.0 0.90 0.80 0.70 0.0 0.20 0.40  0.60 0.80  1.0 1.2  1.4  1.6 1.8  2. 0 Figure 4.22, Radial Variation of Power Within a TRIGA Fuel Rod.
(Data Points from Monte Carlo Calculations [Ahrens 1999a])
C.      Power per rod The Bernath correlation 25 calculates critical heat flux as:
Q AIBO              (Two -TB)
Where the convection heat transfer coefficient for "burnout" condition is, determined by:
hBo      .10~990    (De2+ D) + SLOPE . V With two possible values for the "SLOPE" term:
(1)      IF De< 0.1 ft.,
24  Report KSUNE -Investigation of the Radial Variation of the Fission-Heat Source in a TRIGA Mark III Fuel Element Using MCNP, Ahrens, C., Department of Mechanical and Nuclear Engineering, Kansas State University, Manhattan, Kansas (1999) 25 ANL/RERTE/TM-07-01, Fundamental Approach to TRIGA Steady state Thermal-Hydraulic CHF Analysis Page 4-41
CHAPTER 4: REACTOR                                                                        01/2012 48 SLOPE =        0.
SLOPE =9 9+        -
D.e And the burnout wall temperature term is calculated:
P      V
                                                              +
P54 TwBo = 57- In(P) -          P +15    4 2
The CHF heat flux in is p.c.u./hr-ft , the heat transfer coefficient corresponding to the CHF in 2
p.c.u./hr-ft -C, is the wall temperature at which CHF occurs in &deg;C, Tb is the local bulk coolant temperature in TC, D hydraulic diameter of the coolant passage in feet, D is the diameter of the e
heater surface (heated perimeter divided by n) in feet, P is the pressure in psia, and V is the velocity of the coolant in ft/s. Substituting equivalent terms into the CHF equation results in:
[OIt10890.
[A8 0
(= 80
(  De,      8
                                        +x--*V)"
De+ DiJDe*
57-in(P)                                                                          P_+-i          ,r T
                                                    )~l~)~+5                        4    B Where A is the flow area and WP the wetted perimeter, hydraulic diameter is calculated:
De=- 4-A WP (1) Wetted perimeter:
WP=- r - D *,
2 (2) Flow area:
A=PITCH2        f---r41 4 2          2)
Page 4-42
THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 4 0000 000        ETII        01/2012 TRACE calculates CHFR using the Biasi correlation, however the more accepted CHF correlation for TRIGA reactors is provided by Bernath 26. TRACE calculations completed as described in section 4.6 provides thermal hydraulic parameters used to calculate critical heat flux using the Bernath correlation (and the ratio of the heat flux to the critical heat flux, CHFR). The results of calculations using heat flux and temperature data for 49 0 C water at .6.5 m level is provided in Table 4.18. The minimum CHFR versus power level is provided in Fig. 4.22. As illustrated, the CHFR values agree well and remain much greater than 2 at power levels up to 22.5 kW per unit cell.
Table 4.18, Critical Heat Flux ratio, Bernath Correlation kW      1    2    3      4      5 '    6      7      8      9      10      11    12    13  14    15 1.5    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  96.8 3.0    61.0 53.4  47.2  42.2. 38.5    37.0 35.7: .34.4        35.5  36.6    37.2  39.7 42.9 '46.8  5i.9 4.5    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  35.5 26.9 6.0    35.4 30.9        23.6 21.6        20.8 20.0 19.2          19.7  20.3    20.3  21.3 23.9 24.6    26.7 7.5    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    21.3 9.0    26.0 22.3  19.4  16.9 15.3        14.7 14.1 :13.5        .13.9  14.2    14.C  14.6 16.0 16.2    17.6 10.5    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    14.8 12.0    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    12.6 13.5    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    10.9 15.0    17.6 15.0  12.8  10.9      9.9      9.4    9.0Q 8.6        8.8    9.0    8.6    8.6  9.2  9.1  9.5 16.5    16.4 13.9  11.8  i0.0      9.1      8.6    8.2      7.8    8.0    8.2    7.8    7.8  7.9  8.0  8.3 18.0    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  7.3 19.5    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  6.5 21.0    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  6.1 22.5    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    6.0 26  ANL/RERTE/TM-07-01, op. cit.
Page 4-43
CHAPTER 4: REACTOR                                                                              01/2012 CHAPTER 4: REACTOR
[I 01/2012 Critical Heat Flux Ratio 46.00 41.00 36.00 I
0 31.00 26.00      -                                      I 21.00 16.00 I
11.00 6.00 1.00 4 1.50    4.00 6.50    9.00  11.50  14.00 16.50  19.00  21.50 Unit Cell Power (kW)
Figure 4.23, Critical Heat Flux Ratio (Bernath and Biasi Correlations)
Thermal hydraulic analysis using TRACE (section 4.6) demonstrates that a TRIGA fuel element operating 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, the fuel element that produces the most power (PPEAKROD) is related to the core average power level (PAVE) by:
PFi.AKROI) = PAVh..KPF Parametric variations including peaking factors from 1.3 to 2.0 and the number of fuel elements from 85 to 100 are provided in Table 4.19 and Fig. 4.23. With a peaking factor of 2 and 85 fuel elements, a core at 1913 kW would produce 45 kW in the element producing the highest power.
Table 4.19, Core Power, 45 kW Hot Element Peaking        85            90        100 Factor 1.3        2942          3115      3462 1.4        2732          2893      3214 1.5        2550          2700      3000 1.6        2391          2531        2813 1.7        2250          2382        2647 1.8        2125          2250        2500 1.9        2013          2132        2368 2        1913          2025        2250 Page 4-44
THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 4 0ET I 01/2012 Limiting Core Configuration (Fuel Element Peak Power 45 kW) 3400 3150 290) 2650 0
C-C)  2400 0
U.. 2150  4 -..-  ,-+~
I 190c)4 1.3        1-37    1-44  1.51    1.58  1.65 1.72 179    1.86  1-93 2 Peaking Factor
                                    -        85 ELEMENTS    -g    90ELEMENTS      1--IOELEMENTS Figure 4.24, Core Power, 45 kW Hot Element Based on the calculations, 85 fuel elements with a peaking factor of less than 2.0 provides a large margin to thermal hydraulic limits.
4.6    Core Reactivity As noted in 4.5.1 (A), reactivity worth of material in the core is determined from differential measurements of calibrated control rod worth positions. Verification that the core configuration meets operating limits is similarly determined from the calibrated control rod positions.
As shown in Apoendix 4.1, the rapid fuel temperature response from a pulsed reactivity addition terminates the power increase and causes the reactor to stabilize at a power level corresponding to the fuel temperature consistent with Fig. 4.21. Therefore limits on reactivity are based not on the peak pulse power level, but on the final equilibrium power level associated with the reactivity. A polynomial equation calculating the reactivity deficit based on Fig. 4.24 with an R2 value of 0.99999 is: &#xfd; 6k = -1.75340-" 2 P 4 + 6.06670-10-9"P 3-8.777401 0-6"P2 +8.45380 3"P- 0.072937 An approximation of the power coefficient of reactivity from 100 kW to 1 MW is therefore:
d6k = -7.01360-                2 .P3 + 1.82001-10--P 2-1 .755488.10-6_P +8.45380-1 0-2 dP Page 4-45
CHAPTER 4: REACTOR                                                                                01/2012 Power Coefficient of Reactivity 0.00S    --      .  . . .                            --
OJA00      -    .
1W0    200      3W  400  5WD    f00  700    am0  9W0  1000 Power Level NkW)
Figure 4.25, Power Coefficient of Reactivity Therefore a pulse rod worth limited to 2.8% Ak/k ($4.00) will prevent exceeding steady state power 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 in the 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 pulse occurring 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 inadvertent pulsed reactivity addition leading to prompt critical condition.
There appears to be a significant difference in response in the power level coefficient comparing low power level data to high power level data; the prediction of the power coefficient of reactivity beyond the range of 1000 kW using a simple polynomial fit is not supported. Operating limits on core reactivity are provided in Table 4.20.
Table 4.20, Reactivity Limits
                                                                  %Ak/k                      $
Excess reactivity                                        4.9                    7.00 Shutdown margin[11                                      0.2                    0.182 Moveable experiment worth                                0.7                    1.00 Single experiment worth                                1.75                    2.50 Total experiment worth                                  2.10                    3.00 NOTE [1): most reactive rod fully withdrawn, moveable experiments in the most positive-reactive state Page 4-46
THE UNIVERSITY OF TEXAS TRIGA 11RESEARCH REACTOR
* f        01/2012 SAFETY ANALYSIS REPORT, CHAPTER 4                                                            I Based 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. When significant changes to the core configuration are made, verification that the core meets requirements is accomplished including evaluation that the control rod calibration is valid or re-establishing the control rod worth calibration.
Table 4.21, Limiting Core reactivity Control Rod                      Reference          Current (2011)
Position    Worth      Position    Worth Transient rod                C ring      $3.00        C-1      $3.10 Regulating rod              C ring      $3.71        C-7      $2.82 Shim 1                      D ring      $2.86        D-14      $2.52 Shim 2                      D ring      $2.86        D-6      $3.07 Total Rod Worth                          $12.43                  $11.51 Critical Reactivity                      $5.43                  $5.95 LIMITING                CURRENT Excess Reactivity                        $7.00                  $5.56 Shutdown Margin                          -$1.72                  -$2.85 4.7      Thermal Hydraulic Design This section provides an independent assessment of the expected fuel and cladding thermal conditions, both steady-state and pulse-mode operations, with realiktic modeling of the fuel-cladding gap. Analysis is based on limiting conditions applied to 'a single fuel channel. The correlation of the limiting channel to core average power is sued to validate maximum permissible power level.
Analysis of pulsed-mode behavior is provided in Appendix 4.1, a commonly cited analysis of TRIGA fuel and cladding temperatures associated with pulsing operations. Analysis shows film boiling is not expected, even during~or, after pulsing leading tc*,maximum adiabatic fuel temperatures. The analysis addresses the case of a fuel element at-an-average temperature immediately following a pulse, then estimates cladding temperature and surface heat flux as a function of time after the pulse. The analysis predicts that, if there is no gap resistance between cladding and fuel, film boiling can occur very shortly after a pulse and cladding temperature can reach reaching 470'C. Mechanical stress to the cladding well is below the ultimate tensile strength of the stainless steel at these temperatures. Through comparisons with experimental results, the analysis concludes that an effective gap resistance of 450 Btu hr-1 ft-2 OF- (2550 W m-2 K-) is representative of standard TRIGA fuel and, with that gap resistance, film boiling is not expected.
Page 4-47
CHAPTER 4: REACTOR                                                                      01/2012 Analysis of steady state conditions reveals maximum heat fluxes remain well below the critical heat flux associated with departure from nucleate boiling. The heat transfer model is discussed, followed by the'results.        .
4.7.1  Heat Transfer Model Heat generated in the fuel is conducted through the fuel matrix, transferred by convection across the gap between the fuel matrix and the cladding, conducted through the fuel cladding, and transferred by convection to the cooling water that flows through the core. Fuel centerline temperature can be calculated as the cooling water temperature increased by temperature changes through each physical element from the centerline of the fuel rod to the water coolant.
T,= T11 + ATb,+      AT    + ATg + A T, Where 7 , is the fuel centerline temperature Th is the bulk water temperature AIT,r is the difference in temperature between bulk water and fuel cladding AT,    is the difference in temperature across ATg is the difference in temperature across the gap betweenthe fuel andthe cladding AT,,, is the difference in temperature across the radius of the fuel A standard heat resistance model for this system is:
                          "              h      kc        r, 4 zh
                                                                  , 2 .k f Where q" is the heat flux through the cladding surface h is-the convective heat transfer coefficient associated with the cooling water ro and ri are cladding inner and outer radii kc is the cladding thermal conductivity hg is the gap conductivity kf is the fuel thermal conductivity Thermodynamic values are provided in Table 4.22, with the exception of the convective heat transfer coefficient associated with the cooling water. The gap conductivity of 2840 W m2 K-1 (500 Btu h-' ft -2 F-') is taken from Appendix A. General Atomics reports that fuel conductivity Page 4-48
THE UNIVERSITY OF TEXAS TRIGA 11RESEARCH REACTOR 0
* fl  1  01/2012 SAFETY ANALYSIS REPORT, CHAPTER 4                                                        1 over the range of interest has little temperature dependence. Cladding conductivity across the cladding is temperature dependent, with values quoted at 300 K,400 1(and 600 K.
Table 4.22: Thermodynamic Values Parameter          Symbol      Value              Units Fuel conductivity            kf        18                m- K __
14.9        W m 'K (300 K)
Clad conductivity            kg        16.6        W m-4 K' (400 K) 19.8        W m1 Kz (600 K)
Gap resistance                h9      2840            W m-2 K-1 Clad outer radius            r0    0.018161              m Fuel outer radius              ri    0.018669              m Active fuel length            Lf      0.381                m Axial peaking factor        APF        71/2              N/A The convective heat transfer coefficient is mode dependent and can only be determined in context; TRACE was used to provide heat transfer coefficient data supporting the analysis. The TRAC/RELAP Advanced Computational Engine (TRACE) code is the latest in a series of advanced, best-estimate reactor systems codes developed by the U.S. Nuclear Regulatory Commission for analyzing transient and steady-state neutronic-thermal-hydraulic behavior in light water reactors. It is the product of a long term effort to combine the capabilities of the NRC's four main systems codes    27 (TRAC-P, TRAC-B, RELAP5 and RAMONA) into one modernized computational tool.
The TRACE calculation models a unit cell composed of the area enclosed within a geometry unit defined by fuel pitch. Flow through the unit cell is modeled as a pipe, with model elements represented in Fig. 4.25. The UT TRIGA unit cell is an equilateral triangle, based on hexagonal geometry. Three 30' segments of a fuel element fall within the unit cell, with calculations for heat generation corresponding to 1/2 of the element. For example, calculations assuming 10 kW for the unit cell give indication of thermal response to an element output of 20 kW. The section of the fuel element that contains the fuel matrix (heated length) is modeled separately from the unheated lengths.
The active length of the fuel element was modeled as a TRACE heat source 15 in. (38.1 cm) in length, with the heat exiting through stainless steel cladding. Heat distribution was modeled as sinusoidal variation from a maximum at the center to a minimum modified at the end by extrapolation length of thermal neutrons in graphite. Data was calculated for 15 equally spaced nodes across the span of the simulated fuel element (i.e., 0.0127, 0.0381, 0.0635, 0.0889, 0.114, 0.140, 0.165, 0.191, 0.216, 0.241, 0.267, 0.292, 0.318, 0.343, and 0.368 m).
2'  https://www.nrcsnap.com/snap/-luRins/trace/index.isp 01/2012 Page 4-49
CHAPTER 4: REACTOR                                                                              1 01/2012 Flow entrance and exit has a more complex geometry, and is not modeled explicitly. Special consideration given to the thermal hydraulic characteristics of the fuel end fittings that act as an interface between the flow channel and the grid plates, and the expansion or contraction of flow as it passes into/out of the flow channel. The characteristic losses associated with the entrance and exit includes turbulence effects from sudden expansion and contraction imposed by the grid plates as well as changes in flow direction. These losses are understood in terms of fractional values, or Kfactors. 28 The analytic expression of Kfor expansions/contractions is:
K=F1 l1 L A2 1 The flow path exiting the grid plate and entering the area below the cooling channel undergoes a 450 rotation, followed by another 450 rotation to direct flow along the cooling channel; the associated Kfactors are 0.3429 for each turn. The flow area expands suddenly at the entrance, with a Kfactor approximately 1.0. The loss factor for the cooling channel entrance is therefore 1.68. Similar calculations at the exit of the cooling channel yield a total loss factor of 1.18.
IN")                                1.714-Figure 4.25: Unit Cell Fuel Element Model Hydrostatic pressure is required for TRACE calculations. Pressures at the inlet and outlet to the unit cell were calculated from nominal values of pool level, differential pressure from the confinement system HVAC, and local barometric pressure (Table 4.23). Normal pool level is 8.1 m, with a minimum of 6.5 m. (required by Technical Specifications). Normal pool temperature is about 20&deg;C, with limiting temperature 48&deg;C. Cooling flow enters the lower grid plate at 23.94 in. (0.608 m) above the pool floor and exits the upper grid plate at 51.00 in (1.2954 m). The reactor bay ventilation system operates at a slight vacuum (nominally 0.07 in. H20 or 17.44 Pa) reducing barometric pressure slightly. The average barometric pressure for the Austin area is 28  Handbook of Hydraulics, 5 h Ed. New York: McGraw Hill, King, H.W. and Brater, E. F. (1963) 29  http://www.westerndynamics.com/Download/friclossfittinzs.odf (01/2012)
Page 4-50
THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 4 000 0-0 D 11ETLI                01/2012 reported to range from 28.88 in. to 30.09 in., with an average of the high and low range at 29.485 in., or 99847.67 Pa.
Table 4.23, Hydrostatic Pressures Column      Temp        water        Water            Total water level    height (m)    (&deg;C)        density      Pressure        Hydrostatic (M)                                    (kg/m 3)      (Pa)            Pressure Core top            8.1            6.805          20      998.23      66609.88          166440 Core bottom          8.1            7.492          20      998.23      73338.06          173168 Core top              6.5            5.205          48        988.56      50455.54          150285 Core bottom          6.5            5.892          48        988.56      57118.74          156949 4.7.2    Results The TRACE model was used to calculate temperatures at each of the 15 nodes. The heat flux and the temperature at the inner surface of the cladding were used to calculate the centerline fuel temperature. The temperature data (including TRACE data and centerline fuel temperature in is presented in Tables 4.24-4.27, and graphically (Fig. 4.26). Unit cell and total core flow was calculated (Table 4.28). Finally, recent observations of fuel temperature from installed instrumented fuel elements are compared to calculated fuel centerline temperatures in Table 4.29.
A.        Water Temperature Limiting thermal hydraulic design is based on system response with -the maximum permitted pool water temperature and the "minimum allowed pool water level. TRACE calculations were performed for a range of unit cell power production. Table&4.24 provides the coolant temperatures calculated by TRACE at each node.
Table 4.24, Unit Cell Coolant Temperature (&deg;C) for 490 C 6.5 m Pool Water Unit Ce:.                                                      Node (kW)        1    2      3      4        5      6. 7      8      9    10      11  .. 12      3.3  1.4,  15 1.5    ,    50    50    51      52.    .52      .52    53,. 53    53    53      54    .55      56    57    .57 3.0,        50    51    53      54,      54    ;55    55    55    55  ,.56      57      59      60  .61-    62 4.5      . 50    52. 54      56      56. 56'..57      57      57    58,60            rV62 .64      65    67 6.0        51    52  55      57.,    57      r8    58    59      59    60      62      64  _64      68    70 7.5        51    53    55      58      59      59    60    60    61    61      64    '67        67    72    73 9.0  ... 51    54    56      59.      60      60. 61    62    62    63      66      69      69    74    77 10.5        51    54  57      60      61      62    62    63      64    64      68      71      72    77    79 12.0        52    55  58      61      62      63    64    64      65    66      70      73      74    80    82 13.5        52    55    59      62      63      64    65    65      66    67      71      75      77    82    85 15.0        52    55  59      63      64      65    66    67      67    68      73      77      79    84    87 16.5        52    56  60      64      65      66    67    68    -69    69      74      79      83    86    90 18.0        52    56  60      65*      66      67    68    69      70    71      76      80      85    89    92 Page 4151
CHAPTER 4: REACTOR                                                                                  01/2012 Table 4.24, Unit Cell Coolant Temperature (&deg;C) for 49&deg;C 6.5 m Pool Water Unit Cell                                                  Node (kW)      1    2        3    4      5      6      7    8      9    10  11    12    13    14  15 19.5        53    57      61    66    67      68    69    70    71    72  '77    82    87    91  94 21.0.      53    57    ;61    66      67    68    69    70    71    72    78    83    87    91  95 22.5        53  ,57      61  '66      67    68    69    70    71    72    78    83    87    91  95 B.        Fuel Temperature TRACE calculations provide cladding temperatures directly. Given the cladding temperatures, the standard heat resistance model previously identified in 4.7.1 can be simplified to:
claddnginner +q        h      2 where heat flux and cladding temperatures are calculated in TRACE (Table 4.25 and 4.26). The gap and fuel physical dimensions and thermodynamic properties are constants; based on values in Table 4.22, the terms in parenthesis resolve to 5.39E-4 W m-2 k-1 . About 6% of the coefficient is determined by the gap, and any error associated with gap conductivity is minimized.
Table 4.25a, Outer Clad Temperature (&deg;C) for 49&deg;C and 6.5 m Pool Unit Cell                                                  Node kW        1  2      3      4      5      6      7    8      9    10  11    12      13  14    15 1.5        65  67      70    73    75      76    77    78    77    76  77    76    75  73    72 3.0        76  81      85    89    93      94    96    97    96    95  95    93    91  88    86 4.5        86  91      97 102 106          109    110  113    1I1  110  109    106    103  100    98 6.0      '94 100 -107 -114* 119            121    122  123    i22  122  121    118    114  Il    107 7.5      102  I09 116 123 124              124    125  125    125  124  124    123    122  120  116 9.0      109  117 123 125 125              125    126  126    126  125  125    125    124  123  122 10.5      116  123 "125 '126 126            126    126  127  'i26  126  126    126    125  125  124 12.0      122  125 126 126 127              127    127  127    127  127  127    126    126  125  125 13.5      124  126"  126    i27    127    127    128  110    127  127  127    127    126  126  125 15.0      125  126    127    127    128    128    128  128    128  128 128    127    127  126  126 16.5      126  127    127    128    128    128    128  129    128  128 128    128    127  127  126 18.0      127  127    128    128    129    129    129  129    129  129 128      ".128 128  127  127 19.5      127  128    128    129    129    129    129  129    129  129 129    128    128  127  127 21.0      127  128    128    129    129    129    130  130    129  129. 129    129    128  128  127 22.5      128  128    129    129    130    130    130  130    130  130 i29    129    129  128  127 Page 4-52
THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR                                                              01/2012 SAFETY ANALYSIS REPORT, CHAPTER 4 00 NETL Table 4.25b, Inner Clad Temperature (&deg;C) for 49&deg;C and 6.5 m Pool Unit Cell                                                      Node (kW)          1      2      3      4      5      6      7    :..8        9    .10    11      12  13      14 1.5          66    342  344      347    349    351    352  .353      352      351  351    350 349    348    346 3.0          79    356  360      365    369    370    372    374    .372      371  371    368 366    364    361 4.5          89    368  373      379    384    386    388    391      389      387  386    383 380    377    374 6.0          98    378  385      392    397    400    401    402      402      400  400    396 392    388    384 7.5        107    388  395      402    404    405    405    406      405      405  404    403 401    398    394 9.0        115    397  403      405    407    407    408...408      408-    407  407    406 404    403    401 10.5        122    404  405      408    409    410    410. 411      410      409  40S    408 406    405    404; 12.0        130    406    110    410    411    412    412    413      412      411  411    410 408    407    405 13.5        133    408  410      412    413    414    414    415      414      413  413    411 410    408    407 15.0        135    410  412      413. 415    415    416    417      416      415  415    413 412    410    408 16.5        136    411  413      415    417    417    418    419      418      417  416    415  413    411    410 18.0        138    413  415      417    418    419    420    421      420      419  418    416 415    413    411 19.5        139    414  416      418    420    421    422    422      422      421  420    418 416    414    412 21.0        140    416  418      420    422    423    423    424      423      422  422    420  418    415    413 22.5        142    417  419      421    423    424    425    426      425      424  423    421 '419    417    415 Tabie 4.26a, Heat Flux (Nodes l-18) 49&deg;C 6.5 Pool, Unit Cell                                                Node (kW)          1          2          3            4            5    ..      6          7        8 1.5    -2.72E4      -3.06E4    -3.40E4    -3.74E4      -4.08E4      -4.23E4    -4.38E4  -4.53E4 3.0    -5.44E4      -6.12E4    -6.80E4    -7.48E4      .-8.16E4    -8.46E4      -8.76E4  -9.06E4 4.5    -8.16E4      -9.18E4    -1.02E5    -1.12E5      -1.22E5    -1.27E5      -1.31E5  -1.36E5 6.0    -1.09E5      -1.22E5    -i.36E5    -1.50E5      -1.63E5      -1.69E5    -1.75E5  -1.81E5 7.5    -1.36E5      -1.53E5    -1.70E5    -1.87E5      -2.04E5,    -2.11E5    -2.19E5  -2.27E5 9.0    -1.63E5      -1.84E5    -2.04E5    -2.24E5      -2.45E5      -2.54E5    -.2.63E5  -2.72E5 10.5    -1.90E5      -2.14E5    -2.38E5    -2.62E5      -2.86E5      -2.96E5    -3.06E5  -3.17E5 12.0    -2.18E5      -2,45E5    -2.72E5    -2.99E5      -3.26E5      -3,38E5  _-:3.50E5  -3.63E5 13.5    -2.45E5      -2,75E5    -3.06E5    -3.37E5      -3,67E5      -3.81E5    -3.94E5  -4.08E5 15.0    -2.72E5      -3.06E5    -3.40E5    -3.74E5      -4.08E5      -4.23E5  .- 4.38E5  -4.53E5 16.5    -2.99E5      -3.37E5    -3.74E5    -4.11E5      -4.49E5      -4.65E5    -4.82E5  &#xfd;4.99E5 18.0    -3.26E5      -3.67E5    -4.08E5    -4.49E5      -4.89E5 . -5.07E5        -5.25E5  -5.44E5 19.5    -3.54E5      -3.98E5    -4.42E5    -4.86E5      -5.30E5      -5.50E5    -5.69E5  -5.89E5 21.0    -3.81E5      -4.28E5    -4.76E5    -5.23E5      -5.71E5      -5.92E5    -6.13E5  -6.34E5 22.5    -4.08E5:    -4.59E5    -5.10E5    -5.61E5      -6.12E5      -6.34E5    -6.57E5  -6.80E5 Page 4-53
CHAPTER 4: REACTOR I      01/2012 Table 4.26b, Heat Flux (Nodes 8-15) 49&deg;C 6.5 Pool, Unit Cell                                        Node (kW)I              9        10          11          12          13        14        15 1.5        -4.38E4    -4.23E4    -4.08E4    -3.74E4    -3.40E4  -3.06E4  -2.72E4 3.0        -8.76E4    -8.46E4    -8.16E4    -7.48E4    -6.80E4  -6.12E4  -5.44E4 4.5        -1.31E5    -1.27E5    -1.22E5    -1.12E5    -1.02E5  -9.18E4  -8.16E4 6.0        -1.75E5    -1.69E5    -1.63E5    -1.50E5    -1.36E5  -1.22E5  -1.09E5 7.5        -2.19E5    -2.11E5    -2.04E5    -1.87E5    -1.70E5  -1.53E5  -1.36E5 9.0        -2.63E5    -2.54E5    -2.45E5    -2.24E5    -2.04E5  -1.84E5  -1.63E5 10.5        -3.06E5    -2.96E5    -2.86E5    -2.62E5    -2.38E5  -2.14E5  -1.90E5 12.0        -3.50E5    -3.38E5    -3.26E5    -2.99E5    -2.72E5  -2.45E5  -2.18E5 13.5        -3.94E5    -3.81E5    -3.67E5    -3.37E5    -3.06E5  -2.75E5  -2.45E5 15.0        -4.38E5    -4.23E5    -4.08E5    -3.74E5    -3.40E5  -3.06E5  -2.72E5 16.5        -4.82E5    -4.65E5    -4.49E5    -4.11E5    -3.74E5  -3.37E5  -2.99E5 18.0        -5.25E5    -5.07E5    -4.89E5    -4.49E5    -4.08E5  -3.67E5  -3.26E5 19.5        -5.69E5    -5.50E5    -5.30E5    -4.86E5    -4.42E5  -3.98E5  -3.54E5 21.0        -6.13E5    -5.92E5    -5.71E5    -5.23E5    -4.76E5  -4.28E5  -3.81E5 22.5        -6.57E5    -6.34E5    -6.12E5    -5.61E5    -5.10E5  -4.59E5  -4.08E5 Calculation of maximum fuel temperature was performed using the standard heat resistance model (Table 4.27) modified to use the TRACE data as described above.
Table 4.27, Fuel Centerline Temperatures (&deg;C)
Unit Cell                                                  Node (kW)      1        2      3      4      5      6      7      8    9      10  11    12  13      14    15 1.5    90      95    101  107    112    115    117 119    117    115 114    110 106    102    97 3.0    126    137      147  158    167    172    176 180    176    173 170    161 153    145  136 4.5    161    175      190  205    218    225    231 237    231    226 221    209 197    185  172 6.0    194    212      232 251      268 276        283 289 283        276 270    255 239    222  207 7.5    226    249      272 294      311 318        325 333 325        318 311    295 278    260  241 9.0    258    286      310 330      350 358        367 375 367        358 350    330 311    292  272 10.5    290    319      343 366      388 398        407 417 407        397 388    366 343    321  298 12.0    322    349      375 400      425 437        448 460 448        436 425    400 375    350  325 13.5    349    378      407  436    464    477    489    502  489    476 463    436 407    378  350 15.0    375    407      438 470      502    515    529 S43    529    515 501    470 438    407  375 16.5    400    436      470 504      540    554    570 586    570    554 539    504 470    436  400 18.0    425    464      502 540      577    593    610 627    610    593 576    539 501    463  425 19.5    451    492      533 574      615    633    650 669    650    633 614    574 533    492  451 21.0    476    520      565 608      652    672    691 710    691    672 652    608 564    520  476 22.5    502    549      596 643      690    711    732 753    732    710 690    643 596    549  501 Page 4-54
THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 4 Mt I  01/2012 C.      Temperature Profiles and Flow Rates TRACE calculation provides thermal response for a unit cell. Temperature calculations are based on heat flux (heat per area) and are therefore valid for both the unit cell and the fuel element. However, Fig. 4.25 shows a unit cell to be Y2 of a fuel rod so that the total power generated in a fuel element is twice the unit cell value.
Fuel, cladding, and coolant temperatures based on TRACE data in Tables 4.24, 4.25, and 4.27 are provided in Fig. 4.26 for two unit cell power levels, 10.5 and 22.5 kW. Flow rate versus power for a single fuel element is provided in Fig. 4.25.
Total core flow is the sum of the flow rates of individual fuel rods operating at specific power levels. The power level an element generates is determined by the peaking factors associated the position in the core. Instrumented fuel elements (modified to accept thermocouples) have slightly less fuel mass than other standard fuel elements. Fuel followers have a smaller radius than standard fuel elements. Therefore the power production in thermocouple elements and fuel followers is less, approximately by the ratio of the mass of the element to the mass of a standard fuel element. Table 4.28 provides the data and calculation of total core flow based on a 116 element core operating at 1100 kW. Similar calculations were performed for a 120, 100 and 85 element core over a range of power level with results in Fig. 4.27.
10.5 kW Unit Cell Axial Temperature Profile 675 625                                                                                                              I 575  --
525 475        -7 M
45......                                  ...............................
4-  425                            -                                                              -------
375 325  .-..              .7 275 1                    3                    5                                                    11          13    15 Node
                                - -    Fuel Temp              ---  Inner Clad Temp                -  Outer Clad Temp      - -Coolant Temp Figure 4.26a, Unit Cell Temperature Distribution (10.5 kW)
Page 4-55
CHAPTER 4: REACTOR CHAPTER 4: REACTOR I      01/2012 01/2012 22.5 kW Unit Cell Axial Temperature Profile 975 875 775 675 575 E
9 475 375 275 1          3          5          7          9          11            13          15
                          -  Fuel Temp    .Inner  Cad Temp      -CouterClad    Temp    -    -CoacantTemnp Node Figure 4.26, Unit Cell Temperature Distribution Flow rate versus Unit Cell Power 0.13 0-12 0.11 0.10 009 0.08 0.07 0.36 0.05 0.0]4  -                    i          ; .--          --.    .    .. . . .. .
0~3              yl 1.02541EO5x5 -4.390E-O4xz + I.412BE-O3x+3.0123E402 0.02                                      .992S-01"-
0.00 0    2    4    6    8    10  12    14    16    18    20      22    24 Unit Cd power (W, Figure 4.27, Single Rod Flow Cooling Flow Rate versus Power Level 49&deg;C 6.5 Pool Page 4-56
THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 4                                      0  0 I    01/2012 Table 4.28, Coolant Flow for 1100 kW Operation no.      peaking              Power (kW)                        Flow Rate (kg/s ring    elements      factor    Unit Cell    Element      RING      Unit Cell Element        RING B (6)        4          1.57        7.44        14.89      59.55      0.07        0.15      0.58 B IFE        2          1.57        7.31        14.61      29.23        0.07        0.14      0.29 C (12)      10          1.46        6.92        13.84      138.45      0.07        0.14      1.41 C FFCR        2          1.46        5.78        11.56      23.11      0.07        0.13      0.26 D (18)        16        1.29        6.12        12.23      195.72        0.07        0.14      2.16 D FFCR        2          1.29        5.11        10.21      20.42        0.06        0.13      0.25 E (24)      24          1.07        5.07        10.15      243.52        0.06        0.13      3.02 F (30)      28          0.81        3.84        7.68      215.07        0.06        0.11      3.17 G (30)      28          0.66        3.13        6.26      175.24        0.05        0.10      2.94 TOTAL:      116                                TOTAL:      1100                  TOTAL:      14.08 D.      Comparison to Operational data During calendar year 2011 the UT TRIGA core consisted of 109 standard fuel elements, 2 instrumented fuel elements and 3 fuel followers. Operational data was collected (Table 4.29) to compare calculated fuel temperature values for specific operations with observed indications. Each of the selected values follows an operating interval that approaches steady state fuel temperatures, except for *the 10 kW values (there was a series of 10 kW operations for less than 1 hour during 2011). The power produced by the individual fuel element is calculated as total core power divided by the number of fuel elements (114) and multiplied by the nominal B ring peaking factor (rt/2).
Instrumented fuel elements contain three thermocouples, one at the axial ',midplane with the remaining thermocouples offset 1 inch above and below. Only one thermocouple in each IFE is in use. Fig. 4.24 shows approximately'20 'C difference between the center of the element and positions approximately 1 in. from the center at 10.5 kW unit 6ell power. The, core power is not expected to be homogenous in the B ring; consequently the peaking factors a*re not expected to be uniform for all B ring elements. The position of an IFE and the position of an individual thermocouple within IFE may affect temperature indication; the agreement between FT1 and FT2 measuring channels is therefore considered remarkably good.
Table 4.29, Observed Fuei Temperatures Date            Power (kW)          Observed Temperatures (&deg;C)
Core      B Ring IFE    FT1      FT2        Pool 10/6/2011      10          0.13        26        28        21.1 12/21/2011      100        1.31        86        97        22.8 12/20/2011      500          6.54        240      261          23 12/16/2011      950        12.44        340      359        21.9 Page 4-57
CHAPTER 4: REACTOR I  01/2012 TRACE calculations were performed at power levels corresponding to 100, 500 and 950 kW with pool water temperatures corresponding to the values recorded in Table 4.29. The maximum fuel temperatures from the TRACE calculation are provided in Table 4.30 along with the observed values for comparison.
Table 4.30, Fuel Temperature Comparison Power (kW)                  Fuel Temperatures    (&deg;C)
Core          B Ring IFE    TRACE        FT1          FT2
                              .100              1.3          95          86            97 500            6.5        289        240          261 950            12.4        446        340          359 Fuel Element Temperature Versus Power
                                          -'-CALCJLATED -o-FT I VALUES -FT  2 VALUES 500 450 400 350 3O0 E
a E
250 200 t
U.- 150 100        ...... 4-41 50 0
0    1        2      3      4    5    6      7    8      9      10  11 12  13 Fuel Element Power (kVNI)
Figure 4.28, Comparison of Calculated and Observed Fuel Temperatures The information (provided graphically in Fig. 4.28) shows the temperature from TRACE calculations to be reasonably close and consistently higher than observed values, with the deviation increasing as power level increases. Therefore, modeling bounds actual conditions.
Page 4-58
THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 9                                ..
00o0 0
I            01/2012 9.0      AUXILIARY SYSTEMS 9.1      Confinement System The design of a structure to contain the TRIGA reactor depends on the protection requirements for the fuel elements and the control of exposures to radioactive materials. Fuel elements and other special nuclear materials are protected by physical confinement and surveillance.
The floor of the reactor bay is approximately                              The lower walls of the reactor bay are cast in place concrete. Above grade, the walls are reinforced precast concrete tilt panels, approximately                                with integral columns and embedded reinforcing steel. The wall panels were then set in place vertically using a crane with space left in between each panel for a structural column and temporarily braced. Next the column forms were placed around reinforcing steel extending from the edges of the panels which was interlaced with additional steel reinforcing internal 'to the columns. Concrete was then poured into these forms resulting in a finished wall system with columns that resemble a poured in place design rather than the typical tilt panel welded design.
The roof is sealed using standard tar and gravel techniques. All penetrations in the reactor bay confinement envelope are on the south side, interfacint with the reactor wing offices, machine room spaces, equipment staging area, and confinement (and auxiliary purge) ventilation system..'.
9.2      HVAC (Normal Operations)
Building environment controls use air handling units for ventilation and comfort with cold and hot water coils for temperature and humidity control. There are two separate HVAC systems with three air handling units, located on the fourth level of the reactor bay wing adjacent to the reactor bay. One unit contains both cold and hot water coils in a single duct.system, dedicated to the reactor bay. This system supports confinement functions. The other two units are the cold- and hot-deck components of a double duct system that conditions air in all building zones other than: the reactor bay. A fume/sorting hood is installed in the reactor bay, using a separate exhaust fan and isolation damper that disc"h'arges into a separate roof stack.
Water temperatures of the heating and cooling coils in the air handling units are controlled by set of on-site and off-site systems. The heating system is an ori-site boiler unit with a design capacity set by local building (HVAC) requirements. The cooling system is a PRC chilled water treatment plant with design &#xfd;capacity Set by overall research campus requirements; with thermostats controlling zone or room temperatures. A local instrument air system provides control air for HVAC systems. Controls and air balancing of the two air handling systems provide user comfort and pressure differentials between the reactor bay (confinement) and adjacent zones, and between the adjacent zones and the academic wing of the building.
Page 9-1
CHAPTER 9, AUXILIARY SYSTEMS                                                          1    01/2012 The ventilation system is designed to maintain a series of negative pressure gradients with respect to the building exterior and other building areas, with the reactor bay (confinement) at the lowest pressure. Confinement functions of ventilation control the buildup of radioactive materials generated as a byproduct of reactor operations, and isolate the reactor bay in the event that an abnormal 'release is'detected in the reactor areas. Confinement and isolation is achieved by air control;dampers and leakage prevention material at doors and other room penetration points.
A conceptual diagram of the system is provided in Fig. 9.1. Manual operation controls for both main and purge air systems are in the reactor control room.
wow Figure 9.1, Conceptual Diagram of the Reactor Bay HVAC System An exhaust stack on the roof combines the ventilation exhausts from both the main and the purge systems. As illustratedJin Fig. 9.1, the auxiliary purge system discharge is within the HVAC exhaust stack.. The auxiliarypurge exhaust is a 6 in. (15.24 cm) internal ID and 8.63 in. (21.92 cm) OD. The HVAC exhaust has an 18 in. (45.72 cm).
9.2.1  Design basis Confinement system ventilation has three modes of operation, reactor run mode, quiescent mode, and confinement isolation. The design goal for HVAC system in the reactor run mode is to control the reactor bay, adjacent zones and academic wing of the building at a negative pressure difference relative to ambient atmospheric pressure during 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 during routine operations is discharged through the stack and does not build up in the reactor bay. Release of airborne radioactivity consists mostly of activated 4 1Ar from routine operation. The design goal of the confinement Page 9-2
THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR                        00        B0if  01/2012 SAFETY ANALYSIS REPORT, CHAPTER 9 system ventilation during quiescent mode is to minim; e energy utilization during periods when the reactor is not operated while maintaining pressure in the reactor bay 0.06 in. below atmospheric pressure. The reactor-room confinement is 4esigned to control the exposure of operation personnel and the public from radioact.ive material or its release caused by reactor operation. During potential accident conditions, sensors initiate confinement system isolation when high levels of radioactivity are detected in reactor bay air, e.g. If a fuel element. failure releases fission products or if an experiment with sufficient inventory of radioactive material fails. The confinement isolation secures fans and dampers in the. confinement HVAC, fume/sorting hood, and auxiliary purge 1system. Provisions are made to allow subsequent operation of the auxiliary purge system with the remaining HVAC confinement in isolation.
Release criterion is based on Title 10 Chapter 20 of the U.S. Code of Federal Regulations.
9.2.2  System description During operating modes supply fans draw air from either the return fan or the environment into a conditioning unit that subcools the air:,to control humidity then heats the air for habitability/comfort. Air filtration is the typical design for normal. HVAC operation with fiberglass roughing filters only. The ccnfinement systemn uses heating and cooling in a single unit, the remainder of the building HVAC system has air conditioning split into separate hot and cold decks.
COMM-~
SLPPLY AIR
              -~      ~~
FILTER  a eJ~EA1              ~      1.Ai ISCMfIC Figire 9.2A, Ma'n Reactor Bay HVAC System Page 9-3
CHAPTER 9, AUXILIARY SYSTEMS                                                                    01/2012 Table 9.1, Typical Confinement Vent & Purge Parameters Duct Velocity                            Exit Velocity Aux Purge                  3900  fpm      20    m/s            35.23    m/s Confinement Vent          1800  fpm        9    m/s            26.87    m/s Flow Rate Aux Purge                  1100  cfm    0.52    m3/s Confinement Vent          7200  cfm    3.40    m3/s PAM AIRWO AMIUET *l.T*
* uq.Y      S3 m      IA"M      ISMATM (P. LWIII
                                      ~f.            HVAC OPERAT ION /*ODES
                                                                    -W 6HI. FNG CM all R.P.                  MATISO-D'A= AM    FZCE=
                                                                  *amAM pl&#xfd;cadvs FMOF Fiue92,Mi- RatrByHA              fify SyteCnto Contro    fro"W=M .U~I Figure 9.21B, Main Reactor Bay HVAC Control System Control 9.2.3    Operational analysis and safety function Speed of the confinement system supply fan is regulated to produce 0.06 in. water vacuum in the reactor bay by differential pressure control between the reactor bay and a representative ambient external building measurement point. Additional measurement points in ventilation zones adjacent to the reactor bay are used to maintain differential pressure between the reactor bay and adjacent access areas. Supply air is distributed through a rectangular duct near the ceiling and then to distributed ducts and vents running down the wall and ending near the floor), enhancing mixing and preventing stratification. Air is discharged from the bay through 4 return grills, two parallel ducts to grills near the floor, and two grills near the ceiling. In the reactor run mode the confinement system exhaust fan is controlled to maintain stack velocity designed to exceed the minimum air change specification. Control dampers are located at the supply fan inlet (fresh air intake) and the exhaust fan outlet (discharge to stack), and in a line between the supply and return fans. Confinement system ventilation discharge is through a Page 9-4
iS THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR                            so    e  REI't        01/2012 SAFETY ANALYSIS REPORT, CHAPTER 9 I
stack on the reactor building roof. Schematics of the ventilation system for the reactor bay area and a logic diagram of the ventilation control system sensors and controls are provided in Fig. 9.2A and B.
Table 9.2, Reactor Venti'Eation System Modes MODE SYTMCMOETREACTOR                                  RUN          QUIESCENT          ISOLATION control damper                  CLOSED                    OPEN              CLOSED supply & exhaust                  OPEN                MINIMUM            MINIMUM control dampers Confinement HVAC                                Controlled for Stack Supply Fan                      velocity            Constant Speed            OFF Exhaust fan                Controlled for bay      Controlled for bay        OFF dp                      dp supply    &    exhaust            OPEN                  CLOSED(i            CLOSED control dampers                                                              OPEN[21 Auxiliary Purge System                                                                                              OFF Exhaust fan                        ON                    OFF[1]                    121 OFF or ON supply    &  exhaust    OPEN or CLOSED' 3 1      -  CLOSED"'            CLOSED Fume/Sorting Hood    control dampers Exhaust fan                  ON or OFF[31'              OFF111              OFF NOTE [1]: Mode is set manually NOTE [2]: Provisionshave been made to permit operationof auxiliarypurge system in conferment isolation NOTE [3]: Fume hood is operated manually, as required,and not correlatedto reactoroperation When the reactor is operating (reactor run mode) the system is operated to generate a rate of air exchange exceeding 2 air volumes (4120 M3) per hour. maintain a controlled stack velocity, and regulate negative pressure in the reactor bay. In the reactor run mode, the confinement HVAC supply fan is controlled to maintain the reactor bay at nominal minimum 0.06 in. water.
In the quiescent mode, the confinement ventilation system is balanced for recirculation flow with a small amount of effluent. When the reactor is not operating (quiescent mode), the ventilation system is operated to minimize requirements for conditioning incoming air, in a recirculation mode with a minimal exhaust flow rate and fresh air intake as required to maintain a negative pressure in the reactor bay with respect to adjoining spaces.
In the confinement isolation mode the confinement HVAC and the reactor bay fume/sorting hood are secured; the auxiliary purge system is secured when isolation occurs, but may be manually configured to operate. In the event that airborne radioactive material exceeding a trip set point is detected, the system is designed to establish a shutdown and isolated Page 9-5
CHAPTER 9, AUXIUARY SYSTEMS                                                                        01/2012 condition. Separate controls allow the confinement HVAC and the reactor bay fume/sorting hood to be isolated while the auxiliary purge system can be operated.
Atmospheric dispersion' using a'st4ck, model reqtires:stack discharge 60 (18.23 m) feet above the ground, and at least :2 and 1A times the height of adjacent structures. The nearest structure ismapproximately 80 meter's from the" reactor bay'. Ground elevation in the area is 794 feet, with roof elevation at the stack 843 feet, a distance of 49 feet (14.94 m) above grade. The exhaust stack extends:14 feet (4.24 &#xfd;meters) above the roof level so that the stack discharge is 63 feet (19.202inm). The &#xfd;effective release point above the exhaust'stack can be calculated from the Bryan - Davidson *equation:              -
1.4 (VS.
Where:
Ah is the height of plume rise above release point (m)i 2                        2 D is the diameter of stack (m), confinement vent 0.40122m , auxiliary purge 0.152 m
        /7is the mean wind speed at stack heght (m/s)
V, is the effluent vertical eff!ux velocity (m/s), confinement vent 26.87 m/s, purge 35.23 m/s The effective stack height for the reactor HVAC confinement vent system (in units of meters) is therefore 40.19/{wind velocityl. rn above the stack, and the effective stack height for the auxiliary purge system is 22.25/{wind velocityl above the top of the:stack at 63 feet (19.202 m).
Mixing of the two effluent streams occurs at the exit of the stack.
Pneumatically operated isolation dampers in the confinement system ventilation are located at the supply fan outlet (supply-to 'he reactor bay) and the exhaust fan inlets (return from the reactor bay) near the reactor bay wall penetrations as indicated in Fig. 9.1, as* weli as the fume/soring hood in the reactor bay auxiliary purge system. Controls close the dampers and securethe fans in response to manual or automatic signal initiated by high airborne particulate radioactivity. Loss of instrument air or loss of control power will cause the'dampers to isolate the reactor bay.
9.2.4    Instruments and controls As indicated, the HAVC control, system is controlled' by a, set of temperature, flow, and differential pressure sensors that develop control signals. The signals are used in variable frequency controllers that regulate fan speed to maintain pressure and temperature.
Control room switches establish the operating mode of the confinement ventilation system.
The auxiliary purge system is controlled from the same panel.
Page 9-6
THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR                          OF,          01/2012 SAFETY ANALYSIS REPORT, CHAPTER 9 e Mt Figure 9.3, Confinement System Ventilation Controls Confinement HVAC mode is controlled by a toggle switch labeled "Reactor Off/ Reactor On."
Reactor Off establishes quiescent mode described above. Reactor On establishes the reactor run mode described above.
Alarm indicators on tie control panel provide indication that the diffarential pressures are normal or abnormal. Flow and differential pressure indicators inside the panel provide indication of the zone static pressure, and confinement system and auxiliary purge system velocities.
A continuous air particulate detector located in the reactor bay provides a control signal to initiate confinement isolation when the count rate exceeds a preset level. Indicators at the reactor control console provide alarm level information. A count rate associated with 2,000 pCi/ml detects particulate activity at occupational levels of 10CFR20. The alarm setpoint exceeds 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 the reference concentrations within two hours.
Page 9-7
CHAPTER 9, AUXILIARY SYSTEMS,                                                                                                                  01/2012 9.2.5    Technical Specifications, bases, testing and surveillances Either the confinement ventilation system or the auxiliary purge system is required to be operating when the reactor is operating to control the buildup of gaseous radioactive material in the reactor bay. Ifthe, confinement, ventilation system is operating, instrumentation to initiate confinement .isolation on high airborne contamination levels will be operable. The confinement system.will be: checked periodically to assure proper function. The particulate monitor will be calibrated periodically.
9.3. Auxiliary Purge System A separate, low volume air purge system is designed to exhaust' air that may contain radionuclide products from strategic locations in the reactor bay.'
9.3.1    Design basis The purge system collects and exhausts air from potential sources of neutron activation such as beam tubes, sample transfer systems, rotary specimen rack, and material evolving from the surface of the pool. The purge system filters air in the system. through a rough prefilters followed by a high efficiency particulate: filter. .;Design provisions allow for the addition of charcoal filters if experiment conditions or.other' situations should require the additional protection..          --  .I 9.3.2  System description:
Mai AIR T                "          * ' .        .      .'  ..      .*      ' "LP;i avr    *u        ,      z ARII  PLO                                                                                        WI-      I I  93---1        FILTER                        Li-
                                                                                                            -  q                          ....
ARW RPUE        2 02,.97;1 KEA FILTER E)O4AL5T                  F74"4 Eg7.3J 67ACX                                      3 FUT1JM OWJAL FILTER                                      CA        I-fO
                                        ,      .                                                            MIN.                  M*N
                                                                                                          *P.                          ILzmi I~.ATrI                    Tap~i VALVE                  2    4
                                                                                                                    ---0    -*V
                                                                                                                              .U**I D--AIXIM 'AIR ISO"
* N'    I ,              ..
                                          ...      WI.                      SIM e        "Z L                          ON REACTORBSAYH Figure 9.4A, Purge Air System                    _              Figure 9.4B, Purge Air Controls                    I Page 9-8
THE UNIVERSITY OF TEXAS TRIGA 11RESEARCH REACTOR                  0o0o 00                  01/2012 00 SAFETY ANALYSIS REPORT, CHAPTER 9                                  0 9.3.3  Operational Analysis and Safety Function The primary nuclide of interest is argon-41. Fig. 9.4A and .9.4B are schematics of the auxiliary purge system and its control logic.. Sample ports in the turbulent flow stream of the purge system exhaust .provide for measurement of exhaust activities. The isolationi damper in the purge system is actuated manually, using the.fan control switch. Autornatic iso!atior, o1 the system is generated by the same particulate radiation mcnitcr as is used by.the HVAC confinement ventilation system.
Purge flow is nominally adjusted for continuous operations with approximately 525 cfm from the pool and a similar dilution flow rate from the reactor bay environment. The dilution flow controls effluent humidity from the reactor pool area to limit possible degradation of the purge system HEPA filters. A purge flow of approximately 4 cfm is drawn from the beam port interior when a beam port is used. The beam guide prevents closure of the outer shutter door, and beam port three is normally purged. The rotary specimen rack is purged prior to loading or unloading for about 10 minutes to control personnel exposure and also to remove hydrogen that may evolve from the polyethylene capsules during irradiation. -
The auxiliary purge system may be operated with the confirement HVAC system secured. Since the confinement HVAC operates continucusly except during isolation, confinement HVAC can be secured using thei HVAC Contro!, toggle switch (inside the HVAC control panel, described previously). Since the auxiliary purge system is equipped with HEPA filters and :has the capability for using charcoal filters, operation of the auxiliary purge system could reduce elevated airborne radionuclide contamination in the reactor bay and contain a large fraction of the radionuclides in filtration.. Qperation in this mode requires that the confinement HVAC be secured to prevent unfiltered releases, and then bypassing the confinement isolation trip signal.
9.3.4  Instruments and controls The auxiliary purge system is :controlled from the same panel as the confinement ventilation system. Toggle switches on the controlroormiconfinement HVAC cdhtr6lp3nbl. open dampers to allow the pool surface purge flow, 'and-independently flow from a manual valve manifold accessible on the ground level of the rector bay. The manual valve manifold controls purge flow from the experiment facilities. A separate manually operated: valve' in the same area controls the amount of dilution flow to the purge system.
A flow gage indicates purge stack velocity at the panel. The exhaust point is concentric to the center of the HVAC confinement ventilation exhaust stack.
The auxiliary purge system is monitored by a gaseous effluent radiation detector. The effluent monitor has an alarm setpoint based on ten times the occupational limit or a reference concentration at the ground.
Page 9-9
CHAPTER 9, AUXILIARY SYSTEMS 9.3.5    Technical Specifications, bases, testing and surveillances Ifthe auxiliary purge system is operating, a gaseous effluent monitor will be operating. The auxiliary purge system will have a high efficiency particulate filter. Auxiliary air purge system valve alignment will be checked periodically. The gaseous effluent monitor will be calibrated periodically.
9.4      Fuel storage and handling Special provisions are necessary for the storage of fuel elements that are not in the core assembly. 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 are specially designed racks inside the reactor pool and outside the reactor shield.
Irradiated fuel is manipulated remotely, using a standard TRIGA fuel tool. Irradiated fuel is transferred out of the pool using a transfer cask modeled on the BMI cask TRIGA basket. There are two different loading templates for use with the transfer cask, permitting loading operation either for a single TRIGA fuel element, or to up to three elements. A 5-ton overhead crane is used to move the fuel transfer cask.
9.4.1    Design basis Stored fuel elements are required to have an effective multiplication factor of less than 0.8 for all conditions of moderation. Fuel handling systems and equipment are designed to allow remote operation of irradiated fuel, thus minimizing personnel exposure.
9.4.2    System description..
Space inside the reactor pool is adequate for a large number of fuel racks. The racks are aluminum, suspended from the pool edge by connecting rods.
To facilitate extra storage, 2 racks may be attached to the same connecting rods        by'locating one rack at a different vertical level and offsetting the horizo0tal position.. slightly.
,Outside the react-or pool, rack design is intended to fit in,special storage wells (Fig. 9.4).
Water may be added for shielding or cooling.
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THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 9
                                                                    '.ETII          01/2012 1
Most routine fuel storage is intended to be inside the reactor pool. Outside the reactor pool, supplemental fuel storage is planned for temporary storage of elements transferred to or from the facility, for isolation of fuel elements with clad damage, emergency storage of elements from the reactor pool and core assembly and routine storage o" other radioactive materials.
Temporary storage for some reactor components or experiments may also use the fuel storage Page 9-11
CHAPTER 9, AUXILIARY SYSTEMS                                                                01/2012 racks in the reactor pool. Other locations not in the pool will also provide storage for radioactive non-fuel materials.
A fuel transfer cask modeled after the BMI cask TRIGA basket is used to transfer irradiated fuel.
A standard TRIGA fuel handling tool is used to remotely grapple irradiated fuel elements.
A 5-ton crane is used in conjunction with the fuel handling tool and the transfer cask to allow remote handling of irradiated fuel.
9.4.3    Operational analysis and safety function Bench mark experiments conducted by TRIGA International indicate minimum mass for criticality requires 64 fuel elements in a favorable geometry.
Pool storage racks do not have the capacity or the geometry to support criticality. Spent fuel storage has a higher fuel density in storage, but does not have the capacity to hold 64 fuel elements, and does not have favorable geometry.
The fuel handling tool has been used successfully at the UT TRIGA reactor, including the original reactor on the main campus as well as the current installation. This design is widely used by TRIGA reactors, with good performance history although the first generation tool occasionally released 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 exists in a bottom loading cask. The cask does not provide adequate shielding for close-in work, and all handling is conducted remotely.
The crane exceeds load requirements for spent fuel handling by a large margin. There is little potential for failure.
9.4.4    Instruments and controls New fuel storage is in a locked room on the middle level of the reactor bay. A criticality monitor is installed, with neutron and gamma channels. The system has a local indicator directly outside the storage room, and a remote readout in the control room.
9.4.5    Technical Specifications, bases, testing and surveillances Fuel elements are required to be stored in a configuration with keff less than 0.8. Irradiated fuel is required to be stored in a configuration where convective cooling by water or air is adequate to maintain temperature below the safety limit.
Page 9-12
0 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 9 El Tt r,      01/2012 9.5      Fire protection systems Active fire protection elements generally have automatic operation, manual response, or personnel action for the intended function. Active elements to be considered include automatic fire detection, automatic fire suppression in labs and office spaces, fire information transmission, manual fire suppression and other manual fire control.
Passive fire protection provides fire safety that does not require physical operation or personal response to achieve the intended function. Passive elements include inherent design features, building physical layout, safety-related systems layout, fire barriers, and construction or component materials, and drainage for control of fire protection runoff water. Penetrations in fire barriers have fire resistant ratings compatible with the purpose of the fire barrier.
9.5.1    Design basis The goal of fire protection is to provide reasonable assurance that safety-related systems perform as intended and that other defined loss criteria are met1' 2. For the purpose of fire protection, loss criteria should include protection of safety-related systems, prevention of radioactive releases, personnel protection, minimization of property damage, and maintenance of operation continuity. Three components shall be applied to the fire protection objective. The three components are passive and active fire protection, and fire prevention.
A fire detection, suppression, and information management system is designed to ensure that fires can be detected, suppressed (where possible), and alert response organizations.
Basic design features of the reactor assembly, pool and shield system, and the instrumentation, control, and safety system represent passive fire protection elements. These basic features are sufficient passive protection to protect safety-related systems.
9.5.2    System description Manual protection consists of manual firefighting actions and the systems necessary to support those actions such as extinguishers, pumps, valves, hoses, and the inspection, maintenance and testing of equipment to assure reliability and proper operation. Other manual actions that are elements of active fire protection include utility control, personnel control, and evacuation.
Preplanning and training by facility and emergency personnel ensures awareness of appropriate actions in fire response and possible hazards involved.
'Code of Federal Regulations, Chapter 10 part 20, U.S. Government Printing Office, 1982.
2 Dorsey, N.E., Properties of Ordinary Water-Substance, Reinhold Publishing Corp., New York p. 537.
Page 9-13
CHAPTER 9, AUXILIARY SYSTEMS                                          -                    1/2012 Automatic and manual protection systems in the building include several different type systems. In the academic wing of the building automatic protective actions are provided by a sprinkler system with heat sensitive discharge nozzles, detectors for heat and smoke, and ventilation systems, dampers. The reactor bay wing uses smoke detectors for areas outside the reactor bay that are radiation areas.' The reactor bay ventilation system has smoke detectors that provide a warning of problems within the reactor bay.          Although not a strict safety requirement, a gaseous extinguisher system (halon) is installed to protect the reactor instrumentation, control and safety system.
Manual protection equipment includes a dry stand pipe system in each building stairwell.
Portable extinguishers such as C0 2, halon and- dry chemical are placed in specific locations throughout the building.
Elements of the passive fire protection include the structural construction system and the architectural separation into two separate buildings. Building structural materials are concrete cast in place for foundation, concrete walls, support columns and roof. Steel beam, metal and concrete deck comprises the reactor bay roof. A built-up composition roof with fire barrier materials completes the roof system. The building has pre-cast panels that are cast at the construction site cover 75% of the external perimeter. Metal paneling covers the other 25% of the perimeter. Design and installation of systems'and components are subject to the applicable building codes.                    .                        -"
The common wall between the academic win g.and the reactor bay wing of the building is a fire barrier. Doors between these two building sections and other penetrations such as HVAC chases will conform to applicable codes. Although a few metal stud and plaster board walls have been used in the reactor bay wing, the typical wall system is of concrete block construction.                                .
Design specifications are to meet life-safety requirements appropriate for the conditions. These specifications have requirements for emergency lighting, stairwells and railings, exit doors, and other building safety features.: An emergency shower -aid /yewash are available in the hallway adjacent to laboratory areas.,:
Eachof the three components of the fire protection program is applied to the design, operation and modification of the reactor facility and cornporients. Fir'eprevention is primarily a function of operation rather than design.
9.5.3    Operational analysis and safety function The University of Texas maintains an active fire protection system, with periodic testing and inspections to assure systems are prepared to respond.
The halon system automatically actuates if detectors in two control room ceiling units sense an initiating condition in coincidence. The haion system is equipped with a local alarm to prompt Page 9-14
THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR                      0                    01/2012 SAFETY ANALYSIS REPORT, CHAPTER 9                                            R'..flETI evacuation of the control room prior to system actuation; a manual override can defeat the system if the nature of the event does not require actuation 3f the system.
Fire suppression is used only in areas where there are no, significan't quan iites of radioactive materials or criticality concerns.
9.5.4    Instruments and controls A fire alarm panel transmits status and alarm information to the University of Texas Police Department dispatch station and a campus information network monitor.
9.5.5    Technical Specifications, bases, testing and surveillances There are no Technical Specifications associated with fire protection.
9.5      Communications systems A communication system of typical te!ephone equipment provides basic services between the building and other off-site points. Supplemental features to this system, such as intercom lines between terminals or points within the building and zone speakers for general announcements are to provide additional communication within the building.
9.5.1    Design basis Communications is required to support routine and emergency operations.
9.5.2    System description The telephone system is insta!led and maintained by the university. Connection of the main university telephone system is to standard commercial telephone network.. Telephones with intercom features are to be located at several locations in the building. Locations include the reactor control room, the reactor bay, and several offices. By use of the intercom feature, each of these locpations will be able to access public address speakers in one of several building zones.
A video camera system and a separate intercom system                    supplement the building telecommunication network. These two systems contribute to safe          operation by enhancement of visual and audio communication between the operator and an            experimenter. Each system has 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, as required. A building evacuation alarm system prcmpts personnel to initiate protective actions.
Page 9-15
CHAPTER 9, AUXILIARY SYSTEMS                                                                01/2012 An emergency cell phone is maintained in the control room to compensate for loss of normal communications.        A digital radio is kept in the control room to provide emergency communications on first responder and campus frequencies, and to compensate for loss of normal communications.
9.5.3    Operational analysis and safety function The control room has'adequate capabilities to initiate and coordinate emergency response.
There are multiple provisions specificallyto address failures on normal communications channels.
9.5.4    Instruments and controls As specified above 9.5.5    Technical Specifications, bases, testing and surveillances There are no specific Technical Specifications related to communications, but the reactor Emergency 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 supporting irradiation programs and a series of laboratories in the lab andoffice. wing.
9.6.1    Design basis The design;basisof the NETL laboratories is to allowthe safe arid controlled use of radioactive materials-.                    .
9.6.2    System Description Strategic lab and office wing rooms are equipped with fume hoods and ventilation to control the po.tential.for release of radioactive materials.: One rocm is equipped with two pneumatic transfer systems and a manual port. One system terminates in-a fume hood, monitored by a radiation detector. The other system delivers samples within the tube to a detector. The manual port allows samples to be transferred from the reactor bay to the lab without exiting the reactor bay through normal passageways. A more complete description of the associated laboratories is provided: in Chapter 10.
Page 9-16
0.
THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 9 9  MNtI I    01/2012 9.6.3  Operational analysis and safety function Engineered controls permit safe handling of radioactive materials.
9.6.4  Instruments and controls An installed radiation monitor ensures persornel handlir.g samples from the manual pneumatic sample transfer system are aware of the potential exposure.
9.6.5  Technical Specifications, bases, testing and surveillances There are no specific Technical Specifications related to the laboratories; all operations involved with potential radiation exposure at NETL are managed under the approved reactor Radiation Protection Program.
9.7    Control and storage of reusable components Several experiment facilities that are used in the core are designed to be removed and inserted as required to support various programs.
9.7.1  Design basis Management of experiment facilities is designed to minimize potential exposure to personnel.
9.7.2  System description The 3 element facility, 6 element faiity, pneumatic in-core terminals, and centra! thimble are described in chapter 10. Once irradiated, these facilities are maintained with activated portions in the pool, using pool water as shielding or in other locations typically within the reactor bay 9.7.3  Operational analysis and safety function Maintaining irradiated facilities under water minimizes potential exposure. Corncrete blocks provide temporar/ shielding as needed.
9.7.4  Instruments and controls Instruments and controls associated with specific facilities are addressed in Chapter 10.
Page 9-17
CHAPTER 9, AUXILIARY SYSTEMS                                                            1      01/2012 9.7.5    Technical Specifications, bases, testing and surveillances The basis for Technical Specifications specific to the pool is in Chapter 5, the basis for experiment in Chapter 10.
9.8      Compressed gas systems There are two separate compressed air systems use at the UT facility. One system provides air for laboratories and service connections. One system provides control air.
9.8.1    Design basis Service air is provided to support laboratory and service operations with high capacity applications. Instrument air is intended to support HVAC and reactor operations.
9.8.2    System description One dual compressor system provides oil free compressed air for laboratories and services. The lab air compressor motor is rated at 30 hp. The other system also uses a dual compressor and motor, with 2-stage compressors. The instrument air compressor provides air to HAVC pneumatic controls, pool cooling flow controls. The laboratory air compressor provides aiur to shops and to the transient rod drive system.
9.8.3    Operational analysis and safety function The two systems have dual motors and compressors to provide maximum reliability. The two systems are connected through a manual shut off valve, providing maximum flexibility in the event of a system (or associated air dryer) failure.
Failure of the instrument air system will prevent air from supporting control systems. The pulse rod drive system requires air to couple the drive to the rod; a failure will cause the rod to fall into 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 water system. 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 fail closed, initiating confinement isolation. This is a fail-safe condition, assuring that there is no potential for inadvertent release of radioactive material into the environment in the absence of instrument air.
Page 9-18
THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 9                                      0000 ETI          01/2012 9.8.4  Instruments and controls The 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 surveillances There are no Technical Specificatorns specifically associated with the compressed air systems.
Page 9-19}}

Latest revision as of 14:59, 6 February 2020

University of Texas at Austin - Redacted Supplement to Application Containing Table of Contents, Chapter 4 and Chapter 9 of the Safety Analysis Report
ML12156A196
Person / Time
Site: University of Texas at Austin
Issue date: 01/17/2012
From: Biegalski S
University of Texas at Austin
To: Lising A
Document Control Desk, Division of Policy and Rulemaking
Lising A
References
TAC ME7694
Download: ML12156A196 (97)
v • d • e


Text

UNIVERSITY OF TEXAS AT AUSTIN RESEARCH REACTOR LICENSE NO. R-129 DOCKET NO. 50-602 UNIVERSITY OF TEXAS AT AUSTIN LICENSE RENEWAL APPLICATION JANUARY 17, 2012 REDACTED VERSION*

SECURITY-RELATED INFORMATION REMOVED

  • REDACTED TEXT AND FIGURES BLACKED OUT OR DENOTED BY BRACKETS

DepanmL aoto i""hanical Engineering THE UNIVERSITY OF TEKAS AT AUSTIN Nuclearf.'5i*eerinig waaing taboratory AtArin, "7Txas78758 5.I 2-232-5370 -FAX 512-471- -589- htp,'/wu'A me.nrexas.edul/.-nel/

ATTN: Document Control Desk, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001 Allan Jason Lising Project Manager Division of Policy and Rulemaking Research and Test Reactors Licensing Branch January17, 2012

SUBJECT:

Docket No. 50-602, Information Supplementing Request for Renewal of Facility Operating License R-129 (TAC ME 7694)

REFERENCE:

(1) ML110040316 (2) Letter, 12/12/2011 Docket No. 50-602, Request for Renewal of FacilityOperating License R-129 Sir:

In accordance with USNRC direction (ADAMS ML110040316), a request for renewal of Facility Operating License R-129 (Docket 50-602) was submitted on 12/12/2011. The attached material provides minor editorial corrections and clarification of three chapters previously submitted of the Safety Analysis Report and the Technical Specifications. An additional item is included to support review of the proposed Technical Specifications. In summary:

  • Table of Contents (reflecting updates)
  • Chapter 4, additional figures are provided to better describe the control rod drive mechanisms, and the section on thermal hydraulic analysis was substantially augmented.
  • Chapter 9, operation of the auxiliary purge system and the confinement isolation was revised.
  • Chapter 12, the responsibilities of the Senior Reactor Operator was rewritten to emphasize the role of the Supervisor in reactor operations.
  • Technical Specifications, editorial changes and various improvements were made.
  • Technical Specifications review material: a tabulation of the current Technical Specifications was prepared with a comparison to the proposed, new Technical Specifications.

Your attention in this matter is greatly appreciated, I declare under penalty of perjury that the foregoing is true and correct.

0I THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 0*:.

000 l 01/2012 SAFETY ANALYSIS REPORT Table of Contents Section Page

1. THE FACILITY 1-1 1.1 Introduction 1-1 1.2 Summary and conclusions on principle safety considerations 1-2 1.3 General description of the facility 1-3 A. Site 1-3 B. Building 1-3 C. Reactor 1-3 C.1 Reactor Core. 1-5 C.2 Reactor Reflector. 1-5 D. Reactor Control. 1-6 E. Experiment Facilities. 1-6 E.1 Upper Grid Plate 7L and 3L Facilities 1-6 E.2 Central Thimble 1-6 E.3 Rotary Specimen Rack (RSR) 1-6 E.4 Pneumatic Tubes 1-7 E.5 Beam Port Facilities 1-7 E.5 (1) Beam Port 1 (BP1) 1-7 E.5 (2) Beam Port 2 (BP2) 1-8 E.5 (3) Beam Port 3 (BP3) 1-9 E.5 (4) Beam Port 4 (BP4) 1-10 E.5 (5) Beam Port 5 (BPS) 1-10 F Other Experiment and Research Facilities 1-10 1.4 Overview of shared facilities and equipment 1-10 1.4.3 Reference the other facilities operating history, safety and reliability 1-10 1.5 Summary of operations 1-12 1.6 Compliance with NWPA of 1982 1-12 1.7 Facility history & modifications 1-13 2.0 SITE DESCRIPTION 2-1 2.1 GENERAL LOCATION AND AREA 2-1 2.2 POPULATION AND EMPLOYMENT 2-7 2.3 CLIMATOLOGY 2-11 2.4 GEOLOGY 2-14 2.5 SEISMOLOGY 2-22 2.6 HYDROLOGY 2-22 2.7 HISTORICAL 2-27 3.0 DESIGN OF SYSTEMS, STRUCTURES AND COMPONENTS 3-1 3.1 Design Criteria for Structures, Systems and Components for Safe Reactor Operation 3-2 3.1.1 Fuel Moderator Elements 3-2

SAFETY ANALYSIS REPORT, TABLE OF CONTENTS 01/2012 Table of Contents Section Page 3.1.2 Control Rods 3-3 3.1.3 Core and structural Support 3-4 3.1.4 Pool and Pool Support Systems 3-4 3.1.5 Biological Shielding 3-4 3.1.6 NETL Building/Reactor Bay 3-5 A. Building 3-6 B. Reactor Bay 3-7 3.1.7 Ventilation Systems 3-7 3.1.8 Instruments and Controls 3-8 3.1.9 Sumps and Drains 3-8 3.2 Meteorological Damage 3-9 3.3 Water Damage 3-9 3.4 Seismic Damage 3-10 A. Core and structural Support 3-10 B. Pool and pool cooling 3-10 C. Building 3-10 4.0 Reactor 4-1 4.1 Summary description 4-1 4.2 Reactor Core 4-1 4.2.1 Reactor Fuel 4-2 A. Fuel matrix 4-2 (1) Fabrication 4-3 (2) Physical Properties 4-4 (3) Operational Properties 4-7 (4) Neutronic Properties 4-7 (5) Fuel Morphology & Outgassing 4-8 (6) Zr water reaction 4-9 (7) Mechanical Effects 4-9 (8) Fission Product Release 4-10 B. Cladding 4-10 4.2.2 Control Rods and Drive Mechanisms 4-11 A. Control Rods 4-13 B. Standard Control Rod Drives 4-16 C. Transient Control Rod Drive 4-17 D. Control Functions 4-19 E. Evaluation of the Control Rod System 4-20 4.2.3 Neutron Moderator and Reflector (Core Structure) 4-20 A. Upper grid plate 4-20 B. Reflector 4-23 ii

THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT

  • Ilk, I 01/2012 1/01 Table of Contents Section Page (1) Radial Reflector 4-23 (2) Graphite Rods. 4-24 (3) Axial Reflector 4-24 C. Lower grid plate 4-24 4.2.4 Neutron Startup Source 4-26 4.2.5 Core support structure 4-26 A. Core Support Platform 4-26 B. Safety plate 4-27 4.3 Reactor Pool 4-28 4.4 Biological Shield 4-30 4.5 Nuclear Design 4-32 4.5.1 Normal Operating Conditions 4-32 4.5.2 Nominal Reactivity Worth Values 4-33 4.5.3 Reactor Core Physics 4-32 A. Reference Calculations 4-34 B. Prompt Negative Temperature Coefficient 4-35 4.5.4 Operating Limits 4-39 A. Core Peaking Factors 4-39 B. Power distribution within a Fuel Element. 4-40 C. Power per rod 4-41 4.6 Core Reactivity 4-45 4.7 Thermal Hydraulic Design 4-47 4.7.1 Heat Transfer Model 4-48 4.7.2 Results 4-51 Appendix 4.1, PULSING THERMAL RESPONSE 4.1-1 5.0 REACTOR COOLANT SYSTEMS 5-1 5.1 Summary Description 5-1 5.2 Reactor Pool 5-1 5.2.1 Heat Load 5-2 5.2.2 Pool Fabrication 5-3 5.2.3 Beam Ports 5-3 5.3 Pool Cooling System 5-4 5.3.1 Reactor Pool 5-4 5.3.2 Pool Heat Exchanger 5-5 5.3.3 Secondary Cooling 5-10 5.3.4 Control System 5-10 5.4 Primary Cleanup System 5-11 5.5 Makeup Water System 5-12 5.6 Cooling System Instruments and Controls 5-13 iii

SAFETY ANALYSIS REPORT, TABLE OF CONTENTS 01/2012 Table of Contents Section Page 6.0 ENGINEERED SAFEGUARD FEATURES 6-1 6.1 References 6-1 7.0 INSTRUMENTATION AND CONTROL SYSTEM 7-1 7.1 DESIGN BASES 7-1 7.1.1. NM-1000 Neutron Channel 7-3 7.1.2. NP-1000 Power Safety Channel 7-5 7.1.3. Reactor Control Console 7-6 7.1.4. Reactor Operating Modes 7-7 7.1.5. Reactor Scram and Shutdown System 7-11 7.1.6. Logic Functions 7-12 7.1.7 Mechanical Hardware 7-13 7.2 DESIGN EVALUATION 7-14 8.0 ELECTRIC POWER SYSTEMS 8-1 9.0 AUXILIARY SYSTEMS 9-1 9.1 Confinement System ...... 9-1 9.2 HVAC (Normal Operations) 9-1 9.2.1 Design basis 9-2 9.2.2 System description 9-3 9.2.3 Operational analysis and safety function 9-4 9.2.4 Instruments and Controls 9-6 9.2.5 Technical Specifications, bases, testing and surveillances 9-8 9.3 Auxiliary Purge System 9-8 9.3.1 Design basis 9-8 9.3.2 System description 9-8 9.3.3 Operational Analysis and Safety Function 9-9 9.3.4 Instruments and controls 9-9 9.3.5 Technical Specifications, bases, testing and surveillances 9-10 9.4 Fuel storage and handling 9-10 9.4.1 Design basis 9-10 9.4.2 System description 9-10 9.4.3 Operational analysis and safety function 9-12 9.4.4 Instruments and controls 9-12 9.4.5 Technical Specifications, bases, testing and surveillances 9-12 9.5 Fire protection systems 9-13 9.5.1 Design basis 9-13 9.5.2 System description 9-13 9.5.3 Operational analysis and safety function 9-14 iv

THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR o I 01/2012 0401/ 2012 SAFETY ANALYSIS REPORT Table of Contents Section Page 9.5.4 Instruments and controls 9-15 9.5.5 Technical Specifications, bases, testing and surveillances 9-15 9.5 Communications systems 9-15 9.5.1 Design basis 9-15 9.5.2 System description 9-15 9.5.4 Instruments and controls 9-16 9.5.5 Technical Specifications, bases, testing and surveillances 9-16 9.6 Control, storage, use of byproduct material (including labs) 9-16 9.6.1 Design basis 9-16 9.6.2 System description (drawings, tables) 9-16 9.6.3 Operational analysis and safety function 9-17 9.6.4 Instruments and controls 9-17 9.6.5 Technical Specifications, bases, testing and surveillances 9-17 9.7 Control and storage of reusable components 9-17 9.7.1 Design basis 9-17 9.7.2 System description 9-17 9.7.3 Operational analysis and safety function 9-17 9.7.4 Instruments and controls 9-17 9.7.5 Technical Specifications, bases, testing and surveillances 9-18 9.8 Compressed gas systems 9-18 9.8.1 Design basis 9-18 9.8.2 System description 9-18 9.8.3 Operational analysis and safety function 9-18 9.8.4 Instruments and controls 9-19 9.8.5 Technical Specifications, bases, testing and surveillances 9-19 10.0 EXPERIMENTAL FACILTIES AND UTILIZATION 10-1 10.1 Summary Description 10-1 10.2 In-Core Facilities 10-3 10.2.1 Central Thimb;e (In-Core Facility) 10-4 A. DESCRIPTION 10-4 B. DESIGN & SPECIFICATIONS 10-5 C. REACTIVITY 10-6 D. RADIOLOGICAL ASSESSMENT 10-6 E. INSTRUMENTATION 10-7 F. PHYSICAL RESTRAINTS, SHIELDS, OR BEAM CATCHERS 10-7 G. OPERATING CHARACTERISTICS 10-7 H. SAFETY ASSESSMENT 10-8 10.2.2 Fuel Element Positions (In-Core Facilities) 10-8 V

SAFETY ANALYSIS REPORT, TABLE OF CONTENTS 01/2012 Table of Contents Section Page 10.2.2.1 Pneumatic Sample Transit System 10-8 A. DESCRIPTION. 10-8 B. DESIGN & SPECIFICATIONS. 10-9 C. REACTIVITY 10-10 D. RADIOLOGICAL ASSEMENT 10-11 E. INSTRUMENTATION 10-11 F. PHYSICAL RETRAINTS, SHIELDS, OR BEAM CATCHERS 10-12 G. OPERATING CHARACTERISTICS 10-12 H. SAFETY ASSESSMENT 10-12 10.2.2.2 Three Element Irradiator 10-13 A. DESCRIPTION. 10-13 B. DESIGN & SPECIFICATIONS. 10-13 B (1) Upper and Lower Grid Plate Modifications. 10-13 B (2) Alignment Frame. 10-14 B (3) Three Element Facility Canister. 10-14 C. REACTIVITY 10-16 C (1) Reactivity Calculation 10-17 C (2) Reactivity Measurements 10-18 D. RADIOLOGICAL ASSESSMENT 10-18 E. INSTRUMENTATION 10-19 F. PHYSICAL RESTRAINTS, SHIELDS, or BEAM CATCHERS 10-19 G. OPERATING CHARACTERISTICS 10-19 H. SAFETY ASSESSMENT 10-19 H (1) Cooling 10-19 H (2) Temperature 10-20 H (3) Pressure 10-21 H (4) LOCA potential 10-22 10.2.2.3 6/7 Element Irradiator 10-22 A. DESCRIPTION 10-22 B. DESIGN AND SPESIFICATIONS, 10-22 C. REACTIVITY. 10-23 D. RADIOLOGICAL ASSESSMENT 10-23 E. INSTRUMENTATION 10-23 F. PHYSICAL RESTRAINTS, SHIELDS OR BEAM CATCHERS 10-24 G. OPERATING CHARACTERISTICS 10-24 H. SAFETY ASSESSMENT 10-24 H (1) Temperature (Fuel) 10-24 H (2) Temperature (Lead) 10-24 H (3) Pressure (irradiation Can) 10-24 H (4) Pressure (Lead Sleeve) 10-25 vi

THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 0#-0, 01/2012 SAFETY ANALYSIS REPORT Table of Contents Section Page H (5) Mass 10-25 H (6) Structural 10-25 10.2.3 Rotary Specimen Rack 10-26 A. DESCPIPTION 10-26 B. DESIGN SPEC!,FiC1,1A*ATiN5 10-26 C. REACTIVITY 10-28 D. RADIOLOGICAL ASSESSMENT 10-28 E. INSTRUMENTATION 10-29 F. PHYSICAL RESTRAINTS, SHIELDS OR BEAM CATCHERS 10-29 G. OPERATING CHARACTERISTICS 10-29 H. SAFETY ASSESMENT 10-29 10.3 Beam Ports 10-29 A. DESCRIPTION 10-29 B. DESIGN AND SPECIFICATIONS 10-30 C. REACTIVITY 10-31 D. RADIOLOGICAL ASSESSMENT 10-31 E. INSTRUMENTATION 10-31 F. PHYSICAL RESTRAINTS, SHIELDS, OR BEAM CATCHERS 10-31 G. OPERATING CHARACTERISTICS 10-33 H. SAFETY ASSESSMENT 10-33 10.4 Cold Neutron Source 10-34 A. DESCRIPTION 10-34 B. DESIGN AND SPECIFICATIONS 10-34 C. REACTIVITY 10-37 D. RADIOLOGICAL 10-37 E. INSTRUMENTATION 10-37 F. PHYSICAL RESTRAINTS, SHIELDS, OR BEAM CATCHERS 10-39 G. OPERATING CHARACTERISTICS 10-39 H. SAFETY ANALYSIS 10-40 10.5 Non-reactor experiment facilities 10-41 10.5.1 Neutron generator room 10-41 10.5.2 Subcritical assembly 10-42 10.5.3 Laboratories 10-42 10.5.3.1 Radiochemistry laboratory 10-42 10.5.3.2 Neuron Activation Analysis Laboratory 10-43 10.5.3.3 Radiation detection laboratory 10-43 10.5.3.4 Sample preparation laboratory 10-43 10.5.3.5 General purpose laboratory 10-43 10.6 Experiment Review 10-43 vii

SAFETY ANALYSIS REPORT, TABLE OF CONTENTS 01/2012 Table of Contents Section Page 11 Radiation Protection and Waste Management 11-1 11.1 Radiation Protection 11-1 11.1.1 Radiation Sources 11-1 11.1.1.1 Airborne Radiation Sources 11-1 11.1.1.1.1 Production of Ar-41 in the Reactor Room 11-1 11.1.1.1.2 Radiological Impact of Ar-41 Outside the 11-2 Operations Boundary 11.1.1.2 Liquid Radioactive Sources 11-3 11.1.1.2.1 Radioactivity in the Primary Coolant 11-3 11.1.1.2.2 N-16 Radiation Dose Rates from Primary 11-4 Coolant 11.1.1.3 Solid Radioactive Sources 11-4 11.1.1.3.1 Shielding Logic 11-6 11.1.2 Radiation Protection Program 11-6 11.1.2.1 Management and Administration 11-7 11.1.2.1.1 Level 1 Personnel 11-7 11.1.2.1.2 Level 2 Personnel 11-7 11.1.2.1.3 Level 3 Personnel 11-9 11.1.2.1.4 Level 4 Personnel 11-10 11.1.2.1.5 Other Facility Staff 11-11 11.1.2.2 Health Physic Procedures and Document Control 11-11 11.1.2.3 Radiation Protection Training 11-11 11.1.2.4 Audits of the Radiation Protection Program 11-13 11.1.2.5 Health Physics Records and Record Keeping 11-13 11.1.3 ALARA Program 11-13 11.1.4 Radiation Monitoring and Surveying 11-14 11.1.4.1 Monitoring for Radiation Levels and 11-14 Contamination 11.1.4.2 Radiation Monitoring Equipment 11-15 11.1.4.3 Instrument Calibration 11-15 11.1.5 Radiation Exposure Control and Dosimetry 11-16 11.1.5.1 Shielding 11-16 11.1.5.2 Containment 11-16 11.1.5.3 Entry Control 11-17 11.1.5.4 Personal Protective Equipment 11-17 11.1.5.5 Representative Annual Radiation Doses 11-17 11.1.5.6 Personnel Dosimetry Devices 11-18 11.1.6 Contamination Control 11-18 11.1.7 Environmental Monitoring 11-19 11.2 Radioactive Waste Management 11-19 11.2.1 Radioactive Waste Management Program 11-20 viii

THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 01/2012 SAFETY ANALYSIS REPORT Ru01/201t Table of Contents Section Page 11.2.2 Radioactive Waste Controls 11-20 11.2.2.1 Gaseous Waste 11-20 11.2.2.2 Liquid Waste 11-21 112.2.3 Solid Waste 11-21 11 2.2.4 Mixed Waste 11-21 11.2.2.5 Decommissioning Waste 11-21 11.2.3 Release of Radioactive Waste 11-22 12 Conduct of Operations 12-1 12.1 Organization 12-1 12.1.1 Structure 12-1 12.1.1.1 University Administration 12-1 12.1.1.2 NETL Facility Administration 12-1 12.1.2 Responsibility 12-3 12.1.2.1 Executive Vice President and Provost 12-3 12.1.2.2 Vice President for University Operation 12-3 12.1.2.3 Associate Vice President of Campus Safety And 12-3 Security 12.1.2.4 Director of Nuc!ear Engineering Teaching 12-3 Laboratory 12.1,2.5 Associate Director of Nuclear Engineering 12-3 Teaching Labor3tory 12.1.2.6 Reactor Oversight Committee 12-4 12.1.2.7 Radiation Safety Officer 12-4 12.1.2.8 Radiation Safety Committee 12-4 12.1.2.9 Reactor Supervisor 12-4 12.1.2.10 Health Physicist 12-6 12.1.2.11 Laboratcry Manager. 12-6 12.1.2.12 Reactor Operators 12-6 12.1.2.13 Technical Support 12-6 12.1.2.14 Radiological Controls Technicians 12-6 12.1.2.15 Laboratory Assistants 12-7 12.1.3 Staffing 12-7 12.1.4 Selection and Training of Personnel 12-8 12.1.4.1 Qualifications 12-8 12.1.4.2 Job Descriptions 12-8 12.1.4.2.1 Facility Director 12-8 12.1.4.2.2 Associate Director 12-9 12.1.4.2.3 Reactor Supervisor 12-9 12.1.4.2.4 Health Physicist 12-9 ix

SAFETY ANALYSIS REPORT, TABLE OF CONTENTS 01/2012 Table of Contents Section Page 12.1.4.2.5 Laboratory Manager 12-9 12.1.4.2.6 Reactor Operators 12-9 12.1.4.2.7 Technical Support 12-9 12.1.4.2.8 Radiological Controls Technician 12-10 12.1.4.2.9 Laboratory Assistants 12-10 12.1.5 Radiation Safety 12-10 12.2 Review and Audit Activities 12-10 12.2.1 Composition and Qualifications 12-10 12.2.2 Charter and Rules 12-11 12.2.3 Review Function 12-11 12.2.4 Audit Function 12-12 12.3 Procedures 12-12 12.4 Required Actions 12-13 12.4.1 Safety Limit Violation 12-13 12.4.2 Release of Radioactivity 12-14 12.4.3 Other Reportable Occurrences 12-14 12.5 Reports 12-14 12.5.1 Operating Reports 12-15 12.5.2 Other or Special Reports 12-15 12.6 Records 12-16 12.6.1 Lifetime Records 12-16 12.6.2 Five Year Period 12-16 12.6.3 One Training Cycle 12-17 12.7 Emergency Planning 12-17 12.8 Security Planning 12-17 12.9 Quality Assurance 12-17 12.10 Operator Requalification 12-18 12.11 Startup Program 12-19 12.12 Environmental Report 12-19 13.0 ACCIDENT ANALYSIS 13-1 13.1 Notation and Fuel Properties 13-1 13.2 Accident Initiating Events and Scenarios 13-2 13.3 Maximum Hypothetical Accidents, Single Element Failure in Air 13-5 13.3.1 Assumptions 13-5 13.3.2 Analysis 13-6 A. Radionuclide Inventory Buildup and Decay, Theory 13-7 B. Fission Product Inventory Calculations 13-7 C. Fission Product release 13-10 D. ALl Consequence Analysis 13-11 x

THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR  :, -01/2012 SAFETY ANALYSIS REPORTI0/22 Table of Contents Section Page E. DAC Consequence Analysis 13-14 F. Effluent release Ccnsequence Analysis 13-17 F (1) Atmospheric Dispersion 13-18 F (2) CASE I 13-19 F (3) CASE II 13-20 F (3) Source Term Release Rate 13-22 13.3.3 Results and Conclusions 13-24 13.4 Insertion of Excess Reactivity 13-25 13.4.1 Initial Conditions, Assumptions, and Approximations 13-25 13.4.2 Computational Model for Power Excursions 13-26 13.4.3 Results and Conclusions 13-30 13.5 Loss of Reactor Coolant Accident 13-32 13.5.1 Initial Conditions, Assumptions, and Approximations 13-34 13.5.2 Heat Transfer to Air 13-34 A. Buoyancy Forces 13-35 B. Friction Losses 13-35 C. Losses from Flow Restrictions 13-35 13.5.7 Radiation Levels from the Unccvered Core 13-39 13.5.8 Results and Conclusions 13-43 13.6 Loss of Coolant Flow 13-43 13.6.1 Initialing Events and Scenarios 13-43 13.6.2 Accident Analysis and Determination of Consequences 13-43 13.7 Mishandling or Malfunction of Fuel 13-44 13.7.1 Initiating Events and Scenarios 13-44 13.7.2 Analysis 13-44 13.8 Experiment Malfunction 13-44 13.8.1 Accident Initiating Events and Scenarios 13-44 13.8.2. Analysis and Determination of Consequences 13-45 A. Administrative Controls 13-45 B. Reactivity Considerations 13-45 C. Fueled Experiment Fission Product Inventory 13-46 D. Explosives 13-47 13.9 Loss of Normal Electric Power 13-49 13.9.1 Initiating Events and Scenarios 13-49 13.9.2 Accident Analysis and Determination of Consequences 13-49 13.10 External Events 13-49 13.10.1 Accident Initiating Events and Scenarios 13-49 13.10.2 Accident Analysis and Determination of Consequences 13-50 13.11 Experiment Mishandling or Malfunction 13-50 13.11.1 Initiating Events and Scenarios 13-50 xi

SAFETY ANALYSIS REPORT, TABLE OF CONTENTS 01/2012 Table of Contents Section Page 13.11.2 Accident Analysis and Determination of Consequences 13-50 Appendix 13.1, T-6 DEPLETION ANALYSIS INPUT FILE FOR SCALE CALCULATION 13.1-1 Appendix 13.2, ORIGEN ARP INPUT 13.2-1 Appendix 13.3, MCNP INPUT FOR LOCA DOSES 13.3-1 15.0 FINANCIAL QUALIFICATIONS 15-1 15.1 Financial Ability to Operate a Nuclear Research Reactor 15-1 15.2 Financial Ability to Decommission the Facility 15-1 15.3 Bibliography 15-1 Appendix 15.1, STATUTES AND EXCERPTS REGARDING UT 15.1-1 Appendix 15.2, FIVE-YEAR OPERATING COST ESTIMATE 15.2-1 Appendix 15.3, Letter of Intent, Ultimate Decommissioning 15.3-1 Appendix 15.4, DECOMMISSIONING COST ESTIMATE 15.4-1 APPENDIX 15.5, FUELS ASSISTANCE CONTRACT 15.5-1 xii

THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 0*" 01/2012 SAFETY ANALYSIS REPORT  ?** REM LIST OF FIGURES Page Figure 1.1, UT TRIGA Mark II Nuclear Research Reactor 1-4 Figure 1.2, Core and Support Structure Details 1-5 Figure 1.3, Beam Ports 1-8 Figure 1.4A, Days of Operation per Year 1-12 Figure 1.4B, Burnup per Year 1-12 Figure 2.1, STATE OF TEXAS COUNTIES 2-2 Figure 2.2, TRAVIS COUNTY 2-3 Figure 2.3, CITY OF AUSTIN 2-4 Figure 2.4, JJ PICKLE RESEARCH CAMPUS 2-5 Figure 2.5, LAND USAGE AROUND JJ PICKLE RESEARCH CAMPUS, 2007 2-6 Figure 2.6, 2009 ZIP CODE BOUNDARIES 2-10 Figure 2.7, AUSTIN CLIMATOLOGY DATA 2-11 Figure 2.8, AUSTIN WIND ROSE DATA 2-12 Figure 2.9, TROPICAL STORM PATHS WITHIN 50 NAUTICAL MILES OF AUSTIN, TEXAS (ALL 2-21 RECORDED HURRICANES RATED H1 AND UP)

Figure 2.10, TROPICAL STORM PATHS WITHIN 50 NAUTICAL MILES OF AUSTIN, TEXAS (ALL 2-21 RECORDED STORMS RATED TROP OR SUBTROP)

Figure 2.11, BALCONES FAULT ZONE 2-23 Figure 2.12, TEXAS EARTHQUAKE DATA 2-24 Figure 2.13, TEXAS EARTHQUAKE DATA 2-25 Figure 2.14, LOCAL WATER AQUIFERS 2-26 Figure 2.15, RESEARCH CAMPUS AREA 1940 2-27 Figure 2.16, PICLKE RESEARCH CAMPUS 1960 2-28 Figure 2.17, BALCONES RESEARCH CENTER 1990 2-29 Figure 4.1: H/Zr Phase Diagram 4-6 Figure 4.2A, Zr-H Transport Cross Section & TRIGA Thermal Neutron Spectra 4-7 Figure 4.2B, Fuel Temperature Coefficient of Reactivity 4-7 Figure 4.3, Thermal Pressurization in Fuel and Hydriding Ratios 4-9 Figure 4.4A, Temperature and Cladding Strength for 0.2% Yield 4-11 Figure 4.4B, Temperature, Cladding Strength, and Stress 4-12 Figure 4.5, Lower Gird Plate Control Rod Positions 4-14 Figure 4.6, Standard Control Rod Configuration 4-15 Figure 4.7, Standard/Stepper Motor Control Rod Drive 4-16 Figure 4.8, Transient Rod Drive 4-18 Figure 4.9a, UT TRIGA Core 4-21 Figure 4.9b, Core Top View 4-21 Figure 4.10a, 6/7-Element Facility Grid 4-22 Figure 4.10b, Upper Grid Plate Cut-out for 6/7-Element Grid 4-22 Figure 4.1la, Reflector Top Assembly 4-23 Figure 4.11b, Reflector Bottom Assembly 4-23 xiii

SAFETY ANALYSIS REPORT, TABLE OF CONTENTS 01/2012 LIST OF FIGURES Page Figure 4.12b, Graphite Reflector Through port Detail 4-23 Figure 4.12c, Graphite Reflector, Radial & Piercing-Beam Ports 4-23 Figure 4.13a, Tangential Beam Port Insert 4-23 Figure 4.13b, Radial Beam Port insert 4-23 Figure 4.13c, Inner Shroud Surface 4-24 Figure 4.14, Reflector Component and Assembly Views 4-25 Figure 4.15, Fuel Element Adapter 4-26 Figure 4.16, Core Support Views 4-27 Figure 4.17, Core and Support Structure Views 4-27 Figure 4.18, Safety Plate 4-28 Figure 4.19a, Pool 4-29 Figure 4.19b, Side View 4-29 Figure 4.19c, Top View 4-29 Figure 4.20, Biological Shielding, Base Dimensions 4-31 Figure 4.21, Reactivity Loss with Power 4-34 Figure 4.22, Radial Variation of Power Within a TRIGA Fuel Rod. (Data Points from Monte 4-41 Carlo Calculations [Ahrens 1999a])

Figure 4.23, Critical Heat Flux Ratio (Bernath and Biasi Correlations) 4-44 Figure 4.24, Core Power, 45 kW Hot Element 4-45 Figure 4.25, Power Coefficient of Reactivity 4-46 Figure 4.25: Unit Cell Fuel Element Model 4-50 Figure 4.26a, Unit Cell Temperature Distribution (10.5 kW) 4-55 Figure 4.26b, Unit Cell Temperature Distribution (22.5 kW) 4-56 Figure 4.27, Single Rod Flow Cooling Flow Rate versus Power Level 49°C 6.5 Pool, 4-56 Figure 4.28, Comparison of Calculated and Observed Fuel Temperatures 4-58 Figure 5.1A, Pool Fabrication 5-4 Figure 5.1B, Cross Section 5-4 Figure 5.C, Beam Orientation 5-4 Figure 5.2, Pool Cooling System 5-4 Figure 5.3, Pool Cleanup System 5-11 Figure 5.4, Cooling and Cleanup Instrumentation 5-13 FIGURE 7.1, CONTROL SYSTEM BLOCK DIAGRAM 7-3 Figure 7.2, NEUTRON CHANNEL OPERATING RANGES 7-4 Figure 7.3, Auxiliary Display Panel 7-5 Figure 7.3, LAYOUT OF THE REACTOR CONTROL CONSOLE 7-6 Figure 7.4, CONSOLE CONTROL PANELS 7-8 Figure 7.5, TYPICAL VDEO DISPLAY DATA 7-9 Figure 7.6, ROD CONTROL PANEL 7-9 Figure 7.7, LOGIC DIAGRAM FOR CONTROL SYSTEM 7-13 Figure 9.1, Conceptual Diagram of the Reactor Bay HVAC System 9-2 Figure 9.2A, Main Reactor Bay HVAC System 9-3 Figure 9.2B, Main Reactor Bay HVAC Control System Control 9-4 xiv

THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR *a**

00F 01/2012 SAFETY ANALYSIS REPORT LIST OF FIGURES Page Figure 9.3, Confinement System Ventilation Contrcos 9-7 Figure 9.4A, Purge Air System 9-8 Figure 9.4B, Purge Air Controls 9-8 Figure 9.5A, Storage Well 9-11 Figure 9.5b, Fuel Storage Closure 9-11 Figure 10.1, Core Grid Plate Design and Dimensions 10-3 Figure 10.2, Reactor Core Diagram 10-4 Figure 10.3, Central Thimble Union Assembly 10-5 Figure 10.4, Three Element Irradiator 10-16 Figure 10.5, Rotary Specimen Rack Diagram 10-28 Figure 10.6, Rotary Specimen Rack Raceway Geometry 10-28 Figure 10.7, Rotary Specimen Rack Rotation Control Box 10-28 Figure 10.8, Beam Port Layout 10-30 Figure 10.9, A1230 Cryomech Cryorefrigerator and Cold Head 10-35 Figure 10.10, Cryomech Cold-Head and Vacuum Box 10-36 Figure 10.11, TCNS Vacuum Jacket and Other Instruments (units in cm) 10-36 Figure 10.12, Silicone Diode and Heater Relative to Cold-Head 10-37 Figure 10.13, Neon and Mesitylene Handling System with Pressure Transducers 10-38 Figure 10.14, Shielding around TCNS Facility 10-40 Figure 10.15, Thermo MP 320 Neutron Generator at NETL 10-41 Figure 10.16, Subcritical Assemblies 10-42 Figure 12.1, University Administration 12-2 Figure 12.2, NETL Facility Administration 12-2 Figure 13.1, Ratio of Radionuclide Inventory to ALl 13-13 Figure 13.2, Ratio of Radionuclide Concentration to 10CFR 20 DAC Values 13-14 Figure 13.3, FUEL Temperature and Pulsed Reactivity 13-35 Figure 13.4A, Pulse Measurements 13-31 Figure 13.4B, Fuel Temperature and Peak Pulse Power 13-31 Figure 13.5A, Cooling Time 13-37 Figure 13.5B, Cooling Time and Power Density 13-38 Figure 13.6, Core Model 13-41 Figure 13.7A, Bay Model Top View 13-41 Figure 13.7B, Bay Model Cross Section 13-41 Figure 13.8A, Building Model 13-42 Figure 13.8B, MCNP Side View 13-42 Figure 13.8C, Top View 13-42 xv

SAFETY ANALYSIS REPORT, TABLE OF CONTENTS 01/2012 Table Page Table 1.1, SHUTDOWN OR DECOMMISSIONED U.S. TRIGA REACTORS 1-10 Table 1.2, U.S. OPERATING RESEARCH REACTORS USING TRIGA FUEL 1-10 Table 2.1, AUSTIN AND TRAVIS COUNTY POPULATION TRENDS 2-8 Table 2.2, TRAVIS COUNTY 2009 AUSTIN POPULATION DENSITY DISTRIBUTION BY ZIP CODE 2-9 Table 2.3, 1982 METEOROLOGICAL DATA FOR AUSTIN TEXAS 2-15 Table 2.4, HISTORICAL METEOROLOGICAL DATA FOR AUSTIN TEXAS 2-16 Table 2.5, HISTORICAL METEOROLOGICAL DATA FOR AUSTIN TEXAS 2-17 Table 2.6, HISTORICAL METEOROLOGICAL DATA FOR AUSTIN TEXAS 2-18 Table 2.7, HISTORICAL METEOROLOGICAL DATA FOR AUSTIN TEXAS 2-19 Table 2.8, TRAVIS COUNTY TORNADO FREQUENCIES 2-20 Table 2.9 GROUND WATER ACTIVITY 2-26 Table 3.1, SSC Vulnerability 3-2 Table 4.1, TRIGA Fuel Properties 4-3 Table 4.2, Physical Properties of High-Hydrogen U-ZrH 4-4 Table 4.3, U-ZrH Volumetric Specific Heat Capacity (Cp) 4-6 Table 4.4, Summary of Control Rod Design Parameters 4-13 Table 4.5, Control Rod Information 4-15 Table 4.6, Summary of Reactor SCRAMs 4-19 Table 4.7, Summary of Control Rod Interlocks 4-19 Table 4.8, Upper Grid Plate Penetrations 4-21 Table 4.9, Displaced Fuel Spaces 4-22 Table 4.10, Lower Grid Plate Penetrations 4-25 Table 4.11, Reactor Coolant System Design Summary 4-28 Table 4.12, Significant Shielding and Pool Levels 4-32 Table 4.13, Control Rod Worth 4-33 Table 4.14, Reactivity Values 4-33 Table 4.15, GA-4361 Calculation Model 4-35 Table 4.16, Selected TRIGA II Nuclear Properties 4-35 Table 4.17, UT TRIGA Data 4-36 Table 4.18, Critical Heat Flux ratio, Bernath Correlation 4-43 Table 4.19, Core Power, 45 kW Hot Element 4-44 Table 4.20, Reactivity Limits 4-46 Table 4.21, Limiting Core reactivity 4-47 Table 4.22, Thermodynamic Values 4-49 Table 4.23, Hydrostatic Pressure 4-51 Table 4.24, Coolant Temperature for 49°C 6.5 m Pool 4-51 Table 4.25a, Outer Cladding Temperature (°C) for 49°C and 6.5 m Pool 4-52 Table 4.25b, Inner Cladding Temperature (°C) for 49"C and 6.5 m Pool 4-53 Table 4.26a, Heat Flux (Nodes 1-9) 49°C 6.5 Pool, 4-53 Table 4.26b, Heat Flux (Nodes 10-15) 49°C 6.5 Pool 4-54 Table 4.27, Peak Fuel Centerline Line Temperature (K)49°C 6.5 Pool, 4-54 Table 4-28, Coolant Flow for 1100 kW Operation 4-57 Table 4-29, Observed Fuel Temperatures 4-57 Table 4-30, Fuel Temperature Comparison 4-58 Table 5.1, Reactor Coolant System design Summary 5-2 Table 5.2, Heat Exchanger, Heat Transfer and Hydraulic Parameters 5-9 xvi

THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 0 j 01/2012 Table 9.1, Typical Confinement Vent & Purge Parameters 9-4 Table 9.2, Reactor Ventilation System Modes 9-5 Table 10.1: Composition of Al 6061 1.0-6 Table 10.2: Activation Products in Central Thimble 6061 Aluminum Alloy after 60 Year 10-7 Irradiation Table 10.3 Characteristic Dimension of UT-TRIGA PTS 10-10 Table 10.4: Activation of Pneumatic Transit System Cadmium Liner 10-11 Table 10.5: Flux Measurements in Pneumatic Transit System zt 100 kW 10-12 Table 10.6: Activity of Three Element Irradiator Cd Liner 10-19 Table 10.7: Rotary Specimen Rack Gears 10-27 Table 10.8: Items to be Addressed in Safety Analysis for Experiments 10-44 Table 11.1, Representative Solid Radioactive Sources 11-5 Table 11.2, Representative Radiation Detection Instrumentation 11-15 Table 11.3, Representative Occupational Exposures 11-17 Table 13.1. Neutronic Properties of TRIGA Mkll ZrH1.6 Fue! Elements. 13-1 Table 13.2, Dimensions of TRIGA Mkll ZrH1.6 Fuel Elementsl 13-1 Table 13.3, Thermal and Mechanical Properties of TRIGA Mkll ZrH1.6 Fuel Elements and 13-2 Type 304 Stainless Steel Cladding Table 13.4, UT TRIGA Core-Conditions Basis for Calculations 13-2 Table 13.5, Relevant IOCFR20 Appendix B Values 13-5 Table 13.6, SCALE T-6 Sequence Continuous Burnup Parameters 13-8 Table 13.7A, 1 MTU Gaseous Fission Product Inventory for 3.5 kW Case (Ci) 13-8 Table 13.7B, 1 MTU Particulate Fission Product Inventory (Ci) 13-9 Table 13.8A. Gaseous Fission product Release from Single Element (lVCi) 13-10 Table 13.8B. Particulate Fission Product Release from Single Element 13-11 Table 13.9A, Fraction of Gaseous Fission Product Inventory to 10CFR20 ALl 13-12 Table 13.9B, Fraction of Particulate Fission Product Inventory to IOCFR20 ALl 13-12 Table 13.10A, Fraction of Instantaneous Gaseous Fission Product Inventory to 10CFR20 13-14 DAC[1]

Table 13.10B, Fraction of Instantaneous Particulate Fission Product Inventory to 10CFR20 13-15 DAC [1]

Table 13.11, DAC Ratios for All Cases 13-16 Table 13.12, Reactor Bay Atmosphere Following MHA Compared to Effluent Limit 13-17 Table 13.13: BRIGGS URBAN DISPERSION PARAMETERS 13-18 Table 13.14, Calculated ?/Q Values 13-21 Table 13.15, Reactor Bay Atmosphere Following MHA Compared to Effluent Limit 13-21 Table 13.16, Calculated Plume Meander Factor (M) for < 6 m s-1 Winds 13-21 Table 13.17, Minimum Dispersion Parameters by Stability Class 13-22 Table 13.18, Minimum ?/Q by Stability Class 13-22 Table 13.19, Effluent Limit Ratio to Release Concentrations 13-23 Table 13.20, Low Power Pulsed Reactivity Response 13-28 Table 13.21, Initial Power 880 kW Pulsed Reactivity Response 13-30 Table 13.22, Gamma Source Term 13-39 Table 13.23, Height/Thickness Dimensions of Unit Cell 13-40 Table 13.24, Unit Cell Areas 13-40 Table 13.25, Material Characterization 13-40 Table 13.26, Post LOCA Doses 13-42 xvii

SAFETY ANALYSIS REPORT, TABLE OF CONTENTS 01/2012 Table 13.27, Calculations Supporting Limits on Fueled Experiments 13-46 Table 13.28, Material Strengths 13-48 Table 13.29, Container Diameter to Thickness Ratio 13-49 xviii

THE UNIVERSITY OF TEXAS TRIGA 11RESEARCH REACTOR 01/2012 SAFETY ANALYSIS REPORT, CHAPTER 4 i ET0/

4.0 Reactor This chapter will discuss the reactor core (fuel, control rods, reflector and core support, neutron source, core structure), reactor pool, biological shielding, nuclear design (normal operating conditions, and operating limits), and thermal hydraulic design.

4.1 Summary description The University of Texas Nuclear Engineering Teaching Laboratory (NETL) is home to a General Atomics' TRIGA Mark II research reactor. This installation follows 25 years (1963-1988) of successful operation of a TRIGA reactor at Taylor Hall on the main campus.

The basic TRIGA design uses U-ZrH 1.6 fuel clad with stainless steel in natural water convection cooling mode during operation, with a maximum decay heat that can be removed by natural convection of either water or air. The reactor is located in an open pool of purified, light water that serves as a heat sink during operations at power. Nuclear properties and characteristics control heat generation; thermodynamic characteristics of the fuel and the coolant control heat removal and temperature response. Maximum fuel temperature is the principle design constraint. 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 evolve into space around the fuel matrix; the hydrogen at elevated temperature can generate pressure inside the cladding. Temperature that produces stress greater than the yield strength for the stainless 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 mass exceeding critical loading (i.e., excess reactivity) is required to compensate for the negative fuel temperature coefficient, as well as potential experiments, fission product poisons, and fuel burnup. There are several major experiment facilities that could affect core reactivity, as described in Chapter 10. Experiment program requirements vary widely; limits are imposed on the reactivity effects of experiments. The amount of excess reactivity determines the maximum possible power, and therefore the maximum possible fuel temperature.

4.2 Reactor Core The University of Texas at Austin TRIGA II reactor core is configured in a hexagonal prism volume bounded by aluminum plates at the upper and lower surfaces (grid plates), and surrounded by a cylinder of graphite (aluminum clad) acting as a neutron reflector. Sections of the reflector are cut away to support experimental facilities, including beam ports and a rotating specimen rack. The core assembly is supported by structural aluminum, and includes an aluminum plate that serves to limit downward travel of control elements.

Page 4-1

CHAPTER 4: REACTOR 01/2012 4.2.1 Reactor Fuel The TRIGA fuel system was'developed around the concept of inherent safety, with fuel and cladding designed to withstand all credible envikonmental and radiation conditions during its lifetime at the reactor site. A TRIGA fuel element consists of (A) a central fueled region containing fuel matrix, bounded by an axial reflector and (B) stainless steel end caps at the top and 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 cladding temperature to prevent cladding rupture. The fuel lattice structure that comprises the NETL TRIGA 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 is adequate to limit maximum steady state operating temperature. ; The TRIGA fuel matrix exhibits a large prompt negative temperature coefficient ofý,reactivity. The maximum fuel temperature resulting from sudden insertion of all available excess reactivity would cause power excursion to terminate before any core damage is possible. Limits on core lattice excess reactivity and individual fuel element temperature therefore are interrelated. The maximum possible TRIGA fuel fission product inventory is limited by fissionable material loading. The maximum TRIGA fuel decay heat produced by fission product inventory can be removed by natural convection in air or water.

Handling, transport, and storage of TRIGA fuel elements at the NETL, fresh and irradiated, are described in Chapter 9, Auxiliary Systems.

A. Fuel matrix A TRIGA fuel element consists of a central fueled region containing fuel matrix, bounded by an axial reflector (with a molybdenum disk as a protective interface between the fuel and the lower graphite/axial reflector,- and stainless steel end caps-at the top and bottom with a stainless steel cladding.

The basic safety limit for the TRIGA reactor system is the fuel temperature; this applies for both the steady-state and pulse mode of operation. Twe, limiting temperatures are of interest, depending on the type of, TRIGA fuel used. The TRIGA fuel which is considered low hydride, that with an H/Zr ratio of less than ..5, has a lower temperature !imit than fuel with a higher H/Zr ratio. Fig. 4.1. indicates that the higher hydride compositions are single phase and

.are not subject to the large volume changes associated with the phase transformations at approximately 530'C in the lower hydrides. Also, it has been noted' that the higher hydrides lack any significant thermal diffusion of hydrogen. These two facts preclude concomitant volume changes. The important properties of delta phasetU-ZrH are given in Table 4.1.

'GA-3618, Thermal Migration of Hydrogen in Uranium-Zirconium Alloys, Marten U. et. Al., General Dynamics, General Atomics Division (1962)

Page 4-2

THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 4 R~eETI I 01/2012 Graphite dummy elements may be used to fill grid positions in the core. The dummy elements are of the same general dimensions and construction as the fuel-moderator elements. They are clad in aluminum and have a graphite length of Table 4.1, TRIGA Fuel Properties Property Mark III Dimensions Outside diameter, Do = 2ro Inside diameter, Dj= 2ri Overall length Length of fuel zone, L Length of graphite axial reflectors End fixtures and cladding Cladding thickness Burnable poisons Uranium content Weight percent U 23 5 U enrichment percent 235U content Physicalpropertiesof fuel excluding cladding H/Zr atomic ratio Thermal conductivity (W cm-' K-1 )

Heat capacity [T >0 0 C] (J cm 3 K')

Mechanicalpropertiesof delta phase U-ZrI-I0 Elastic modulus at 20'C Elastic modulus at 650'C Ultimate tensile strength (to 650'C)

Compressive strength (20'C)

Compressive yield (20'C)

(1) Fabrication A uranium loaded zirconium hydride was found to produce desired moderating characteristics and acceptably low parasitic neutron absorption with strong temperature feedback and high heat capacity. Feedstock of between (or recycled material) are cast in controlled atmosphere, high-temperature induction furnace. 2 Fuel element castings are machined to cylinders of approximately 5 inches in length. A center hole is drilled the length of the cylinder. Additional machining is required for fuel meat to be 2 TRIGA International: A New TRIGA Fuel Fabrication Facility at CERCA - Gerard Ilarbormier, Jean-Claude Ottone, CFIRCA, Proceedings of the 1997 TRTR Annual meeting Page 4-3

CHAPTER 4: REACTOR 01/2012 fabricated into instrumented fuel assemblies (IFEs, described below) and fuel element followers. The cylinders are heated in a high temperature electric furnace with a hydrogen atmosphere. The exterior and center surface: exposed to hydrogen induces the cylindrical fuel meat to hydride, with a target Zr:H ratio of 1'.6; :.A pure zirconium filler rod is placed in the center hole to maintain nearly uniform thermo-hydraulic properties. Each TRIGA fuel element contains 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 seal in:the upper end fixture, and a leadout tube provides a watertight conduit carrying the leadout wires 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 the transient rod (air filled follower) and the standard rods (fuel fo!lowers) respectively.

(2) Physical Properties The zirconium-hydrogen system is essentially a simple eutectoid, with at least four separate hydride phases. The delta and epsilon phases are respectively face-centered cubic and face-centered tetragonal hydride phases. The two phase delta + epsilon region exists between ZrHI.64 and ZrH1 .7 4 at room temperature, and closes at ZrH 1.7 at 455°C. From 455°C to about 10500 C, the delta phase is supported by a broadening range of H/Zr ratios. Other important properties observed for the delta phase U-ZrH are listed in Table 4.2.

The ratio of Zr-H plays a significant role ir determining physicai properties. The H:ZR material has a cubic structure in the delta-phase at ratios greater than 1.4. in lower H:Zr ratios (< 1.5) a phase change occurs at about 955°F (535°C) with large dens:ity differences between the phases leading to potential for deformation (swelling, and cracking). For hydrogen to zirconium atom ratios greater. than 1.5, the matrix is single phase (delta or epsilon) and does not exhibit phase separation with ,thermal cycling; Thermal diffusion of hydrogen is rrtinimal in higher ratios as well, minimizing potential for dceformaticrn from evolutiion of hydrogen gas. Any hydrogen gas is in equilibrium with: the matrix, substantially retained by the cladding, Losses through the cladding from hydrogen migration are about 1%for cladding temperature about 93 0 0F (500 0C).

Table 4.2, Physical Properties of High-Hydrogen U-ZrH Property Temperature Value Units Thermal Conductivity, 93°C - 650°C 0.22 W cm 6 psi 20°C 0 9.1x10 6 psi Elastic Modulus 650 C 6.0x10 Ultimate Tensile Strength 20°C 2.4 x10 4 psi Compressive Strength 20°C 6.0 x10 4 psi Compressive Yield 20'C 3.5 x10 4 psi Heat of Formation 298°C 37.75 kcal g-molr Page 4-4

00o°NlTi 01/2012 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 4 000o nE t At ratios greater than 1.6 there can be a shift to higher density tetragonal. Higher hydride compositions are single phase. and are not subject to the large volume changes associated with the phase transformations at approximately 530TC as in :the: lower hydrides. The stability extends from the minimum on the scale (OTC).to the maximum onwthe:scale (950TC), indicating no volume changes from morphology which might stress cladding occur for a target ratio of 1.6 other than thermal expansion. Significantly, zirconium hydrides at these ratios lack any significant thermal diffusion of hydrogen under isothermal conditions' Undernon-isothermal conditions, hydrogen migrates from high temperature regions to low temperature regions, with equilibrium disassociation pressures lower after redistribution. Hydrogen' dissociates slightly from the fuel matrix at high temperatures, and is re-absorbed :into the matrix at lower temperatures, with the equilibrium hydrogen dissociation pressure a function of both the composition and temperature. The equilibrium hydrogen dissociation pressure is governed by the composition and temperature. For ZrH 1 .6, the equilibrium hydrogen pressure is one atmosphere at about 760 0 C. Hydrogen dissociation pressures of hydrides are similar in alloys up to about 75 weight per cent uranium. For the delta and epsilon phases, dimensional changes from hydrogen migration are not significant. In the delta .phase, equilibrium disassociation pressures are related by:

K log pK +

With:

P pressure (atm)

T= temperature (K)

K1= -3.8415 + 38.6433-X - 34.2639.X + 9.282122X3 K2= -31.2981 + 23.5741,X -. 6.0280.X2 X= hydrogen to zirconium atom ratio At a ratio of 1.7 the equilibrium disassociation pressure corresponds to a temperature of about 1400F (300°C). The density of ZrH. decreases as hydrogen ratio increases; from low ratios to the delta phase (H:Zr of 1.5) the density change is high' with little cha&ngefor further increases.

Massively. hydrided bulk density' is-reported to be about 2% Ylower than x~ray diffraction analysis. For TRIGA-fuel with aZr:H ratio of 1:1.6, the uranium density,.volumefractioh, and weight fraction are related by:

WU p(A)=

0.177-0.125. WU and WU= 0.177-pu (A) 1 + 0. 125.-pu (A)

Page 4-5

CHAPTER 4: REACTOR CHPE,: ECO I 01/2012 0121 po, (A) = 19.07. V)J(A) where pu(A)= Uranium density WU - Uraniuri Weight fraction V1 = volume'fraaction of uranium in the U-ZrH1 .6 alloy Thermal conductivity ha*Ibeen determined from short-pulse heating techniques. Using thermal diffusivity values, density, and specific heat the thermal conductivity of uranium zirconium with a Zr:H ratio of 1:1.6 is 0.042 +/-+0.002 ca[-1 s-5 cm 'C -.

Volumetric specific heat is a function of temperature and composition. Table 4.3 lists values for variations in Zr:H and w% U based on a O°C reference, showing variation less than 10%.

Table 4.3, U-ZrH Volumetric Specific Heat Capacity (Cp)

ZrH W% U Value Units U-ZrH 1.6 8.5 2.04 + 4.17x10 W s .cmr3 3

U-ZrH. 7 20 2.17 + 4.36x10 W s -cr 850 w 600 P

450 I

I I h 0F 02 I I 04

@4.8 I

06 I

08 I

to I ---

12 0 16 L.

I LB m

I1 Zw HYDROGEN CONTENT (Di Zra Figure 4.1: H/Zr Phase Diagram Page 4-6:

01/2012 THE UNIVERSITY OF TEXAS TRIGA 11RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 4 (3) Operational Properties The neutronic properties of ZrH are the primary motivation for incorporation in TRIGA fuel development. The morphology of ZrH, in particular hydrogen diffusion in the material, imposes limits during operation. Ultimately, personnel exposure related to TRIGA fuel is limited during normal operations and abnormal events by retaining fission products in the fuel elements. It is well known that zirconium can undergo a reaction with water that releases hydrogen, with subsequent potential for a mixture that can be detonated. Such a reaction has the potential-,to release a large fraction of fission product inventory of affected fuel elements, but is not likely given characteristics of operation and properties of the fuel matrix. Fuel element changes occur during operation from thermal stress, which can affect fuel performance. Fuel cladding prevents migration of fission products for the fuel element, but in the absence of cladding it is not likely that all fission products will escape the fuel meat. Finally, thermal effects related to fuel matrix from steady state and pulsing operations are considered.

(4) Neutronic Properties A large fraction of neutron moderation occurs through interactions with hydrogen in the fuel matrix. The zirconium hydride structure has a profound effect on neutron scattering at low energies because of zirconium-hydrogen binding, with distinct lattice energy levels of 0.13 eV and about 0.25 eV found in scattering experiments. Thermal neutrons that interact wi~th.hydrogen in the lattice (where neutron energy is below the lattice energies) therefore have potential to gain energy. Because the fission cross section has 1/v dependence in the thermal range, increasing thermal neutron energy decreases fission probability. If fuel temperature increases, thermal excitation creates more of these relatively high-efnergy lattice centers as indicated in Fig. 4.2a. When the rate of fission is high enough to create elevated fuel temperatures, the elevated fuel temperatures decrease the rate of fission. This phenomenon is responsible for an extremely, high feedback of negative reactivity from fuel temperature illustrated in Fig. 4.2b. Maximum possible fuel temperature and maximum theoretical power level are therefore a function of the amount of fuel in the reactor.

I00 USTAINLESS STEEL CLAD

-12 8.5 WT-% U-ZrHj.6O CORE 00 " 400-C

-a a 260 401 ba 20' 0 L 0.01 0.... ........... . l.0 NEUTRON ENERGY (eVI 0 200 400 600 800 1000 /POO

'TEMPERATURE tIC)

Figure 4.2A, Zr-H Transport Cross Section & TRIGA Figure 4.2B, Fuel Temperature Coefficient of Thermal Neutron Spectra Reactivity Page 4-7

CHAPTER 4: REACTOR 1 01/2012 (5) Fuel Morphology & Outgassing As noted previously, during fuel fabrication the ratio of hydrogen to zirconium is enhanced by thermally induced diffusion in an atmosphere of pressurized hydrogen. During reactor operation, temperature gradients influence hydrogen diffusivity to promote outgassing, bounded by temperature induced pressurization of the hydrogen in free volume of the cladding. Pressure inside the fuel element does not intrinsically pose a challenge to fuel element integrity, and will be considered as part of cladding performance in a later section. At a given temperature, higher H:Zr ratios (in the absence of phase 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 of free hydrogen gas at higher temperatures also produces more partial gas pressure in the free volume of the element. Calculations performed with a higher mass fraction of uranium result in an increase in the partial pressure of hydrogen by as much as a factor of four. 3 The fuel rod diameter is on the order of the path length of neutron from generation to absorption, 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 outer surface of the fuel. Fuel rod temperature gradient during normal, steady-state operations is monotonically decreasing from a peak at the center of the fuel rod. Routine power changes occur at a rate that allows quasi-steady state thermal equilibrium, but pulsing operations do not. As a consequence, power distribution and development of temperature gradients in steady-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 than during steady state. Diffusion rates are finite, and the diffusion coefficient for thermal diffusion of hydrogen in zirconium 4 (ranging from 4x10 5 to 2x10 8 cm 2 s-1, and requiring days to equilibrate) lags the time cons-cant for the temperature changes. The temperature gradient during the transient peaks near the surface of the fuel rod rather than the center, and rapidly vanishes as the system comes to equilibrium. Therefore thermal gradients in pulsing bias hydrogen diffusion towards the center of the fuel rod with only a small region near the surface having a gradient that promctes outgassing. Surface cooling from endothermic gas emission lowers the surface temperature and therefore tends to iower the diffusion constant at the fuel rod surfaces. Re-absorption occurs where hydride surfaces are at relatively lower temperatLres. There is evidence that low permeability oxide films on fuel surfaces retard mass transfer. Local heat transfer effects cause the surface temperature to be lower than that which would occur during adiabatic conditions.

3 Journal of NUCLEAR SCIENCE and TECHNOLOGY, Vol. 37, No. 10, p. 887-892 (October 2000); Estimation of Hydrogen Redistributionin Zirconium Hydride under Temperature Gradient 4 Congreso Internacional de Metalugia y Materiales, Primeras Jornadas Internacionales de Materiales Nucleares (19 al 23 de Octubre de 2009, Buenos Airesm Argentina; Some Peculiarities of Hydrogen Behavior and Delayed Hydride Cracking in Zirconium Based Reactor Alloys, Shmakov, R.N. Singh Page 4-8

THE UNIVERSITY OF TEXAS TRIGA ImRESEARCH REACTOR 000 01/2012 SAFETY ANALYSIS REPORT, CHAPTER 4 0O

%03 1027 5

Z

/1,1/ o,,o~Ao DATA FROM GA-8129 AND NAA-SR-9374 600 700 80o 900 ICo I100 200 1300 TEMPERATURE( C)

Figure 4.3, Thermal Pressurization in Fuel and Hydriding Ratios Long term operations with steady state fuel temperatures exceeding 750'C (1023*K) may have time- and temperata ure-dependent fuel growth.5 Mechanisms contributing to the growth are identified as fission recoils and gaseous fission products, strongly influenced by thermal gradients. Analysis of steady state operating fuel,temperatures is provided in section 4.6, with pulsing operations fuel temperatures in Appendix 4.1.

(6) Zr water reaction Among th.e.chemical properties of U-ZrH: and ZrH, :the. reaction rate of the hydride with water is of particular interest. Since the hydriding reaction ýis exothermic, water will react more readily with zirconium than with zirconium hydride systems. Zirconium is frequently used in contact with water in reactors, and the zirconium-waterreaction is not a safety hazard.

Experiments carried out at GA Technologies show that.the zirconium hydride systems have a relatively low chemical reactivity with respect to water and air . These tests have involved the, quenching with water of both powders and solid specimens of.U-ZrH after~heating to as high as; 850°C, and of solid U-Zr alloy after, heating to as high as 1200*C. Tests have also been made to' determine the extent to which fission.products are removed from the surfaces of thefuel' elements at room temperature. Results prove that, .because of the high resistance to leaching, a large fraction of the fission products is retained in even completely unclad U-ZrH fuel..:

(7) Mechanical Effects At room temperature the hydride is like ceramic and shows little ductility. However, at the elevated temperatures of interest for pulsing, the material is found to be more ductile. The effect of very large thermal stress on hydride fuel bodies has been observed in hot cell observations to cause relatively widely spaced cracks which tend to be 5General Atomics Technical Report E-117-833 6

NUREG/CR-2387 Credible Accidents for TRIGA and TRIGA Fueled Reactors, S. C. Hawley,S. C. and Kathren, R. L.,

PNL-4208 (1982)

Page 4-9 I

CHAPTER 4: REACTOR A 01/2012 either radial or normal to the central axis and do not interfere with radial heat flow. Since the segments tend to be orthogonal, their relative positions appear to be quite stable. During fabrication, a molybdenum disk is placed between the lowest fuel mass and the lower axial-graphite reflector, minimizing potential for interaction that might affect the graphite and cause position changes in fuel meat that has developed surface imperfections. Anticipated mechanical effects from operation of the reactor are not expected to create conditions that challenge fuel performance.

7 (8) Fission Product Release Early in development of U-ZrHx fuel, experiments were performed 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 test conducted, with water flowing across the surface of the foil to remove fission products for analysis. The test was performed for 1 day and for 8 days with the total fractional fission product loss calculated to be between 10-7 and 10-s from preferential leaching of radionuclides, with gasses evolving from depths of 2.6 plm in the foil, and particulate from 22 A. Acceptable 8 upper values for release fraction are 1.0 x 10-4 for noble gases and iodine contained within the fuel, 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 noble gases, which is in SARs for other reactor facilities [NUREG-1390, 1990].

B. Cladding The 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 to release of those fission products that migrate to escape the fuel matrix surface. As noted previously, the free hydrogen in the space within the fuel element pressurizes the interior of the fuel element when fuel temperature is elevated during reactor operations. Power levels are acceptable if they do not result in temperatures that produce stress from the gas pressure that challenges the integrity of the cladding. A cylinder is considered a thin shell if wall thickness is less than about 10% of the radius and the classic equation for hoop stress created by internal pressure is:

o= P.r/t where:

oe is the hoop stress P is internal pressure r is inside radius t is the wall thickness 7 General Atomic report GA-655, Uranium-Zirconium-Hydride Fuel Elements, Merten, Stone, Wallace (1959) 8 NUREG/CR-2387, op. cit.

Page 4-10

THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 4 R~eETL I 01/2012 For stress is times the internal pressure. Fig.

4.4A provides temperature dependent ultimate strength and the 0.2% yield, and Fig. 4.4B shows where the hoop stress induced by the internal pressure intersects with ultimate strength. This intersection corresponds to a fuel temperature of 950'C for cladding temperatures greater than 500'C.

cc 103 400 500 600 700 800 900 1000 i'300 TEMPERATURE ( 0 C)

Figure 4.4A, Temperature and Cladding Strength for 0.2% Yield Page ,4-11

CHAPTER 4: REACTOR I 01/2012 Therefore, if fuel and cladding temperature remains below 950°C with cladding temperatures greater than 500°C, the stainless- steel cladding will not fail from overpressure. For cladding temperatures less than 500TC, hydrogen pressure from peak fuel temperature of 1150TC would not produce a stress in the clad in excess of its ultimate strength. The limiting fuel temperature and pressure is therefore the design basis for the UT TRIGA fuel. TRIGA fuel with a hydrogen to zirconium ratio of at least 1.65 has been pulsed to temperatures of about 1150TC without damage to the clad 9 .

1OS.

ULTIM~ATE 5TRENGTH 304. SS 1: S T R E NZ rH 1 .

65 o*,

1-io2.

10 2_-;

500 600 700 800 900 1000 1100 TEMPERATURE (*C)

Figure 4.4B, Temperature, Cladding Strength, and Stress 9 "Annual Core Pulse Reactor," General Dynamics, General Atomics Division report GACD 6977 (Supplement 2),

Dee. J. B., et. A].

Page 4-12

THE UNIVERSITY OF TEXAS TRIGA IIRESEARCH REACTOR N0ETL 01/2012 0.00 _

SAFETY ANALYSIS REPORT, CHAPTER 4 There are several reasons why the gas pressure should be less for the transient conditions than the equilibrium condition values would predict. For example, the gas diffusion rates are finite; surface cooling is believed to be caused by endothermic gas emission which tends to lower the diffusion constant at the surface. Reabsbr ption takes placelconcurrently on the cooler hydride surfaces away from the hot spot. There is evidence for a low permeability oxide film on the fuel surface. Some iocal, heat transfer does take place during the pulse time to cause a less than adiabatic true surface temperature.

4.2.2 Control Rods and Drive Mechanisms The control rods and drive mechanisms consist of (A) control rods, (B) standard, (or stepper) control rod drives, (C)transient rod drives, (D)control functions, and (E) system operation. The UT TRIGA reactor was installed with 4 control rods, three standard rods magnetically coupled to the control rod drive, and one pulse rod pneumatically coupled to the control rod drive. One of the standard rods, the regulating rod, is capable of being either automatically controlled with instrumentation and control systems described in Chapter 7 or manually from the reactor control console. The other control rods are manually shimmed. Principle design parameters for the control rods are provided in Table 4.4.

A. Control Rods The standard/stepper control rods (regulating and shim) are sealed 304 stainless steel tubes approximately 43 in. (109 cm) long by 1.35 in. (3.43 cm) in diameter in which the uppermost 6.5 in. (16. 5 cm) section is an air void, followed by 15 in. (38.1 cm) of a neutron absorber, solid boron carbide. Standard control rods have a fuel follower attached so that as the control rod is withdrawn from the core the water channel is filled with a fuel element as illustrated in Fig. 4.6.

The fuel follower, 15 in. (0.381 cm) of U-ZrH1 .6 fuel, is immediately below the neutron absorber of the standard control rods. The bottom 6.5 in. (16.5 cm) of the standard control rod is an air void. Table 4.4 summarizes control rod design parameters.

Table 4.4, Summary of Control Rod Design Parameters Cladding Material Aluminum SS 304 OD 1.25 in. 3.18 cm 1.35 in. 3.43 cm Length 36.75 in.. 93.35 cm 43.13 in. 109.5 cm ...

Wall thickness 0.028 in. 0.071 cm 0.02 in. 0.051 cm Poison Section Material Boron Carbide OD 1.19 in. 3.02 cm 1.31 in. 3.32 cm Length 15 in. 38.1 cm 14.25 in. 36.20 cm Follower Section Material Air U-ZrH.1 6 OD 1.25 in. 3.18 cm 1.31 in 3.34 cm Length 20.88 in. 53.02 cm Page 4-13

CHAPTER 4: REACTOR 1 01/2012 The transient (also called safety-transient or pulse) rod is a sealed, 36.75 in. (93.35 cm) long by 1.25 in. (3.18 cm) diameter tube containing boron in graphite as a neutron absorber. Below the absorber is an air filled follower section. The absorber section is 15 in. (38.1 cm) long and the follower is 20.88 in. (53.02 cm) long. The transient rod passes through the core in a perforated aluminum guide tube. The tube receives its support from the safety plate and its lateral positioning from both grid plates. It extends approximately 10 in. (25.4 cm) above the top grid plate. Water passage through the tube is provided by a large number of holes distributed evenly over its length. A locking device is built into the lower end of the assembly.

Control rods are withdrawn out of the core through the upper grid plate; when fully inserted the followers extend down through the lower grid plate. All fuel element position penetrations in the upper grid plate are identical; the lower grid plate (an excerpt in Fig. 4.5, fully described later in Chapter 4) has a set of 11 penetrations in the C and D rings (shaded in gray and black in Fig. 4.5, black representing the current configuration) with the same diameter as the upper grid plate. One of these penetrations in reserved for the central thimble (position Al) while the others are available for use as control rod positions. A safety plate is mounted below the lower grid plate as shown in Fig. 4.6, so that the control rod cannot exit the core region in the downward direction.

Figure 4.5, Lower Gird Plate Control Rod Positions Control rod worth is principally a function of control rod dimensions and location, experiment facilities in the core, with lessor influence by fuel and control rod burnup. Estimated control rod from the 1991 preliminary safety analysis report is provide in Table 4.5, along with the worth of each control rod as measured in June 2011. Sections of the control rod are separated and secured by 1-inch magneform fittings.

Page 4-14

THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 4 NETI I 01/2012 Table 4.5, Control Rod Informaton Rod Location Diameter Estimated,(1991). Current (2011)

In. cm. %Ak/k $S Transient Rod C Ring 1.25 3.18 2.1 3.00 3.10 Regulating Rod C ring 1.35 3.43 2.6 3.71 2.82 Shim 1 D ring 1.35 3.43 2.0 2.86 2.52 Shim 2 D ring 1.35 3.43 2.0 2.86 3.07

-U° Air 16.5 cm T.I J B4C F 38.1 cm LN U-ZrH 38.1 cm Air K,

6.5cmc..

Figure 4.6, Standard Control Rod ConfiguratioIn A threaded fitting at the end of each control rod connects to a series of shafts that connect to control rod drive mechanisms mounted' on a bridge that spans the reactor pool. The top section of the connecting shafts for standard rods passes through a hole in the bottom of a tube supported by the control rod drive housing. The tube is designed with slots that provide a hydraulic cushion for the rod during a scram, and also prevent the bottom of the control, rod from impacting the safety plate.

The shaft is secured to a cylinder that rests on the bottom of the housing when the rod is fully inserted. The top of the cylinder is secured to an iron core, engaged by an electromagnet for fail-safe control. The electromagnet is at the bottom of a small shaft controlled by the control rod drive mechanism. When the electromagnet is energized, the iron core is coupled to the drive unit.

Page 4m15

CHAPTER 4: REACTOR CHPE :RACO I 01/2012 121 The top section of the transient rod is connected to a single acting pneumatic cylinder which operates on a fixed piston, that couples the connecting rods to the drive. The transient rod drive is mounted on a steel frame that. bolts to the bridge. Any value from zero to a maximum of 15 in. (38.1. cm,):. of rod may:ibe withdrawn from the core; rod travel is limited by administrative control-not to exceed to the maximum licensed step insertion of reactivity.

B. Standard Control Rod Drives The rod drive mechanism for the standard rod drives is an electric stepping-motor-actuated linear drive equipped with a magnetic coupler and a positive feedback potentiometer. A stepping motor drives a pinion gear and a 10-turn potentiometer via a chain and pulley gear mechanism. The potentiometer is used to provide rod position information.

~MAGNET WIRE CONDUIT

-MAGNET DOWN ADJUSTMENT SCREW' MAGNET DRAW TUBE -MOTOR BIAS ADJUSTMENT T CENTER SWITCH ROD DOWN LIMIT SWITCH-'.. Y_-, IAOUNTING PLATE WIRE CONDUIT PULL-ROD ROD MOTOR PULL-ROD DRAW TUBE

  • PULL LOCK

.CONNECTING ROD Figure 4.7, Standard/Stepper Motor Control Rod Drive Page 4-16

THE UNIVERSITY OF TEXAS TRIGA 11RESEARCH REACTOR 000 o f 01/2012 SAFETY ANALYSIS REPORT, CHAPTER 4 00_ I The pinion gear engages a rack attached to the magnet draw tube. An electromagnet, attached to the lower end of the draw tube, engages an iron 3rm-nature. The armature is screwed and pinned into the upper end of a connecting rod that terminates at its lower end in the control rod. When the stepping motor is energized (via the rod' control UP/DOWN switch on the reactor control console), the pinion gear shaft rotates, thus raising the magnet draw tube, the armature and the connecting rod will raise with the draw tube so that the control rod is withdrawn from the reactor core. In the event of a reactor scram, the magnet' is de-energized and the armature will be released. The connecting rod, the piston, and the control rod will then drop, thus reinserting the control rod.

Stepping motors operate on phase-switched direct current power. The motor shaft advances 200 steps per revolution (1.8 degrees per' step). Since current is maintained -on the'motor windings when the motor is not being stepped, a high holding torque is maintained. The torque versus speed characteristic of a stepping motor is greatly dependent on the drive circuit used to step the motor. To optimize the torque characteristic for the motor frame size, a Translator Module was selected to drive the stepping motor. This combination of stepping motor and translator module produces the optimum torque at the operating speeds of the control rod drives. 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 Drive The safety transient control rod*drive is operated with a pneumatics rod drive. Operation of the transient rod drive is controlled from'the reactor control console. The transient rod is a scrammable rod operated in both pulse and steady-state modes of reactor operation. During steady state operation, the transient rod will function as an alternate safety rod with air continuously supplied to the rod. Rod position is thus controlled by'Operation of an electric motor that positions the air drive cylinder. The position of the transient control rod and its associated reactivity worth will generally dictate removal of the rod as the first step of a startup for steady-state operation. Rod withdrawal speed is about 28 ipm (1.E9cm/s).

The transient rod drive is a single-acting pneumatic cylinder with its piston attached to the transient rod through a connecting rod assembly., The piston rod passes through an air seal at the lower end of the cylinder. Compressed air is supplied to the lower end of the cylinder from an accumulator tank when a three -way solenoid valve located in the piping between the accumulator and cylinder is energized. The compressed air. drives the piston upward in the cylinder 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 the end of its travel, the piston strikes the anvil of an oil filled hydraulic shock absorber, which has a spring return, and which decelerates the piston at a controlled rate over its last 2 in. (5 cm.) of travel. When the solenoid is de-energized, a solenoid valve cuts off the compressed air supply and exhausts the pressure in the cylinder, thus allowing the piston to drop by gravity to its original position and restore the transient rod to a position fully inserted in the reactor core.

Page 4-17

CHAPTER 4: REACTOR I 01/2012 V EN.T VALVE Figure 4.8, Transient Rod Drive The extent of transient rod withdrawal from the core during a pulse is determined by raising or lowering the de~coupled cylinder, thereby controlling the distance the piston travels when air is applied. The cylinder has external threads running most of its length, which engage a series of ball bearings contained in a ball-nut mounted in the drive housing. As the ball-nut is rotated by a worm gear, the cylinder moves up or down depending onr!the direction of worm gear rotation.

A ten-turn Potentiometer driven by the worm shaft provides a signal indicating the position of the cylinder and the distance the transient rod will be ejected from the core. Motor 'operation for pneumatic cylinder positioning is controlled by a switch on the reactor control console. The magnet power key switch on the control console power supply prevents unauthorized firing of the transient rod drive.

Attached to and extending downward from the transient rod drive housing is the rod guide support, which serves several purposes. The air inlet connection near the bottom of the cylinder 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 rod assembly that moves the transient rod. The flanged connector stops the downward movement of the transient rod when the connector strikes the damp pad at the bottom of the rod guide support. A microswitch is mounted on the outside of the guide tube with its actuating lever Page 4-18

THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 4

  • eNETI I 01/2012 extending 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 the instrument 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 withdrawn position and the rod drops into the core from the fu!! out position in less than 1 second.

D. Control Functions Instrumentation and controls provide protective actions through the control rod system, as described in Table 4.6. A trip signal from the reactor protection system or the reactor control systems will deenergize the electro magnets and the pulse rod air solenoid valve previously described which allows gravity to insert the control rods.

Table 4.6, Summary of Reactor SCRAMs Limiting Trip Setpoint Measuring Channel Steady Pulse Actual Setpoint State SS - 1050 (NPP/NP) 1080 NM 1100 kW 2000 MW - 1 Pl 910 NP Maximum thermal power Pulse - 1910 NPP Power Channel High power 110% 110%

Detector High Voltage 80% 8C%

High Fuel Temperature 550°C Magnet current loss Manual Scram DAC and CSC watchdog timers In addition, the reactor control system (described in Chapter 7) has interlocks to prevent various conditions from developing. Table 4.7 is a summary of the functions.

Table 4.7, Summary of Control Rod Interlocks INTERLOCK SETPOINT FUNCTION/PURPOSE Inhibit standard rod motion if nuclear instrument S2 startup channel reading is less than instrument sensitivity/ensure nuclear instrument startup channel is operating Pulse Rod Interlock Pulse rod inserted Prevent applying power to pulse rod unless rod inserted/prevent inadvertent pulse Prevent withdrawal of more than 1 rod/Limit Muti dWithdrawal s , me maximum reactivity addition rate (does not apply in automatic flux control)

Prevent withdrawing standard control rods in pulse Pulse Mode Interlock Mode switch in Hi Pulse mode

.Pulse-Power Interlock 10 kW Prevent pulsing if power level is greater than 10 kW These safety settings are conservative in the sense that if they are adhered to, the consequence of normal or abnormal operation would be fuel and clad temperatures well below the safety limits indicated in the reactor design bases. Because of the conservatism in these safety Page 4-19

CHAPTER 4: REACTOR 01/2012 settings, it is reasonable that at some later date less restrictive safety system settings could be assigned in conjunction -with upgrading of the reactor to operate at higher steady-state power levels or in the pulsing mode while using the same fuel and core configuration.

Administrative limitations are imposed for the excess reactivity, transient conditions and coolant water temperature as follows:

1) Maximum core excess reactivity of 4.9% Ak/k ($7.00) with a shutdown margin of at least 0.2% 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, and

3) Core inlet water temperature of 48.9°C.

E. Evaluation of the Control Rod System The reactivity worth and speed of travel for the control rods are adequate to allow complete control of the reactor system during operation from a shutdown condition to full power. The TRIGA system does not rely on speed of control for reactor safety; scram times for the rods are measured periodically to monitor potential degradation of the control rod system. The inherent shutdown mechanism (temperature feedback) of the TRIGA prevents unsafe excursions and the control system is used only for the planned shutdown of the reactor and to control the power level in steady state operation. A scram. does not challenge the control integrity or operation, or affect the integrity or operation of other~reactor systems.

4.2.3 Neutron Moderator and Reflector (Core Structure)

The UT TRIGA core is supported within a reflector assembFy. The reflector assembly supports (A) an upper grid plate, (B) core barrel and reflector, and (C) lower grid plate, shown in Fig.

4.9a/b. The upper and lower grid plates provide alignment and support for the fuel elements.

A. Upper grid plate The upper grid plate provides alignment for fuel elements and control rods, and (in conjunction with the top fuel assembly) space for cooling flow. The.top grid. plate is fabricated from a circular aluminum plate 5/8 inches (1.59 cm.) thick and 21..6 in. (55.245 cm) diameter, anodized to resist wear and corrosion. The top of the upper grid plate is 59 in. (150 cm.) above the bottom of the pool. diameter are established on a triangular pitch of 1.714 in. (4.35 cm), separated by radial fuel arrays integrated 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 and guide tubes, the pneumatic transfer tube, and the central thimble. Small 0.203 in. (8 mm) holes at various positions in the top grid plate permit insertion of wires or foils into the core to obtain Page 4-20

THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 0 0 No NET_

T 01/2012 SAFETY ANALYSIS REPORT, CHAPTER 4 flux data. The flux probe holes are counter sunk/chamfered to (820) to 0.31 in. (11 mm). The center fuel element position is reserved as an experimental facility. The outermost fuel positions in the radial arrays are not fabricated for fuel insertion. Upper grid plate penetrations are summarized in Table 4.8.

01<

Ito. 2W

' /

Figure.4.9a, UT TRIGA Core Figure 4.9b, Core Top View The 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 grid plate using % in. (0.953 cm) holes centered along each of the hexagonal faces of the G ring fuel positions.

Table 4.8, Upper Grid Plate Penetrations

-Penetration Function..: Size "

Fuel Elements -, 1.505 in. (3.8227 cm), diameter 3Telement - .1.2 in. (3.048 cm). radius !

6/7-Element 2.2 in. (5.588 cm) radius Upper grid plate alignment 3/8 in. (0.9525 cm) diameter Flux probes 0.203 in. (0.5156 cm) diameter..

Fuel positions are :indexed by !etters denoting a "ring" where elements are colfinear with respect to the adjacent radial array fuei positions; A is the centralring position and G is furthest from the center. One radial array is used as a reference position, and the fuel positions range from 1 at the index to the maximum value for the ring, except for the G ring. Since the vertices of 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 facility encompassing the, central thimble since the central thimble does not contain fuel).

Page 4-21

CHAPTER 4: REACTOR I 01/2012 The inserts mesh in slots milled in the circular grid plate cutouts; 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 element facilities permit specimen as large as 2.4 in. (6.1 cm).

In addition to the experiment facilities that replace fuel positions, the current core configuration 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 or potentially displaced by core equipment. For control rods, only currently used positions are identified; there are alternate positions useable for control rods.

Table 4.9, Displaced Fuel Spaces Facility Core Location .

Page 4-22

0 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 4 0 4 nETL I 01/2012 B. Reflector The 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 fuel region.

(1) Radial Reflector. The radial reflector is a 10.2 in. (25.91 cm) graphite ring with an inner diameter 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 surface of the top section (Fig. 4.9a), with flat welded plates tying the top and bottom sections to the lift points. Angle plate structures are welded on the outer perimeter as points to secure the power level detectors. A 3 inch (7.62 cm.) wide well is fabricated in the top section (Fig. 4.11b),

and blocks with threaded penetrations are welded at the inner perimeter of the well to allow securing the rotary specimen rack (an experimental assembly) in the well.

Figure 4.11a, Reflector Top Assembly Figure 4.11b, Reflector Bottom Assembly The 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 of the reactor core, with one "through port" (Fig. 4.12a) and a 10 in. (25.3 cm.) cylinder cut from the inner surface to allow a 3 inch wide experimental facility surrounding the core.

1 Figure 4.12a, Graphite Reflector, Through Port Figure 4.12b, Graphite Reflector Through port Detail Page 4-23

CHAPTER 4: REACTOR 01/2012 I 01/2012 CHAPTER 4: REACTOR ------ a--

Figure 4.12c, Graphite Reflector, Radial & Piercing-Beam Ports The through port has a rectangular water-filled cut-out between the core shroud and the beam port penetration (Fig. 4.12b). Aluminum canisters that mate with the beam ports are nested in the reflector in two of the beam ports, one radial and one tangential (Fig. 4.12c, Fig. 4.13a/b).

The third beam port (radial) penetrates the core shroud (Fig. 4.13c).

Figure 4.13a, Tangential Beam Fort Insert Figure 4.13b, Radial Beam Port inert Figure 4.13c, Inner Shroud Surface (2) Graphite Rods. Graphite dummy elements may be uFAd to fP!. grid positions not filled by the fuel-moderator elemerts or other core compound, ,. They are of the same general dimensions and construction aý. the fuel-moderator elements, !ýut are f;lled entirely with graphite and are clad witi  !,'.hr*ir, :m.

(3) Axial Reflector. Graphite cylinders are placed above and below the fuel in the fuel elements. Fuel element construction was previously discussed.

C. Lower grid plate The lower grid plate (Fig. 4.14) provides alignment for fuel elements and control rods, and (in conjunction with the top fuel assembly) space for cooling flow. The lower (or bottom) grid Page 4-24

THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 4 0E~

I 01/2012 plate is fabricated from a circular aluminuLr. p*2t.e ý..75 inches (3.81 cm.), anodized to resist wear and corrosion. The top of the bottom *id piate is S.9 in. (25.19 cm.) above the bottom of the pool. The bottom grid plate is fabricated with fuel position penetrations and penetrations matching the flux probe holes on toe sarne, center 5s the upper grid plate, but also contains penetrations that support alignment cf the 3, G-,af:d 7 element facilities (Table 4.10). All but 11 fuel penetrations in the lower grid plate are sarnfller than the diameter of the fuel element and chamfered to provide a surface supporting triflutes on the bottom of the fuel element elements.

Table 4.10, Lower Grid Plate Penetrations Penetration Function Size Central thimble 1.505 Control Rod 1.505 Flux Hole Probes 8 mm 3-Element Alignment 3/8 in.

Lower grid plate alignment Lower Grid Plate Support Lower Grid Plate 11 Reflector Canister Bottom View Grid Plate in Core Shroud Figure 4.14, Reflector Component and Assembly Views Ten lower grid plate penetrations are the same diameter as the penetration in the upper grid plate, providing clearance for the central thimble and control rods. Since only 4 controls rods Page 4-25

CHAPTER 4: REACTOR 01/2012 are installed, unused control rod positions (i.e., large diameter holes) can be used for fuel with an adapter to support positioning the fuel above the lower grid plate (Fig. 4.15).

Figure 4.15, Fuel Element Adapter 4.2.4 Neutron Startup Source The reactor license permits the use of sealed neutron sources, including a is a standard sealed neutron source, encapsulated in stainless steel. The source is maintained in an aluminum-cylinder source holder of approximately the same dimensions as a fuel element. The source holder is manufactured as upper and lower (threaded) sections. The top of the lower section is at the horizontal centerline of the core. A soft'aluminum ring provides sealing against water leakage into the cavity.

The source holder may be positioned in any one of the fuel positions defined by the upper and lower grid plates. The upper end fixture of the source holder is similar to that of the fuel element; the source holder can be installed or removed with the fuel handling tool. In addition, the upper end fixture has a small hole through which one end of a stainless steel wire may be inserted to facilitate handling operation from the top of the tank.

4.2.5 Core support structure.

The core support structure includes (A)a platform supporting the reflector and core structure, and (B)a "safety plate" thaftprevents the control rods in a failure mode from falling out of the core.

A. Core Support Platform The reflector assembly rests on a platform (Fig. 4.16) constructed of structural angle 6061-T5 aluminum with a 3 in. x 3 in. x %in. (7.62 cm x 7.62 cm x 0.953 cm) web. Aluminum 6061-T651 plate is used for safety plate support pads (%in., 1.905 tmP), cross braces (% in., 0.953 cm.), and platform support pads (Y/in., 1.27 cm.). Angle aluminumr is inserted 9 iii. (22.86 cm)frorn two edges to support the safety plate, with angle bracing on the edges perpendicular to the safety plate supports.

Page 4-26

THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 4 IMeETI I 01/2012 Core Support Top View Core Support Side View Core Support Side View Figure 4.16, Core Support Views The platform top surface is 30 X in. X 30 Y4 in., with the top surface 16 Y4 in. above the pool floor. Surfaces are matte finished for uniform appearance with shot cleaning and peening using glass beads (MIL-STD -852).

-h 1 Core and Support Structure Assembly Core and Support Assembly Isometric Figure 4.17, Core and Support Structure Views B. Safety plate The safety plate (Fig 4.18) limits the distance that a control rod can fall to less than 17.44 in.

(44.30 cm) below the top surface of the lower grid plate. The safety plate is an aluminum plate Y2 in. (1.27 cm.) thick, 12 in. (30.48 cm) X 13.5 in. (34.29 cm), anodized to resist wear and corrosion (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 previously described, the bottom grid plate has a set of through-penetrations for optional placement of control rods. A special adapter is required to support fuel elements when these locations are used for fuel. The adapters have a central alignment pin that fits within holes in the safety plate, and an offset keeper-pin that prevents the adapter from rotating around the central pin.

Page 4-27

CHAPTER 4: REACTOR 01/2012 Figure 4.18, Safety Plate 4.3 Reactor Pool The reactor pool is a 26 foot, 11.5 in. (8.2169 m) tall tank formed by the union of two half-cylinders with a radius of 6 1/2z feet separated by 6 Y feet (1.9812 m). The bottom of the pool is 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 minimum level of 6.5 m (21.35 ft.) required for operations. Volume of water in pool (excluding the reflector, beam tubes and core-metal) is 40.57 mi3 and 32.50 m 3 for the nominal and minimum-required levels. Table 4.11 summarizes reactor coolant system design.

Table 4.11, Reactor Coolant System Design Summary Material Aluminum plate (6061)

Reactor Tank Thickness Y4 in. (0.635 cm)

Volume (maximum) 11000 gal (41.64 M 3)

Pipes .. Aluminum 6061 Coolant Lines Iron-Plastic Liner, 316 SS Ball and Stem Fittings . . Aluminum (Victaulic)

Type . Centrifugal Coolant Pump Material Stainless Steel Capacity 250 gpm (15.8 Ips)

Type Shell & Tube Materials (shell) " Carbon steel Materials (tubes) 304 stainless steel HeatExchanger Heat Duty Flow Rate (shell), "

Flow Rate (tubes) . -

Tube Inlet .... ..

Typical Heat Exchanger Iu e uOutlet Operating Parameters Shell Inlet Shell Outlet Page 4-28

THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 00 T 01/2012 SAFETY ANALYSIS REPORT, CHAPTER 4 _ 0000 The pool (Figs. 4.19a/b/c) is fabricated from sheets of 0.25 in. (0.635 cm) 6061 aluminum in 4 vertical sections welded to a Y2 in. thick aluminum plate. Full penetration inspection was performed 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 the beam 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 threaded connection to perform a leak test during fabrication. A 2 in. X 2 in. X Y4 in. (square) aluminum channel was rolled and welded to the upper edge of the tank.

Beam port penetrations are fabricated around the core to allow extractior of radiatior beams to support experiments. The beam ports are centered 90.2 cm (35 in.) above the pool floor, 7.2 cm (2.83 in.) below the core centerline. The section of the beam ports that are an integral part of the pool include an in-pool section, interface with the pool wall, and a section.

extending outside of the pool.

In pool sections are 6 in. (15.4 cm) in diameter, with a 0.635 cm (0.25 in.) wall thickness. The in pool section for BP 1 and 5 is 6 in. (15 cm), while the remaining in-pool beam port sections are much longer. Supports (2 in. X 2 in. X Y4 in. aluminum angle bracket) are welded at the bottom of the pool and directly onto. BP 2, 3, and 4 because of the extended lengths. BP 2 and 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 extends approximately 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 (14.8125 in.) from the in-pool end. BP 4 extends 43 cm into the pool (16.94 in.) with the support 7.62 cm (3 in.) from the in pool end. Beam port 1 and 5 are aligned in a single beam line. A flight tube inserted into BP 1/5 extends through the reflector near the core shroud; BP 1 and 5 are equipped with a bellows to seal a neutron flight-tube. Beam ports 2, 3, and 4 are sealed at the in-pool end. BP 2 is tangential to the core shroud, offset 34.29 cm (13 /2 in.) from core center rotated 3 0 0 with respect to BP 3. Beam port 3 is 90' with respect to BP 1/5, aligned to the center of the core. Alignment of BP 4 is through the core center, rotated 60' from BP 3.

The beam port interface with the pool wall includes a reinforcing flange on the inner pool wall. The flange is 3/8 in. thick, 11 in. in diameter. The flange is welded on the outer Page 4-29

CHAPTER 4: REACTOR 01/2012 diameter 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 pool walls. A stainless steel (304) ring is machined for a slip fit over the extension. The ring is welded to 6 5/8 in. diameter stainless steel pipe (SST 304W/ASTM 312) extending the flight tube 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, the in-pool beam port supports are welded to the pool floor.

Detection of potential pool leakage could occur in a number of ways.

1. Pool water level is maintained approximately 8.1 m above the pool floor, and monitored with an alarm on the control room console. A sudden decrease in pool water will create a condition that alerts the reactor operator at the controls.
2. Losses to evaporation are compensated by makeup water. Makeup water usage is closely monitored, and changes in makeup requirements or increases in makeup water that do not correspond to power level operation are a primary pool-leak indicator.
3. French drains around. the reactor pool shielding ,foundation are collected in a

.sump, and sampled periodically. Increases in radiation levels from the sump

..(particularly tritium) could indicate pool leakage.,

4.4 Biological Shield Pool water system and. shield structure (Fig. 4.20), design combine to control the effective radiation levels from, the, operation of the reactor.: One goal of the design is a radiological exposure constraint of 1 mrem/hour for accessible areas of the pool and shield system. Dose

.levels assume a full power operation level of 1.500 megawatts (thermal). Radiation doses above the pool' and at specific penetrations into or through the shield may exceed the design goal. The reference. case design is a solid structure without any system penetrations.

Tank assembly is by shop fabrication. A protective layer of epoxy paint and bitumen coal tar pitch with paper provides a barrier between the aluminum pool tank and the reactor shield concrete.

Page 4-30

THE UNIVERSITY OF TEXAS TRIGA 11RESEARCH REACTOR N 0001/2012 0

SAFETY ANALYSIS REPORT, CHAPTER 4 thick foundation pad supports the reactor pool and shield structure.

Standard weight concrete,' comprises the foundation pad. High density

concrete, of Five beam tubes atthe level of the reactor -provide experimental access, to:reactor neutron and gamma radiations. Two of the tubes combine to penetrate the complete reactor pool and shield structure from one side to the 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 leakage when the tubes are not in use. Beam port details are discussed in Chapter 10. A summary of significant component elevations and control functions is provided in Table 4.12.

Page 4-31

CHAPTER 4: REACTOR CHAPTER 4: REACTOR I 01/2012 01/2012 Table 4.12, Significant Shielding and Pool Levels Levei Notes Parameter of Interest (meters)

CONCRETE PAD FLOOR SAFETY PLATE GRID PLATE CORE BOTTOM BEAM PORT CL CORE CL CORE TOP 1 GRID PLATE MAIN LOWER SHIELDING TRANSITIONAL CONCRETE STEP SHIFT TO HIGH DENSITY C()NCRETE MIN CORE LEVEL (TS)

VACUUM-BREAKERS LOW POOL LEVEL SCRAM LOW POOL LEVEL op LOW POOL LEVEL ALARM NORMAL POOL LEVEL HIGH POOL LEVEL HIGH POOOL LEVEL ALARM'1 TOP OF TOP LEVEL 4.5 Nuclear Design The characteristics and operating parameters of this reactor,ý have been calculated and extrapolated using experience and data obtained from existing TRIGA reactors as bench marks in evaluating the calculated data. There are several TRIGA systems with essentially the same core and reflector relationship as this TRIGA so the values presented here are felt to be accurate to within 5% but, of course, are influenced by specific core configuration details as well as operational details. An operational core of 3 fuel followed control rods, and one air followed control rod is to be arranged in 5 rings with a central, water filled hole. Dimension of the active fueled core, approximated as cylinder, 15 in. (

cylinder radius is calculated as the average radius ofa hexagonal fuel array with 4.5.1 Normal Operating Conditions Reactivity worth of core components is generally determined by calculation and/or comparison of the reactivity worth associated with the difference in the reactivity worth of control rod positions in the critical condition, component installed and component removed. The 1992 UT SAR provided data indicating estimated worth of the control rods (Table 4.13). Control rod worth is influenced by core the experiment configuration, with significant impact from the large in core irradiation sites. Table 4.13 provides the worth of the control rods in the current configuration (3 element facility in Eli, F13, and F14). Change in core configuration require validation that control rod worth is not affected by the experiment facility, or re-establishment Page 4-32

THE UNIVERSITY OF TEXAS TRIGA 11RESEARCH REACTOR 00 A TI 01/2012 SAFETY ANALYSIS REPORT, CHAPTER 4 ....

of the control rod worth followed by verification that the limiting conditions for operation are met.

Table 4.13 Control Rod Worth Reference Current (2011),

Control Rd . Position Worth Position  ; Worth.

Transient rod C ring 2.1% Ak/k $3.00 C-1 $3.10 Regulating rod C ring 2.6% Ak/k $3.71 C-7 $2.82 Shim 1 D ring . 2.0% Ak/k $2.86 D-14 $2.52 Shim 2 D ring 2.0% Ak/k $2.86 D-6 $3.07 4.5.2 Nominal Reactivity Worth Values Reactivity values for core components based on calculations and observations are provided in Table 4.14, with Technical Specifications values in bold face type. Current values are based on measurements; nominal values are calculations frOm indicated sources.

'Table 4.14, Reactivity Values

$"TS CURRENT NOMINAL Parameter LIMIT VALUE VALUE 10 Reactor Reference Data Notebook, Safety Analysis, report Table 4-5; SAR Table 4-6 indicates CT Fuel $0.90, CT Void -$0.15, PNT Void -$0. 10, RSR void -0.20 "3-Element Experiment Authorization 12 Significant deviation from values in 3-Element Experiment Authorization (cf. E-Ring -$0.50 & D-Ring $0.95)

Page 4-33

CHAPTER 4: REACTOR I 01/2012 Table 4.14, Reactivity Values TS CURRENT NOMINAL LIMIT VALUE VALUE 4.5.3 Reactor Core Physics The 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 step insertions of reactivity is the prompt negative temperature coefficient (Fig. 4.21) associated with the TRIGA fuel and core design. This temperature coefficient allows a greater freedom in steady-state operation as the effect of incidental reactivity changes occurring from the experimental devices in the core is greatly reduced.

44.0 3.0 i-2.0

.5 t I I n I 0

0 200 100 600 800 1000 POWER (KW)

Figure 4.21, Reactivity Loss with Power Page 4-34

THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 4 SNETL I 01/2012 A. Reference Calculations A reference calculation of neutron flux distribution across the core was performed by General Atomics 13 . The calculations were accomplished on an .BM-7090 using General Atomics (diffusion theory based) codes GAMBLE and GAZE, and GAIV1-I. GAM-l is a fast neutron (using P1 treatment), temperature dependent (using methods developed by Nordhiem) cross section calculations for neutrons above 1 eV. GATHER-I was used to calculated cross sections below 1 eV. Homogenization was accomplished by the transport theory code DSN for group-dependent disadvantage factors (a second homogenization was accomplished for inhomogeneities in cells with control rods). No attempt was made to account for spatial variations in core temperatures. Basic core data for the calculations is provided in Table 4.15, with selected nuclear properties in Table 4.16. The model varies from the UT TRIGA reactor in specification of control rods, with one poison and three aluminum followers, where the UT TRIGA uses one aluminum and three poison followers; since this effects only the homogenization for two discrete cells, the results for core wide parameters is valid. UT TRIGA data is provided in Table 4.17.

Table 4.15, GA-4361 Calculation Model Area Volume volume Radius Fraction in. cm crn2 cm3 Cell Region U-ZrH 1.7 0.7175. 1.822 10.429 397.34 0.6308 SS Cladding 0.7375 1.873 0.592 22.56 0.0358 Water 0.9032 2.294 5.511 209.98 0.3334 TOTAL na na 16.532 629.88 1.0000 Table 4.16, Selected TRIGA II Nuclear Properties Number of cells 80 91 Fuel Temperature 23°C 200°C 1 eV to 10 MeV 1;a 0.00660 0.00675 if 0.00135 0.00135 Flux/watt 2.46x10 7 2.21x10 7 p'll 0.9405 0.9481 0 to 1 eV 1" 0.0873 0.0794 if 0.0526 0.0472 7 7 1.08x10 Flux/watt 1.11x10

% of fissions 94.6 94.5 Vave cm/s 2.73x10 5 2.94x10s Eae eV 0.0391 0.0455 NOTE 1: Resonance escape probability 13 GA-4361, CalculatedFluxes and Cross Sections for the TRIGA Reactors, G. B. West. August 1963 Page 4-35

CHAPTER 4: REACTOR 1 01/202 Table 4.17, UT TRIGA Data Core Configuration Ref Cold Clean Critical Loading : 64 elements Ref Operational Loading 90 elements Actulal'Initial Criticality Fuel' elemeht pitch 0.043536 cm

  • " Coolantlvolume to cell ratio .32.86%

... .Fuel Elem ents

  • Cladding, .. SS 304.

Fuel matrix .J-ZrH..6 Fuel Mass 2.5 kg Uranium fraction 8.5%

Enrichment 19.5%,

Nuclear Parameters Prompt neutron lifetime,( f) 41 ps Effective delayed neutron.

0 .007 fraction (13)

Prompt negative temperature 0 coefficient (a) 1x104 Ak/k C B. Prompt Negative Temperature Coefficient .

GA Technologies, the designer of the reactor, has developed techniques to calculate the temperature coefficient accurately and therefore predict the transient behavior of the reactor.

This temperature coefficient arises primarily from a change in the disadvantage factor resulting from the heating of the uranium zirconium hydride fuel-moderator elements. The coefficient is prompt because the fuel is intimately mixed with a large portion of the moderator and thus fuel and solid moderator temperatures rise simultaneously. A quantitative calculation of the temperature coefficient requires knowledge of the energy dependent distribution of thermal neutron flux in'the reactor.

The basic physical processes which occur when the fuel-'-moderator elements are heated can be described as follows: the rise in temperature of the hydride increases the probability.that a thermal neutr'on in the fuel element will gain energy from an excited state of an oscillating hydrogen atom in the lattice. As the neutrons gain e.nergy-from the ZrH, their mean free path is increased appreciably. Since the average chord length. in:the fuel element is comparable to a mean free path, the probability of escape from the fuel element before capture.is increased. In the water the neutrons are rapidly thermalized so that the ,capture and escape probabilities are relatively insensitive to the energy with which the neutron enters the water. The*heating of the moderator mixed with the fuel thus causes the spectrurn to h1airden more in the fuel than in the water. As a result, there is a temperature dependent disadvantage factor for the unit cell in the core which decreases the ratio of absorptions in the fuel to total cell absorptions as the fuel element temperature is increased. This brings about a shift in the core neutron balance, giving a loss of reactivity.

Page 4-36

THE UNIVERSITY OF TEXAS TRIGA IIRESEARCH REACTOR l' Tl 01/2012 SAFETY ANALYSIS REPORT, CHAPTER 4 ÷ n The temperature coefficient then, depends on spatial variations of the thermal neutron spectrum over distances of the order of a mean free path with large changes of mean free path occurring because of the energy change in a single collision. A quantitative description of these processes requires a knowledge of the differential slow neutron energy transfer cross section in water and zirconium hydride, the energy dependence of the transport cross section of hydrogen as bound in water and zirconium hydride, the energy dependence of the capture and fission cross sections of all relevant materials, and a multigroup transport theory reactor description 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 GA Technologies: GGC-3"4 , GAZE-2 1 s, and GAMBLE-5 16, as well as DTF-IV 17, an Sn multigroup transport code written at Los Alamos. Neutron cross sections for energies above thermal (>1 eV) 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 flux derived by solution of the 81 equations for each discrete reactor region composition. This code and its related cross-section library predict the age of each of the common moderating materials to within a few percent of the experimentally determined values and use the resonance integral work of Adler, Hinman, and Nordhein to generate cross sections for resonance materials which are properly averaged over the region spectrum. Thermal cross sections 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 the chemically bound moderator atoms. The bound hydrogen kernels used for hydrogen in the water were generated by the THERMIDOR code 1 8 using thermalization work of Nelkin1 9 . Early thermalization work by McReynolds et a120 on zirconium hydride has been greatly extended at GA Technologies 21, and work by Parks resulted in the SUMMIT t251 code, which was used to generate the kernels for hydrogen as bound in ZrH. These scattering models have been used to predict adequately the water and hydride (temperature dependent) spectra as measured at the GA Technologies linear accelerator as shown in section 4.2.1 (A).

'4 General Atomics Report GA-7157, "Users and Programmer Manual for the GGC-3 Multigroup Cross Section Code," General Dynamics, General Atomic Division (1967) is General Atomics Report GA-3152 "GAZE-2: A One-Dimensional, Multigroup, Neutron Diffusion Theory Code for the IBM-7090," Lenihan, S. R., 'General Dynamics, General Atomic Division (1962) 16 General Atomrics Report GA-818, "GAMBLE A program for the Solution for the Multigroup Neutron-Diffusion Equations in Two Dimensions, with Arbitrary Group Scattering, for the UNIVAC-1108 Computer," 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 Equation with Anisotropic Scatterings, Los Alamos Scientific Laboratory, new Mexico (1965) 18 "THERIMIDOR- A FORTRAN II Code forCalculating the Nelkin Scattering Kernel for Bound Hydrogen (A modification of Robespierre),"Gulf General Atomic, Inc. (unpublished data) Brown, H. D., Jr.

19 "Scattering of Slow Neutrons by Water," Phys. Rev., 11, 741-746, Nelkin, M. S. (1960) 20 "Neutron Thermalization by Chemically-Bound Hydrogen and Carbon," Proc. nd Intl. Conf. Peaceful 2

Used at Energy (A/Conf. 15/F/1540), Geneva, IAEA (1958) 21 General Atomics Report GA-4490 Neutron Interactions in Zirconium Hydride, Whittenmore, W. L.,

General Dynamics, General Atomic Division (1964)

Page 4-37

CHAPTER 4: REACTOR 01/2012 Qualitatively, the scattering of slow neutrons by zirconium hydride can be described by a model in which the hydrogen atom motion is treated as an isotropic harmonic oscillator with energy transfer quantized -in multiples. of 70..14 eV. More precisely, the SUMMIT model uses a frequency spectrum with two branches, one for the optical modes for energy transfer with the bound proton, and the other for-the acoustical modes for energy transfer with the lattice as a whole. The optical modes are represented as a. broad frequency band centered at 0.14 CV, and whose width is adjusted to fit the cross section data of Woods et al. 1281. The low frequency acoustical modes are assumed to have a Debye spectrum with a cutoff of 0.02 eV and a weight determined by an effective mass of 360.

This structure then allows a neutron to slow down by thetransition in energy units of 0.14 eV as long as its energy is above 0.14 eV. Below 0.14 eV the neutron can still lose energy by the inefficient process of exciting acoustic Debye type modes in which the hydrogen atoms move in phase with the zirconium atoms, which in turn move in phase with one another. These modes therefore, correspond to the motion of a group of atoms whose mass is much greater than that of hydrogen, and indeed even greater than the mass of zirconium. Because of the large effective mass, these modes are very inefficient for thermalizing neutrons, but for neutron energies below 0.14 eV they provide the only mechanism for neutron slowing down within the ZrH. (In a TRIGA core, the water also provides for neutron thermalization below 0.14 eV.) In addition, in the ZrH it is possible for a neutron to gain one or more energy units of -0.14 eV in one or several scatterings, from excited Einstein oscillators. Since the number of excited oscillators present in a ZrH lattice increases with temperature, this process of neutron speeding up is strongly temperature dependent and plays an important role in the behavior of ZrH moderated reactors.

Calculations of the temperature coefficient were done in the following steps:

a. Multigroup cross sections were generated by the GGC-3 code for a homogenized unit cell. Separate cross-section sets were generated for each fuel element temperature by use of the temperature dependent hydride kernels and Doppler broadening of the 238 U resonance integral to reflect the proper temperature. Water at room temperature was used for all prompt coefficient calculations.
b. A value for. k- was computed for each fuel element temperature by transport cell calculations, using the P1 approximation. Comparisons have shown 54 and S8 results to be nearly identical. Group dependent disadvantage factors defined as (Dg'/ (1gc (region cell) were calculated for each cell region (fuel, clad, and water).
c. The thermal group disadvantage factors were used as input for a second GGC-3 calculation where cross sections for a homogenized core were generated which gave the same neutron balance as the thermal group portion of the discrete cell calculation.

Page 4-38

THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR *g0 fl0 T I " 01/2012

.---°°O 0

  • SAFETY ANALYSIS REPORT, CHAPTER 4
d. The cross sections for an equivalent honmogenized core were used in a full reactor calculation to determine the contribution to the temperature coefficient due to the increased leakage of therma! neutrons into <the, reflector with increasing hydride temperature. This calculation requires severa! thermnalgrbups, butiransport effects are no longer of major concern. Thus, reactivity calculati6ns :s a fbiit-tion of fuei element temperature have been done on the entire reactor with the use of diffusion theory codes.

Results from the above calculations indicate that more than 50% of the temperature coefficient for a standard TRIGA core comes from the temperature-dependent disadvantage factor or "cell effect", and ~20% each from .Doppler broadening of the 2 3 8 U resonances and temperature dependent 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 .is needed to overcome temperature and .allow the reactor to operate at the higher power levels in steady-state operation. Fig. A.19 shows. the. relationship of reactor power level and associated reactivity loss to achieve a given power level.

4.5.4 Operating Limits The core-wide operating limits associated with nuclear design are based on spatialdistribution of neutron flux that determines*the local peak power production. Therefore (A)., the peaking factors 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 and capabilities. - . .

A. Core Peaking Factors The core.is generally modeled as aýrght cylinder. Neutron flux,.varies along the. axis of a cylindrical reactor using periodic Bessel functions. Neutron flux varies: radially in a cylindrical reactor using period sine functions. The product of these two functions. provides a relationship between average core power and the maximum power at a location within the core. Neutron flux and fission rate also varies significantly across the radius of a TRIGA fuel element; the complexities of the system do notlend themselves to reasonable analytic description.

Core Radial Peaking Factor. Classically, the radial hot-channel factor for a cylindrical reactor (using R as the physical radius and Re as the physical radius and the extrapolation distance) is given 2by:

22 Elements of Nuclear Reactor Design, 2nd Edition (1983), J. Weisman, Section 6.3 Page 4-39

CHAPTER 4: REACTOR 01/2012 1.202 *(R

./[2.4048"(4 Dl However, TRIGA fuel elements are on the order of a mean free path of thermal neutrons.

Consequently, there is a significant change in thermal neutron flux across a fuel element.

Calculated thermal neutron flux data 23 indicates that the ratio of peak to average neutron flux (peaking factor) for TRIGA 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 actual case is 14% less than power calculated using the assumption.

Core Axial Peaking Factor. The axial distribution of power in the hottest fuel element is sinusoidal, with the peak power a factor of rn/2 times the average, and heat conduction radial only. The axial factor for power produced within a fuel element is given by:

g(z) 1.514"cos(',*z in which £ = L / 2 and f.., is the extrapolation length in graphite, namely, 0.0275 m. The value used to calculate power in the limiting -location-within the fuel element is therefore 4% higher than power calculated with the actual peaking factor. Actual power produced in the most limiting actual case is 4% less than power calculated using the assumption; therefore calculated temperatures will bound actual temperatures.

Core Local Peaking Factor. The location on the fuel rod producing the most thermal power with thermal power distributed over N fuel rods is therefore:

q N D, L P

B. Powerdistribution:within a Fuel Element. -

The radial and axial distribution~of the power'within a.fuel.element is given by q.'(r,z)= qfL(r)g(z) in which r is measured from the vertical axis of the fuel element and z is measured along the axis, from the center of the fuel element. The axial peaking factor follows from the previous assumption of the core axial peaking factor, but (since there is a significant flux depression across a TRIGA fuel element) distribution of power produced across the radius of the fuel the radial peaking factor requires a different approach than the previous radial peaking factor for the core. The radial factor within a fuel element is given by:

23 GA-4361, Calculated Fluxes and Cross Sections for TRIGA Reactors (8/14/1963), G.' B. West Page 4-40

THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 4 000 000IIETI I 01/2012 er2 f a+cri+

I + br + dr 2 in which the parameters of the rational polynomial approximation Iare derived from flux-depression calculations for the TRIGA fuel2 4. Values for the coefficients are: a = 0:824461 b =

-0.26315, c = -0.21869, d = -0.01726, and e = +0.04679. The fit is IElustrated anFig. 4.20.

1.3 1.2 1.

L 1.0 0.90 0.80 0.70 0.0 0.20 0.40 0.60 0.80 1.0 1.2 1.4 1.6 1.8 2. 0 Figure 4.22, Radial Variation of Power Within a TRIGA Fuel Rod.

(Data Points from Monte Carlo Calculations [Ahrens 1999a])

C. Power per rod The Bernath correlation 25 calculates critical heat flux as:

Q AIBO (Two -TB)

Where the convection heat transfer coefficient for "burnout" condition is, determined by:

hBo .10~990 (De2+ D) + SLOPE . V With two possible values for the "SLOPE" term:

(1) IF De< 0.1 ft.,

24 Report KSUNE -Investigation of the Radial Variation of the Fission-Heat Source in a TRIGA Mark III Fuel Element Using MCNP, Ahrens, C., Department of Mechanical and Nuclear Engineering, Kansas State University, Manhattan, Kansas (1999) 25 ANL/RERTE/TM-07-01, Fundamental Approach to TRIGA Steady state Thermal-Hydraulic CHF Analysis Page 4-41

CHAPTER 4: REACTOR 01/2012 48 SLOPE = 0.

SLOPE =9 9+ -

D.e And the burnout wall temperature term is calculated:

P V

+

P54 TwBo = 57- In(P) - P +15 4 2

The CHF heat flux in is p.c.u./hr-ft , the heat transfer coefficient corresponding to the CHF in 2

p.c.u./hr-ft -C, is the wall temperature at which CHF occurs in °C, Tb is the local bulk coolant temperature in TC, D hydraulic diameter of the coolant passage in feet, D is the diameter of the e

heater surface (heated perimeter divided by n) in feet, P is the pressure in psia, and V is the velocity of the coolant in ft/s. Substituting equivalent terms into the CHF equation results in:

[OIt10890.

[A8 0

(= 80

( De, 8

+x--*V)"

De+ DiJDe*

57-in(P) P_+-i ,r T

)~l~)~+5 4 B Where A is the flow area and WP the wetted perimeter, hydraulic diameter is calculated:

De=- 4-A WP (1) Wetted perimeter:

WP=- r - D *,

2 (2) Flow area:

A=PITCH2 f---r41 4 2 2)

Page 4-42

THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 4 0000 000 ETII 01/2012 TRACE calculates CHFR using the Biasi correlation, however the more accepted CHF correlation for TRIGA reactors is provided by Bernath 26. TRACE calculations completed as described in section 4.6 provides thermal hydraulic parameters used to calculate critical heat flux using the Bernath correlation (and the ratio of the heat flux to the critical heat flux, CHFR). The results of calculations using heat flux and temperature data for 49 0 C water at .6.5 m level is provided in Table 4.18. The minimum CHFR versus power level is provided in Fig. 4.22. As illustrated, the CHFR values agree well and remain much greater than 2 at power levels up to 22.5 kW per unit cell.

Table 4.18, Critical Heat Flux ratio, Bernath Correlation kW 1 2 3 4 5 ' 6 7 8 9 10 11 12 13 14 15 1.5 106.2 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 96.8 3.0 61.0 53.4 47.2 42.2. 38.5 37.0 35.7: .34.4 35.5 36.6 37.2 39.7 42.9 '46.8 5i.9 4.5 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 35.5 26.9 6.0 35.4 30.9 23.6 21.6 20.8 20.0 19.2 19.7 20.3 20.3 21.3 23.9 24.6 26.7 7.5 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 21.3 9.0 26.0 22.3 19.4 16.9 15.3 14.7 14.1 :13.5 .13.9 14.2 14.C 14.6 16.0 16.2 17.6 10.5 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 14.8 12.0 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 12.6 13.5 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 10.9 15.0 17.6 15.0 12.8 10.9 9.9 9.4 9.0Q 8.6 8.8 9.0 8.6 8.6 9.2 9.1 9.5 16.5 16.4 13.9 11.8 i0.0 9.1 8.6 8.2 7.8 8.0 8.2 7.8 7.8 7.9 8.0 8.3 18.0 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 7.3 19.5 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 6.5 21.0 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 6.1 22.5 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 6.0 26 ANL/RERTE/TM-07-01, op. cit.

Page 4-43

CHAPTER 4: REACTOR 01/2012 CHAPTER 4: REACTOR

[I 01/2012 Critical Heat Flux Ratio 46.00 41.00 36.00 I

0 31.00 26.00 - I 21.00 16.00 I

11.00 6.00 1.00 4 1.50 4.00 6.50 9.00 11.50 14.00 16.50 19.00 21.50 Unit Cell Power (kW)

Figure 4.23, Critical Heat Flux Ratio (Bernath and Biasi Correlations)

Thermal hydraulic analysis using TRACE (section 4.6) demonstrates that a TRIGA fuel element operating 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, the fuel element that produces the most power (PPEAKROD) is related to the core average power level (PAVE) by:

PFi.AKROI) = PAVh..KPF Parametric variations including peaking factors from 1.3 to 2.0 and the number of fuel elements from 85 to 100 are provided in Table 4.19 and Fig. 4.23. With a peaking factor of 2 and 85 fuel elements, a core at 1913 kW would produce 45 kW in the element producing the highest power.

Table 4.19, Core Power, 45 kW Hot Element Peaking 85 90 100 Factor 1.3 2942 3115 3462 1.4 2732 2893 3214 1.5 2550 2700 3000 1.6 2391 2531 2813 1.7 2250 2382 2647 1.8 2125 2250 2500 1.9 2013 2132 2368 2 1913 2025 2250 Page 4-44

THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 4 0ET I 01/2012 Limiting Core Configuration (Fuel Element Peak Power 45 kW) 3400 3150 290) 2650 0

C-C) 2400 0

U.. 2150 4 -..- ,-+~

I 190c)4 1.3 1-37 1-44 1.51 1.58 1.65 1.72 179 1.86 1-93 2 Peaking Factor

- 85 ELEMENTS -g 90ELEMENTS 1--IOELEMENTS Figure 4.24, Core Power, 45 kW Hot Element Based on the calculations, 85 fuel elements with a peaking factor of less than 2.0 provides a large margin to thermal hydraulic limits.

4.6 Core Reactivity As noted in 4.5.1 (A), reactivity worth of material in the core is determined from differential measurements of calibrated control rod worth positions. Verification that the core configuration meets operating limits is similarly determined from the calibrated control rod positions.

As shown in Apoendix 4.1, the rapid fuel temperature response from a pulsed reactivity addition terminates the power increase and causes the reactor to stabilize at a power level corresponding to the fuel temperature consistent with Fig. 4.21. Therefore limits on reactivity are based not on the peak pulse power level, but on the final equilibrium power level associated with the reactivity. A polynomial equation calculating the reactivity deficit based on Fig. 4.24 with an R2 value of 0.99999 is: ý 6k = -1.75340-" 2 P 4 + 6.06670-10-9"P 3-8.777401 0-6"P2 +8.45380 3"P- 0.072937 An approximation of the power coefficient of reactivity from 100 kW to 1 MW is therefore:

d6k = -7.01360- 2 .P3 + 1.82001-10--P 2-1 .755488.10-6_P +8.45380-1 0-2 dP Page 4-45

CHAPTER 4: REACTOR 01/2012 Power Coefficient of Reactivity 0.00S -- . . . . --

OJA00 - .

1W0 200 3W 400 5WD f00 700 am0 9W0 1000 Power Level NkW)

Figure 4.25, Power Coefficient of Reactivity Therefore a pulse rod worth limited to 2.8% Ak/k ($4.00) will prevent exceeding steady state power 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 in the 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 pulse occurring 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 inadvertent pulsed reactivity addition leading to prompt critical condition.

There appears to be a significant difference in response in the power level coefficient comparing low power level data to high power level data; the prediction of the power coefficient of reactivity beyond the range of 1000 kW using a simple polynomial fit is not supported. Operating limits on core reactivity are provided in Table 4.20.

Table 4.20, Reactivity Limits

%Ak/k $

Excess reactivity 4.9 7.00 Shutdown margin[11 0.2 0.182 Moveable experiment worth 0.7 1.00 Single experiment worth 1.75 2.50 Total experiment worth 2.10 3.00 NOTE [1): most reactive rod fully withdrawn, moveable experiments in the most positive-reactive state Page 4-46

THE UNIVERSITY OF TEXAS TRIGA 11RESEARCH REACTOR

  • f 01/2012 SAFETY ANALYSIS REPORT, CHAPTER 4 I Based 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. When significant changes to the core configuration are made, verification that the core meets requirements is accomplished including evaluation that the control rod calibration is valid or re-establishing the control rod worth calibration.

Table 4.21, Limiting Core reactivity Control Rod Reference Current (2011)

Position Worth Position Worth Transient rod C ring $3.00 C-1 $3.10 Regulating rod C ring $3.71 C-7 $2.82 Shim 1 D ring $2.86 D-14 $2.52 Shim 2 D ring $2.86 D-6 $3.07 Total Rod Worth $12.43 $11.51 Critical Reactivity $5.43 $5.95 LIMITING CURRENT Excess Reactivity $7.00 $5.56 Shutdown Margin -$1.72 -$2.85 4.7 Thermal Hydraulic Design This section provides an independent assessment of the expected fuel and cladding thermal conditions, both steady-state and pulse-mode operations, with realiktic modeling of the fuel-cladding gap. Analysis is based on limiting conditions applied to 'a single fuel channel. The correlation of the limiting channel to core average power is sued to validate maximum permissible power level.

Analysis of pulsed-mode behavior is provided in Appendix 4.1, a commonly cited analysis of TRIGA fuel and cladding temperatures associated with pulsing operations. Analysis shows film boiling is not expected, even during~or, after pulsing leading tc*,maximum adiabatic fuel temperatures. The analysis addresses the case of a fuel element at-an-average temperature immediately following a pulse, then estimates cladding temperature and surface heat flux as a function of time after the pulse. The analysis predicts that, if there is no gap resistance between cladding and fuel, film boiling can occur very shortly after a pulse and cladding temperature can reach reaching 470'C. Mechanical stress to the cladding well is below the ultimate tensile strength of the stainless steel at these temperatures. Through comparisons with experimental results, the analysis concludes that an effective gap resistance of 450 Btu hr-1 ft-2 OF- (2550 W m-2 K-) is representative of standard TRIGA fuel and, with that gap resistance, film boiling is not expected.

Page 4-47

CHAPTER 4: REACTOR 01/2012 Analysis of steady state conditions reveals maximum heat fluxes remain well below the critical heat flux associated with departure from nucleate boiling. The heat transfer model is discussed, followed by the'results. .

4.7.1 Heat Transfer Model Heat generated in the fuel is conducted through the fuel matrix, transferred by convection across the gap between the fuel matrix and the cladding, conducted through the fuel cladding, and transferred by convection to the cooling water that flows through the core. Fuel centerline temperature can be calculated as the cooling water temperature increased by temperature changes through each physical element from the centerline of the fuel rod to the water coolant.

T,= T11 + ATb,+ AT + ATg + A T, Where 7 , is the fuel centerline temperature Th is the bulk water temperature AIT,r is the difference in temperature between bulk water and fuel cladding AT, is the difference in temperature across ATg is the difference in temperature across the gap betweenthe fuel andthe cladding AT,,, is the difference in temperature across the radius of the fuel A standard heat resistance model for this system is:

" h kc r, 4 zh

, 2 .k f Where q" is the heat flux through the cladding surface h is-the convective heat transfer coefficient associated with the cooling water ro and ri are cladding inner and outer radii kc is the cladding thermal conductivity hg is the gap conductivity kf is the fuel thermal conductivity Thermodynamic values are provided in Table 4.22, with the exception of the convective heat transfer coefficient associated with the cooling water. The gap conductivity of 2840 W m2 K-1 (500 Btu h-' ft -2 F-') is taken from Appendix A. General Atomics reports that fuel conductivity Page 4-48

THE UNIVERSITY OF TEXAS TRIGA 11RESEARCH REACTOR 0

  • fl 1 01/2012 SAFETY ANALYSIS REPORT, CHAPTER 4 1 over the range of interest has little temperature dependence. Cladding conductivity across the cladding is temperature dependent, with values quoted at 300 K,400 1(and 600 K.

Table 4.22: Thermodynamic Values Parameter Symbol Value Units Fuel conductivity kf 18 m- K __

14.9 W m 'K (300 K)

Clad conductivity kg 16.6 W m-4 K' (400 K) 19.8 W m1 Kz (600 K)

Gap resistance h9 2840 W m-2 K-1 Clad outer radius r0 0.018161 m Fuel outer radius ri 0.018669 m Active fuel length Lf 0.381 m Axial peaking factor APF 71/2 N/A The convective heat transfer coefficient is mode dependent and can only be determined in context; TRACE was used to provide heat transfer coefficient data supporting the analysis. The TRAC/RELAP Advanced Computational Engine (TRACE) code is the latest in a series of advanced, best-estimate reactor systems codes developed by the U.S. Nuclear Regulatory Commission for analyzing transient and steady-state neutronic-thermal-hydraulic behavior in light water reactors. It is the product of a long term effort to combine the capabilities of the NRC's four main systems codes 27 (TRAC-P, TRAC-B, RELAP5 and RAMONA) into one modernized computational tool.

The TRACE calculation models a unit cell composed of the area enclosed within a geometry unit defined by fuel pitch. Flow through the unit cell is modeled as a pipe, with model elements represented in Fig. 4.25. The UT TRIGA unit cell is an equilateral triangle, based on hexagonal geometry. Three 30' segments of a fuel element fall within the unit cell, with calculations for heat generation corresponding to 1/2 of the element. For example, calculations assuming 10 kW for the unit cell give indication of thermal response to an element output of 20 kW. The section of the fuel element that contains the fuel matrix (heated length) is modeled separately from the unheated lengths.

The active length of the fuel element was modeled as a TRACE heat source 15 in. (38.1 cm) in length, with the heat exiting through stainless steel cladding. Heat distribution was modeled as sinusoidal variation from a maximum at the center to a minimum modified at the end by extrapolation length of thermal neutrons in graphite. Data was calculated for 15 equally spaced nodes across the span of the simulated fuel element (i.e., 0.0127, 0.0381, 0.0635, 0.0889, 0.114, 0.140, 0.165, 0.191, 0.216, 0.241, 0.267, 0.292, 0.318, 0.343, and 0.368 m).

2' https://www.nrcsnap.com/snap/-luRins/trace/index.isp 01/2012 Page 4-49

CHAPTER 4: REACTOR 1 01/2012 Flow entrance and exit has a more complex geometry, and is not modeled explicitly. Special consideration given to the thermal hydraulic characteristics of the fuel end fittings that act as an interface between the flow channel and the grid plates, and the expansion or contraction of flow as it passes into/out of the flow channel. The characteristic losses associated with the entrance and exit includes turbulence effects from sudden expansion and contraction imposed by the grid plates as well as changes in flow direction. These losses are understood in terms of fractional values, or Kfactors. 28 The analytic expression of Kfor expansions/contractions is:

K=F1 l1 L A2 1 The flow path exiting the grid plate and entering the area below the cooling channel undergoes a 450 rotation, followed by another 450 rotation to direct flow along the cooling channel; the associated Kfactors are 0.3429 for each turn. The flow area expands suddenly at the entrance, with a Kfactor approximately 1.0. The loss factor for the cooling channel entrance is therefore 1.68. Similar calculations at the exit of the cooling channel yield a total loss factor of 1.18.

IN") 1.714-Figure 4.25: Unit Cell Fuel Element Model Hydrostatic pressure is required for TRACE calculations. Pressures at the inlet and outlet to the unit cell were calculated from nominal values of pool level, differential pressure from the confinement system HVAC, and local barometric pressure (Table 4.23). Normal pool level is 8.1 m, with a minimum of 6.5 m. (required by Technical Specifications). Normal pool temperature is about 20°C, with limiting temperature 48°C. Cooling flow enters the lower grid plate at 23.94 in. (0.608 m) above the pool floor and exits the upper grid plate at 51.00 in (1.2954 m). The reactor bay ventilation system operates at a slight vacuum (nominally 0.07 in. H20 or 17.44 Pa) reducing barometric pressure slightly. The average barometric pressure for the Austin area is 28 Handbook of Hydraulics, 5 h Ed. New York: McGraw Hill, King, H.W. and Brater, E. F. (1963) 29 http://www.westerndynamics.com/Download/friclossfittinzs.odf (01/2012)

Page 4-50

THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 4 000 0-0 D 11ETLI 01/2012 reported to range from 28.88 in. to 30.09 in., with an average of the high and low range at 29.485 in., or 99847.67 Pa.

Table 4.23, Hydrostatic Pressures Column Temp water Water Total water level height (m) (°C) density Pressure Hydrostatic (M) (kg/m 3) (Pa) Pressure Core top 8.1 6.805 20 998.23 66609.88 166440 Core bottom 8.1 7.492 20 998.23 73338.06 173168 Core top 6.5 5.205 48 988.56 50455.54 150285 Core bottom 6.5 5.892 48 988.56 57118.74 156949 4.7.2 Results The TRACE model was used to calculate temperatures at each of the 15 nodes. The heat flux and the temperature at the inner surface of the cladding were used to calculate the centerline fuel temperature. The temperature data (including TRACE data and centerline fuel temperature in is presented in Tables 4.24-4.27, and graphically (Fig. 4.26). Unit cell and total core flow was calculated (Table 4.28). Finally, recent observations of fuel temperature from installed instrumented fuel elements are compared to calculated fuel centerline temperatures in Table 4.29.

A. Water Temperature Limiting thermal hydraulic design is based on system response with -the maximum permitted pool water temperature and the "minimum allowed pool water level. TRACE calculations were performed for a range of unit cell power production. Table&4.24 provides the coolant temperatures calculated by TRACE at each node.

Table 4.24, Unit Cell Coolant Temperature (°C) for 490 C 6.5 m Pool Water Unit Ce:. Node (kW) 1 2 3 4 5 6. 7 8 9 10 11 .. 12 3.3 1.4, 15 1.5 , 50 50 51 52. .52 .52 53,. 53 53 53 54 .55 56 57 .57 3.0, 50 51 53 54, 54 ;55 55 55 55 ,.56 57 59 60 .61- 62 4.5 . 50 52. 54 56 56. 56'..57 57 57 58,60 rV62 .64 65 67 6.0 51 52 55 57., 57 r8 58 59 59 60 62 64 _64 68 70 7.5 51 53 55 58 59 59 60 60 61 61 64 '67 67 72 73 9.0 ... 51 54 56 59. 60 60. 61 62 62 63 66 69 69 74 77 10.5 51 54 57 60 61 62 62 63 64 64 68 71 72 77 79 12.0 52 55 58 61 62 63 64 64 65 66 70 73 74 80 82 13.5 52 55 59 62 63 64 65 65 66 67 71 75 77 82 85 15.0 52 55 59 63 64 65 66 67 67 68 73 77 79 84 87 16.5 52 56 60 64 65 66 67 68 -69 69 74 79 83 86 90 18.0 52 56 60 65* 66 67 68 69 70 71 76 80 85 89 92 Page 4151

CHAPTER 4: REACTOR 01/2012 Table 4.24, Unit Cell Coolant Temperature (°C) for 49°C 6.5 m Pool Water Unit Cell Node (kW) 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 19.5 53 57 61 66 67 68 69 70 71 72 '77 82 87 91 94 21.0. 53 57 ;61 66 67 68 69 70 71 72 78 83 87 91 95 22.5 53 ,57 61 '66 67 68 69 70 71 72 78 83 87 91 95 B. Fuel Temperature TRACE calculations provide cladding temperatures directly. Given the cladding temperatures, the standard heat resistance model previously identified in 4.7.1 can be simplified to:

claddnginner +q h 2 where heat flux and cladding temperatures are calculated in TRACE (Table 4.25 and 4.26). The gap and fuel physical dimensions and thermodynamic properties are constants; based on values in Table 4.22, the terms in parenthesis resolve to 5.39E-4 W m-2 k-1 . About 6% of the coefficient is determined by the gap, and any error associated with gap conductivity is minimized.

Table 4.25a, Outer Clad Temperature (°C) for 49°C and 6.5 m Pool Unit Cell Node kW 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 1.5 65 67 70 73 75 76 77 78 77 76 77 76 75 73 72 3.0 76 81 85 89 93 94 96 97 96 95 95 93 91 88 86 4.5 86 91 97 102 106 109 110 113 1I1 110 109 106 103 100 98 6.0 '94 100 -107 -114* 119 121 122 123 i22 122 121 118 114 Il 107 7.5 102 I09 116 123 124 124 125 125 125 124 124 123 122 120 116 9.0 109 117 123 125 125 125 126 126 126 125 125 125 124 123 122 10.5 116 123 "125 '126 126 126 126 127 'i26 126 126 126 125 125 124 12.0 122 125 126 126 127 127 127 127 127 127 127 126 126 125 125 13.5 124 126" 126 i27 127 127 128 110 127 127 127 127 126 126 125 15.0 125 126 127 127 128 128 128 128 128 128 128 127 127 126 126 16.5 126 127 127 128 128 128 128 129 128 128 128 128 127 127 126 18.0 127 127 128 128 129 129 129 129 129 129 128 ".128 128 127 127 19.5 127 128 128 129 129 129 129 129 129 129 129 128 128 127 127 21.0 127 128 128 129 129 129 130 130 129 129. 129 129 128 128 127 22.5 128 128 129 129 130 130 130 130 130 130 i29 129 129 128 127 Page 4-52

THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 01/2012 SAFETY ANALYSIS REPORT, CHAPTER 4 00 NETL Table 4.25b, Inner Clad Temperature (°C) for 49°C and 6.5 m Pool Unit Cell Node (kW) 1 2 3 4 5 6 7  :..8 9 .10 11 12 13 14 1.5 66 342 344 347 349 351 352 .353 352 351 351 350 349 348 346 3.0 79 356 360 365 369 370 372 374 .372 371 371 368 366 364 361 4.5 89 368 373 379 384 386 388 391 389 387 386 383 380 377 374 6.0 98 378 385 392 397 400 401 402 402 400 400 396 392 388 384 7.5 107 388 395 402 404 405 405 406 405 405 404 403 401 398 394 9.0 115 397 403 405 407 407 408...408 408- 407 407 406 404 403 401 10.5 122 404 405 408 409 410 410. 411 410 409 40S 408 406 405 404; 12.0 130 406 110 410 411 412 412 413 412 411 411 410 408 407 405 13.5 133 408 410 412 413 414 414 415 414 413 413 411 410 408 407 15.0 135 410 412 413. 415 415 416 417 416 415 415 413 412 410 408 16.5 136 411 413 415 417 417 418 419 418 417 416 415 413 411 410 18.0 138 413 415 417 418 419 420 421 420 419 418 416 415 413 411 19.5 139 414 416 418 420 421 422 422 422 421 420 418 416 414 412 21.0 140 416 418 420 422 423 423 424 423 422 422 420 418 415 413 22.5 142 417 419 421 423 424 425 426 425 424 423 421 '419 417 415 Tabie 4.26a, Heat Flux (Nodes l-18) 49°C 6.5 Pool, Unit Cell Node (kW) 1 2 3 4 5 .. 6 7 8 1.5 -2.72E4 -3.06E4 -3.40E4 -3.74E4 -4.08E4 -4.23E4 -4.38E4 -4.53E4 3.0 -5.44E4 -6.12E4 -6.80E4 -7.48E4 .-8.16E4 -8.46E4 -8.76E4 -9.06E4 4.5 -8.16E4 -9.18E4 -1.02E5 -1.12E5 -1.22E5 -1.27E5 -1.31E5 -1.36E5 6.0 -1.09E5 -1.22E5 -i.36E5 -1.50E5 -1.63E5 -1.69E5 -1.75E5 -1.81E5 7.5 -1.36E5 -1.53E5 -1.70E5 -1.87E5 -2.04E5, -2.11E5 -2.19E5 -2.27E5 9.0 -1.63E5 -1.84E5 -2.04E5 -2.24E5 -2.45E5 -2.54E5 -.2.63E5 -2.72E5 10.5 -1.90E5 -2.14E5 -2.38E5 -2.62E5 -2.86E5 -2.96E5 -3.06E5 -3.17E5 12.0 -2.18E5 -2,45E5 -2.72E5 -2.99E5 -3.26E5 -3,38E5 _-:3.50E5 -3.63E5 13.5 -2.45E5 -2,75E5 -3.06E5 -3.37E5 -3,67E5 -3.81E5 -3.94E5 -4.08E5 15.0 -2.72E5 -3.06E5 -3.40E5 -3.74E5 -4.08E5 -4.23E5 .- 4.38E5 -4.53E5 16.5 -2.99E5 -3.37E5 -3.74E5 -4.11E5 -4.49E5 -4.65E5 -4.82E5 ý4.99E5 18.0 -3.26E5 -3.67E5 -4.08E5 -4.49E5 -4.89E5 . -5.07E5 -5.25E5 -5.44E5 19.5 -3.54E5 -3.98E5 -4.42E5 -4.86E5 -5.30E5 -5.50E5 -5.69E5 -5.89E5 21.0 -3.81E5 -4.28E5 -4.76E5 -5.23E5 -5.71E5 -5.92E5 -6.13E5 -6.34E5 22.5 -4.08E5: -4.59E5 -5.10E5 -5.61E5 -6.12E5 -6.34E5 -6.57E5 -6.80E5 Page 4-53

CHAPTER 4: REACTOR I 01/2012 Table 4.26b, Heat Flux (Nodes 8-15) 49°C 6.5 Pool, Unit Cell Node (kW)I 9 10 11 12 13 14 15 1.5 -4.38E4 -4.23E4 -4.08E4 -3.74E4 -3.40E4 -3.06E4 -2.72E4 3.0 -8.76E4 -8.46E4 -8.16E4 -7.48E4 -6.80E4 -6.12E4 -5.44E4 4.5 -1.31E5 -1.27E5 -1.22E5 -1.12E5 -1.02E5 -9.18E4 -8.16E4 6.0 -1.75E5 -1.69E5 -1.63E5 -1.50E5 -1.36E5 -1.22E5 -1.09E5 7.5 -2.19E5 -2.11E5 -2.04E5 -1.87E5 -1.70E5 -1.53E5 -1.36E5 9.0 -2.63E5 -2.54E5 -2.45E5 -2.24E5 -2.04E5 -1.84E5 -1.63E5 10.5 -3.06E5 -2.96E5 -2.86E5 -2.62E5 -2.38E5 -2.14E5 -1.90E5 12.0 -3.50E5 -3.38E5 -3.26E5 -2.99E5 -2.72E5 -2.45E5 -2.18E5 13.5 -3.94E5 -3.81E5 -3.67E5 -3.37E5 -3.06E5 -2.75E5 -2.45E5 15.0 -4.38E5 -4.23E5 -4.08E5 -3.74E5 -3.40E5 -3.06E5 -2.72E5 16.5 -4.82E5 -4.65E5 -4.49E5 -4.11E5 -3.74E5 -3.37E5 -2.99E5 18.0 -5.25E5 -5.07E5 -4.89E5 -4.49E5 -4.08E5 -3.67E5 -3.26E5 19.5 -5.69E5 -5.50E5 -5.30E5 -4.86E5 -4.42E5 -3.98E5 -3.54E5 21.0 -6.13E5 -5.92E5 -5.71E5 -5.23E5 -4.76E5 -4.28E5 -3.81E5 22.5 -6.57E5 -6.34E5 -6.12E5 -5.61E5 -5.10E5 -4.59E5 -4.08E5 Calculation of maximum fuel temperature was performed using the standard heat resistance model (Table 4.27) modified to use the TRACE data as described above.

Table 4.27, Fuel Centerline Temperatures (°C)

Unit Cell Node (kW) 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 1.5 90 95 101 107 112 115 117 119 117 115 114 110 106 102 97 3.0 126 137 147 158 167 172 176 180 176 173 170 161 153 145 136 4.5 161 175 190 205 218 225 231 237 231 226 221 209 197 185 172 6.0 194 212 232 251 268 276 283 289 283 276 270 255 239 222 207 7.5 226 249 272 294 311 318 325 333 325 318 311 295 278 260 241 9.0 258 286 310 330 350 358 367 375 367 358 350 330 311 292 272 10.5 290 319 343 366 388 398 407 417 407 397 388 366 343 321 298 12.0 322 349 375 400 425 437 448 460 448 436 425 400 375 350 325 13.5 349 378 407 436 464 477 489 502 489 476 463 436 407 378 350 15.0 375 407 438 470 502 515 529 S43 529 515 501 470 438 407 375 16.5 400 436 470 504 540 554 570 586 570 554 539 504 470 436 400 18.0 425 464 502 540 577 593 610 627 610 593 576 539 501 463 425 19.5 451 492 533 574 615 633 650 669 650 633 614 574 533 492 451 21.0 476 520 565 608 652 672 691 710 691 672 652 608 564 520 476 22.5 502 549 596 643 690 711 732 753 732 710 690 643 596 549 501 Page 4-54

THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 4 Mt I 01/2012 C. Temperature Profiles and Flow Rates TRACE calculation provides thermal response for a unit cell. Temperature calculations are based on heat flux (heat per area) and are therefore valid for both the unit cell and the fuel element. However, Fig. 4.25 shows a unit cell to be Y2 of a fuel rod so that the total power generated in a fuel element is twice the unit cell value.

Fuel, cladding, and coolant temperatures based on TRACE data in Tables 4.24, 4.25, and 4.27 are provided in Fig. 4.26 for two unit cell power levels, 10.5 and 22.5 kW. Flow rate versus power for a single fuel element is provided in Fig. 4.25.

Total core flow is the sum of the flow rates of individual fuel rods operating at specific power levels. The power level an element generates is determined by the peaking factors associated the position in the core. Instrumented fuel elements (modified to accept thermocouples) have slightly less fuel mass than other standard fuel elements. Fuel followers have a smaller radius than standard fuel elements. Therefore the power production in thermocouple elements and fuel followers is less, approximately by the ratio of the mass of the element to the mass of a standard fuel element. Table 4.28 provides the data and calculation of total core flow based on a 116 element core operating at 1100 kW. Similar calculations were performed for a 120, 100 and 85 element core over a range of power level with results in Fig. 4.27.

10.5 kW Unit Cell Axial Temperature Profile 675 625 I 575 --

525 475 -7 M

45...... ...............................

4- 425 - -------

375 325 .-.. .7 275 1 3 5 11 13 15 Node

- - Fuel Temp --- Inner Clad Temp - Outer Clad Temp - -Coolant Temp Figure 4.26a, Unit Cell Temperature Distribution (10.5 kW)

Page 4-55

CHAPTER 4: REACTOR CHAPTER 4: REACTOR I 01/2012 01/2012 22.5 kW Unit Cell Axial Temperature Profile 975 875 775 675 575 E

9 475 375 275 1 3 5 7 9 11 13 15

- Fuel Temp .Inner Cad Temp -CouterClad Temp - -CoacantTemnp Node Figure 4.26, Unit Cell Temperature Distribution Flow rate versus Unit Cell Power 0.13 0-12 0.11 0.10 009 0.08 0.07 0.36 0.05 0.0]4 - i  ; .-- --. . .. . . .. .

0~3 yl 1.02541EO5x5 -4.390E-O4xz + I.412BE-O3x+3.0123E402 0.02 .992S-01"-

0.00 0 2 4 6 8 10 12 14 16 18 20 22 24 Unit Cd power (W, Figure 4.27, Single Rod Flow Cooling Flow Rate versus Power Level 49°C 6.5 Pool Page 4-56

THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 4 0 0 I 01/2012 Table 4.28, Coolant Flow for 1100 kW Operation no. peaking Power (kW) Flow Rate (kg/s ring elements factor Unit Cell Element RING Unit Cell Element RING B (6) 4 1.57 7.44 14.89 59.55 0.07 0.15 0.58 B IFE 2 1.57 7.31 14.61 29.23 0.07 0.14 0.29 C (12) 10 1.46 6.92 13.84 138.45 0.07 0.14 1.41 C FFCR 2 1.46 5.78 11.56 23.11 0.07 0.13 0.26 D (18) 16 1.29 6.12 12.23 195.72 0.07 0.14 2.16 D FFCR 2 1.29 5.11 10.21 20.42 0.06 0.13 0.25 E (24) 24 1.07 5.07 10.15 243.52 0.06 0.13 3.02 F (30) 28 0.81 3.84 7.68 215.07 0.06 0.11 3.17 G (30) 28 0.66 3.13 6.26 175.24 0.05 0.10 2.94 TOTAL: 116 TOTAL: 1100 TOTAL: 14.08 D. Comparison to Operational data During calendar year 2011 the UT TRIGA core consisted of 109 standard fuel elements, 2 instrumented fuel elements and 3 fuel followers. Operational data was collected (Table 4.29) to compare calculated fuel temperature values for specific operations with observed indications. Each of the selected values follows an operating interval that approaches steady state fuel temperatures, except for *the 10 kW values (there was a series of 10 kW operations for less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> during 2011). The power produced by the individual fuel element is calculated as total core power divided by the number of fuel elements (114) and multiplied by the nominal B ring peaking factor (rt/2).

Instrumented fuel elements contain three thermocouples, one at the axial ',midplane with the remaining thermocouples offset 1 inch above and below. Only one thermocouple in each IFE is in use. Fig. 4.24 shows approximately'20 'C difference between the center of the element and positions approximately 1 in. from the center at 10.5 kW unit 6ell power. The, core power is not expected to be homogenous in the B ring; consequently the peaking factors a*re not expected to be uniform for all B ring elements. The position of an IFE and the position of an individual thermocouple within IFE may affect temperature indication; the agreement between FT1 and FT2 measuring channels is therefore considered remarkably good.

Table 4.29, Observed Fuei Temperatures Date Power (kW) Observed Temperatures (°C)

Core B Ring IFE FT1 FT2 Pool 10/6/2011 10 0.13 26 28 21.1 12/21/2011 100 1.31 86 97 22.8 12/20/2011 500 6.54 240 261 23 12/16/2011 950 12.44 340 359 21.9 Page 4-57

CHAPTER 4: REACTOR I 01/2012 TRACE calculations were performed at power levels corresponding to 100, 500 and 950 kW with pool water temperatures corresponding to the values recorded in Table 4.29. The maximum fuel temperatures from the TRACE calculation are provided in Table 4.30 along with the observed values for comparison.

Table 4.30, Fuel Temperature Comparison Power (kW) Fuel Temperatures (°C)

Core B Ring IFE TRACE FT1 FT2

.100 1.3 95 86 97 500 6.5 289 240 261 950 12.4 446 340 359 Fuel Element Temperature Versus Power

-'-CALCJLATED -o-FT I VALUES -FT 2 VALUES 500 450 400 350 3O0 E

a E

250 200 t

U.- 150 100 ...... 4-41 50 0

0 1 2 3 4 5 6 7 8 9 10 11 12 13 Fuel Element Power (kVNI)

Figure 4.28, Comparison of Calculated and Observed Fuel Temperatures The information (provided graphically in Fig. 4.28) shows the temperature from TRACE calculations to be reasonably close and consistently higher than observed values, with the deviation increasing as power level increases. Therefore, modeling bounds actual conditions.

Page 4-58

THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 9 ..

00o0 0

I 01/2012 9.0 AUXILIARY SYSTEMS 9.1 Confinement System The design of a structure to contain the TRIGA reactor depends on the protection requirements for the fuel elements and the control of exposures to radioactive materials. Fuel elements and other special nuclear materials are protected by physical confinement and surveillance.

The floor of the reactor bay is approximately The lower walls of the reactor bay are cast in place concrete. Above grade, the walls are reinforced precast concrete tilt panels, approximately with integral columns and embedded reinforcing steel. The wall panels were then set in place vertically using a crane with space left in between each panel for a structural column and temporarily braced. Next the column forms were placed around reinforcing steel extending from the edges of the panels which was interlaced with additional steel reinforcing internal 'to the columns. Concrete was then poured into these forms resulting in a finished wall system with columns that resemble a poured in place design rather than the typical tilt panel welded design.

The roof is sealed using standard tar and gravel techniques. All penetrations in the reactor bay confinement envelope are on the south side, interfacint with the reactor wing offices, machine room spaces, equipment staging area, and confinement (and auxiliary purge) ventilation system..'.

9.2 HVAC (Normal Operations)

Building environment controls use air handling units for ventilation and comfort with cold and hot water coils for temperature and humidity control. There are two separate HVAC systems with three air handling units, located on the fourth level of the reactor bay wing adjacent to the reactor bay. One unit contains both cold and hot water coils in a single duct.system, dedicated to the reactor bay. This system supports confinement functions. The other two units are the cold- and hot-deck components of a double duct system that conditions air in all building zones other than: the reactor bay. A fume/sorting hood is installed in the reactor bay, using a separate exhaust fan and isolation damper that disc"h'arges into a separate roof stack.

Water temperatures of the heating and cooling coils in the air handling units are controlled by set of on-site and off-site systems. The heating system is an ori-site boiler unit with a design capacity set by local building (HVAC) requirements. The cooling system is a PRC chilled water treatment plant with design ýcapacity Set by overall research campus requirements; with thermostats controlling zone or room temperatures. A local instrument air system provides control air for HVAC systems. Controls and air balancing of the two air handling systems provide user comfort and pressure differentials between the reactor bay (confinement) and adjacent zones, and between the adjacent zones and the academic wing of the building.

Page 9-1

CHAPTER 9, AUXILIARY SYSTEMS 1 01/2012 The ventilation system is designed to maintain a series of negative pressure gradients with respect to the building exterior and other building areas, with the reactor bay (confinement) at the lowest pressure. Confinement functions of ventilation control the buildup of radioactive materials generated as a byproduct of reactor operations, and isolate the reactor bay in the event that an abnormal 'release is'detected in the reactor areas. Confinement and isolation is achieved by air control;dampers and leakage prevention material at doors and other room penetration points.

A conceptual diagram of the system is provided in Fig. 9.1. Manual operation controls for both main and purge air systems are in the reactor control room.

wow Figure 9.1, Conceptual Diagram of the Reactor Bay HVAC System An exhaust stack on the roof combines the ventilation exhausts from both the main and the purge systems. As illustratedJin Fig. 9.1, the auxiliary purge system discharge is within the HVAC exhaust stack.. The auxiliarypurge exhaust is a 6 in. (15.24 cm) internal ID and 8.63 in. (21.92 cm) OD. The HVAC exhaust has an 18 in. (45.72 cm).

9.2.1 Design basis Confinement system ventilation has three modes of operation, reactor run mode, quiescent mode, and confinement isolation. The design goal for HVAC system in the reactor run mode is to control the reactor bay, adjacent zones and academic wing of the building at a negative pressure difference relative to ambient atmospheric pressure during 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 during routine operations is discharged through the stack and does not build up in the reactor bay. Release of airborne radioactivity consists mostly of activated 4 1Ar from routine operation. The design goal of the confinement Page 9-2

THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 00 B0if 01/2012 SAFETY ANALYSIS REPORT, CHAPTER 9 system ventilation during quiescent mode is to minim; e energy utilization during periods when the reactor is not operated while maintaining pressure in the reactor bay 0.06 in. below atmospheric pressure. The reactor-room confinement is 4esigned to control the exposure of operation personnel and the public from radioact.ive material or its release caused by reactor operation. During potential accident conditions, sensors initiate confinement system isolation when high levels of radioactivity are detected in reactor bay air, e.g. If a fuel element. failure releases fission products or if an experiment with sufficient inventory of radioactive material fails. The confinement isolation secures fans and dampers in the. confinement HVAC, fume/sorting hood, and auxiliary purge 1system. Provisions are made to allow subsequent operation of the auxiliary purge system with the remaining HVAC confinement in isolation.

Release criterion is based on Title 10 Chapter 20 of the U.S. Code of Federal Regulations.

9.2.2 System description During operating modes supply fans draw air from either the return fan or the environment into a conditioning unit that subcools the air:,to control humidity then heats the air for habitability/comfort. Air filtration is the typical design for normal. HVAC operation with fiberglass roughing filters only. The ccnfinement systemn uses heating and cooling in a single unit, the remainder of the building HVAC system has air conditioning split into separate hot and cold decks.

COMM-~

SLPPLY AIR

-~ ~~

FILTER a eJ~EA1 ~ 1.Ai ISCMfIC Figire 9.2A, Ma'n Reactor Bay HVAC System Page 9-3

CHAPTER 9, AUXILIARY SYSTEMS 01/2012 Table 9.1, Typical Confinement Vent & Purge Parameters Duct Velocity Exit Velocity Aux Purge 3900 fpm 20 m/s 35.23 m/s Confinement Vent 1800 fpm 9 m/s 26.87 m/s Flow Rate Aux Purge 1100 cfm 0.52 m3/s Confinement Vent 7200 cfm 3.40 m3/s PAM AIRWO AMIUET *l.T*

  • uq.Y S3 m IA"M ISMATM (P. LWIII

~f. HVAC OPERAT ION /*ODES

-W 6HI. FNG CM all R.P. MATISO-D'A= AM FZCE=

  • amAM plýcadvs FMOF Fiue92,Mi- RatrByHA fify SyteCnto Contro fro"W=M .U~I Figure 9.21B, Main Reactor Bay HVAC Control System Control 9.2.3 Operational analysis and safety function Speed of the confinement system supply fan is regulated to produce 0.06 in. water vacuum in the reactor bay by differential pressure control between the reactor bay and a representative ambient external building measurement point. Additional measurement points in ventilation zones adjacent to the reactor bay are used to maintain differential pressure between the reactor bay and adjacent access areas. Supply air is distributed through a rectangular duct near the ceiling and then to distributed ducts and vents running down the wall and ending near the floor), enhancing mixing and preventing stratification. Air is discharged from the bay through 4 return grills, two parallel ducts to grills near the floor, and two grills near the ceiling. In the reactor run mode the confinement system exhaust fan is controlled to maintain stack velocity designed to exceed the minimum air change specification. Control dampers are located at the supply fan inlet (fresh air intake) and the exhaust fan outlet (discharge to stack), and in a line between the supply and return fans. Confinement system ventilation discharge is through a Page 9-4

iS THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR so e REI't 01/2012 SAFETY ANALYSIS REPORT, CHAPTER 9 I

stack on the reactor building roof. Schematics of the ventilation system for the reactor bay area and a logic diagram of the ventilation control system sensors and controls are provided in Fig. 9.2A and B.

Table 9.2, Reactor Venti'Eation System Modes MODE SYTMCMOETREACTOR RUN QUIESCENT ISOLATION control damper CLOSED OPEN CLOSED supply & exhaust OPEN MINIMUM MINIMUM control dampers Confinement HVAC Controlled for Stack Supply Fan velocity Constant Speed OFF Exhaust fan Controlled for bay Controlled for bay OFF dp dp supply & exhaust OPEN CLOSED(i CLOSED control dampers OPEN[21 Auxiliary Purge System OFF Exhaust fan ON OFF[1] 121 OFF or ON supply & exhaust OPEN or CLOSED' 3 1 - CLOSED"' CLOSED Fume/Sorting Hood control dampers Exhaust fan ON or OFF[31' OFF111 OFF NOTE [1]: Mode is set manually NOTE [2]: Provisionshave been made to permit operationof auxiliarypurge system in conferment isolation NOTE [3]: Fume hood is operated manually, as required,and not correlatedto reactoroperation When the reactor is operating (reactor run mode) the system is operated to generate a rate of air exchange exceeding 2 air volumes (4120 M3) per hour. maintain a controlled stack velocity, and regulate negative pressure in the reactor bay. In the reactor run mode, the confinement HVAC supply fan is controlled to maintain the reactor bay at nominal minimum 0.06 in. water.

In the quiescent mode, the confinement ventilation system is balanced for recirculation flow with a small amount of effluent. When the reactor is not operating (quiescent mode), the ventilation system is operated to minimize requirements for conditioning incoming air, in a recirculation mode with a minimal exhaust flow rate and fresh air intake as required to maintain a negative pressure in the reactor bay with respect to adjoining spaces.

In the confinement isolation mode the confinement HVAC and the reactor bay fume/sorting hood are secured; the auxiliary purge system is secured when isolation occurs, but may be manually configured to operate. In the event that airborne radioactive material exceeding a trip set point is detected, the system is designed to establish a shutdown and isolated Page 9-5

CHAPTER 9, AUXIUARY SYSTEMS 01/2012 condition. Separate controls allow the confinement HVAC and the reactor bay fume/sorting hood to be isolated while the auxiliary purge system can be operated.

Atmospheric dispersion' using a'st4ck, model reqtires:stack discharge 60 (18.23 m) feet above the ground, and at least :2 and 1A times the height of adjacent structures. The nearest structure ismapproximately 80 meter's from the" reactor bay'. Ground elevation in the area is 794 feet, with roof elevation at the stack 843 feet, a distance of 49 feet (14.94 m) above grade. The exhaust stack extends:14 feet (4.24 ýmeters) above the roof level so that the stack discharge is 63 feet (19.202inm). The ýeffective release point above the exhaust'stack can be calculated from the Bryan - Davidson *equation: -

1.4 (VS.

Where:

Ah is the height of plume rise above release point (m)i 2 2 D is the diameter of stack (m), confinement vent 0.40122m , auxiliary purge 0.152 m

/7is the mean wind speed at stack heght (m/s)

V, is the effluent vertical eff!ux velocity (m/s), confinement vent 26.87 m/s, purge 35.23 m/s The effective stack height for the reactor HVAC confinement vent system (in units of meters) is therefore 40.19/{wind velocityl. rn above the stack, and the effective stack height for the auxiliary purge system is 22.25/{wind velocityl above the top of the:stack at 63 feet (19.202 m).

Mixing of the two effluent streams occurs at the exit of the stack.

Pneumatically operated isolation dampers in the confinement system ventilation are located at the supply fan outlet (supply-to 'he reactor bay) and the exhaust fan inlets (return from the reactor bay) near the reactor bay wall penetrations as indicated in Fig. 9.1, as* weli as the fume/soring hood in the reactor bay auxiliary purge system. Controls close the dampers and securethe fans in response to manual or automatic signal initiated by high airborne particulate radioactivity. Loss of instrument air or loss of control power will cause the'dampers to isolate the reactor bay.

9.2.4 Instruments and controls As indicated, the HAVC control, system is controlled' by a, set of temperature, flow, and differential pressure sensors that develop control signals. The signals are used in variable frequency controllers that regulate fan speed to maintain pressure and temperature.

Control room switches establish the operating mode of the confinement ventilation system.

The auxiliary purge system is controlled from the same panel.

Page 9-6

THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR OF, 01/2012 SAFETY ANALYSIS REPORT, CHAPTER 9 e Mt Figure 9.3, Confinement System Ventilation Controls Confinement HVAC mode is controlled by a toggle switch labeled "Reactor Off/ Reactor On."

Reactor Off establishes quiescent mode described above. Reactor On establishes the reactor run mode described above.

Alarm indicators on tie control panel provide indication that the diffarential pressures are normal or abnormal. Flow and differential pressure indicators inside the panel provide indication of the zone static pressure, and confinement system and auxiliary purge system velocities.

A continuous air particulate detector located in the reactor bay provides a control signal to initiate confinement isolation when the count rate exceeds a preset level. Indicators at the reactor control console provide alarm level information. A count rate associated with 2,000 pCi/ml detects particulate activity at occupational levels of 10CFR20. The alarm setpoint exceeds 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 the reference concentrations within two hours.

Page 9-7

CHAPTER 9, AUXILIARY SYSTEMS, 01/2012 9.2.5 Technical Specifications, bases, testing and surveillances Either the confinement ventilation system or the auxiliary purge system is required to be operating when the reactor is operating to control the buildup of gaseous radioactive material in the reactor bay. Ifthe, confinement, ventilation system is operating, instrumentation to initiate confinement .isolation on high airborne contamination levels will be operable. The confinement system.will be: checked periodically to assure proper function. The particulate monitor will be calibrated periodically.

9.3. Auxiliary Purge System A separate, low volume air purge system is designed to exhaust' air that may contain radionuclide products from strategic locations in the reactor bay.'

9.3.1 Design basis The purge system collects and exhausts air from potential sources of neutron activation such as beam tubes, sample transfer systems, rotary specimen rack, and material evolving from the surface of the pool. The purge system filters air in the system. through a rough prefilters followed by a high efficiency particulate: filter. .;Design provisions allow for the addition of charcoal filters if experiment conditions or.other' situations should require the additional protection.. -- .I 9.3.2 System description:

Mai AIR T " * ' . . .' .. .* ' "LP;i avr *u , z ARII PLO WI- I I 93---1 FILTER Li-

- q ....

ARW RPUE 2 02,.97;1 KEA FILTER E)O4AL5T F74"4 Eg7.3J 67ACX 3 FUT1JM OWJAL FILTER CA I-fO

, . MIN. M*N

  • P. ILzmi I~.ATrI Tap~i VALVE 2 4

---0 -*V

.U**I D--AIXIM 'AIR ISO"

  • N' I , ..

... WI. SIM e "Z L ON REACTORBSAYH Figure 9.4A, Purge Air System _ Figure 9.4B, Purge Air Controls I Page 9-8

THE UNIVERSITY OF TEXAS TRIGA 11RESEARCH REACTOR 0o0o 00 01/2012 00 SAFETY ANALYSIS REPORT, CHAPTER 9 0 9.3.3 Operational Analysis and Safety Function The primary nuclide of interest is argon-41. Fig. 9.4A and .9.4B are schematics of the auxiliary purge system and its control logic.. Sample ports in the turbulent flow stream of the purge system exhaust .provide for measurement of exhaust activities. The isolationi damper in the purge system is actuated manually, using the.fan control switch. Autornatic iso!atior, o1 the system is generated by the same particulate radiation mcnitcr as is used by.the HVAC confinement ventilation system.

Purge flow is nominally adjusted for continuous operations with approximately 525 cfm from the pool and a similar dilution flow rate from the reactor bay environment. The dilution flow controls effluent humidity from the reactor pool area to limit possible degradation of the purge system HEPA filters. A purge flow of approximately 4 cfm is drawn from the beam port interior when a beam port is used. The beam guide prevents closure of the outer shutter door, and beam port three is normally purged. The rotary specimen rack is purged prior to loading or unloading for about 10 minutes to control personnel exposure and also to remove hydrogen that may evolve from the polyethylene capsules during irradiation. -

The auxiliary purge system may be operated with the confirement HVAC system secured. Since the confinement HVAC operates continucusly except during isolation, confinement HVAC can be secured using thei HVAC Contro!, toggle switch (inside the HVAC control panel, described previously). Since the auxiliary purge system is equipped with HEPA filters and :has the capability for using charcoal filters, operation of the auxiliary purge system could reduce elevated airborne radionuclide contamination in the reactor bay and contain a large fraction of the radionuclides in filtration.. Qperation in this mode requires that the confinement HVAC be secured to prevent unfiltered releases, and then bypassing the confinement isolation trip signal.

9.3.4 Instruments and controls The auxiliary purge system is :controlled from the same panel as the confinement ventilation system. Toggle switches on the controlroormiconfinement HVAC cdhtr6lp3nbl. open dampers to allow the pool surface purge flow, 'and-independently flow from a manual valve manifold accessible on the ground level of the rector bay. The manual valve manifold controls purge flow from the experiment facilities. A separate manually operated: valve' in the same area controls the amount of dilution flow to the purge system.

A flow gage indicates purge stack velocity at the panel. The exhaust point is concentric to the center of the HVAC confinement ventilation exhaust stack.

The auxiliary purge system is monitored by a gaseous effluent radiation detector. The effluent monitor has an alarm setpoint based on ten times the occupational limit or a reference concentration at the ground.

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CHAPTER 9, AUXILIARY SYSTEMS 9.3.5 Technical Specifications, bases, testing and surveillances Ifthe auxiliary purge system is operating, a gaseous effluent monitor will be operating. The auxiliary purge system will have a high efficiency particulate filter. Auxiliary air purge system valve alignment will be checked periodically. The gaseous effluent monitor will be calibrated periodically.

9.4 Fuel storage and handling Special provisions are necessary for the storage of fuel elements that are not in the core assembly. 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 are specially designed racks inside the reactor pool and outside the reactor shield.

Irradiated fuel is manipulated remotely, using a standard TRIGA fuel tool. Irradiated fuel is transferred out of the pool using a transfer cask modeled on the BMI cask TRIGA basket. There are two different loading templates for use with the transfer cask, permitting loading operation either for a single TRIGA fuel element, or to up to three elements. A 5-ton overhead crane is used to move the fuel transfer cask.

9.4.1 Design basis Stored fuel elements are required to have an effective multiplication factor of less than 0.8 for all conditions of moderation. Fuel handling systems and equipment are designed to allow remote operation of irradiated fuel, thus minimizing personnel exposure.

9.4.2 System description..

Space inside the reactor pool is adequate for a large number of fuel racks. The racks are aluminum, suspended from the pool edge by connecting rods.

To facilitate extra storage, 2 racks may be attached to the same connecting rods by'locating one rack at a different vertical level and offsetting the horizo0tal position.. slightly.

,Outside the react-or pool, rack design is intended to fit in,special storage wells (Fig. 9.4).

Water may be added for shielding or cooling.

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THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 9

'.ETII 01/2012 1

Most routine fuel storage is intended to be inside the reactor pool. Outside the reactor pool, supplemental fuel storage is planned for temporary storage of elements transferred to or from the facility, for isolation of fuel elements with clad damage, emergency storage of elements from the reactor pool and core assembly and routine storage o" other radioactive materials.

Temporary storage for some reactor components or experiments may also use the fuel storage Page 9-11

CHAPTER 9, AUXILIARY SYSTEMS 01/2012 racks in the reactor pool. Other locations not in the pool will also provide storage for radioactive non-fuel materials.

A fuel transfer cask modeled after the BMI cask TRIGA basket is used to transfer irradiated fuel.

A standard TRIGA fuel handling tool is used to remotely grapple irradiated fuel elements.

A 5-ton crane is used in conjunction with the fuel handling tool and the transfer cask to allow remote handling of irradiated fuel.

9.4.3 Operational analysis and safety function Bench mark experiments conducted by TRIGA International indicate minimum mass for criticality requires 64 fuel elements in a favorable geometry.

Pool storage racks do not have the capacity or the geometry to support criticality. Spent fuel storage has a higher fuel density in storage, but does not have the capacity to hold 64 fuel elements, and does not have favorable geometry.

The fuel handling tool has been used successfully at the UT TRIGA reactor, including the original reactor on the main campus as well as the current installation. This design is widely used by TRIGA reactors, with good performance history although the first generation tool occasionally released 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 exists in a bottom loading cask. The cask does not provide adequate shielding for close-in work, and all handling is conducted remotely.

The crane exceeds load requirements for spent fuel handling by a large margin. There is little potential for failure.

9.4.4 Instruments and controls New fuel storage is in a locked room on the middle level of the reactor bay. A criticality monitor is installed, with neutron and gamma channels. The system has a local indicator directly outside the storage room, and a remote readout in the control room.

9.4.5 Technical Specifications, bases, testing and surveillances Fuel elements are required to be stored in a configuration with keff less than 0.8. Irradiated fuel is required to be stored in a configuration where convective cooling by water or air is adequate to maintain temperature below the safety limit.

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0 THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 9 El Tt r, 01/2012 9.5 Fire protection systems Active fire protection elements generally have automatic operation, manual response, or personnel action for the intended function. Active elements to be considered include automatic fire detection, automatic fire suppression in labs and office spaces, fire information transmission, manual fire suppression and other manual fire control.

Passive fire protection provides fire safety that does not require physical operation or personal response to achieve the intended function. Passive elements include inherent design features, building physical layout, safety-related systems layout, fire barriers, and construction or component materials, and drainage for control of fire protection runoff water. Penetrations in fire barriers have fire resistant ratings compatible with the purpose of the fire barrier.

9.5.1 Design basis The goal of fire protection is to provide reasonable assurance that safety-related systems perform as intended and that other defined loss criteria are met1' 2. For the purpose of fire protection, loss criteria should include protection of safety-related systems, prevention of radioactive releases, personnel protection, minimization of property damage, and maintenance of operation continuity. Three components shall be applied to the fire protection objective. The three components are passive and active fire protection, and fire prevention.

A fire detection, suppression, and information management system is designed to ensure that fires can be detected, suppressed (where possible), and alert response organizations.

Basic design features of the reactor assembly, pool and shield system, and the instrumentation, control, and safety system represent passive fire protection elements. These basic features are sufficient passive protection to protect safety-related systems.

9.5.2 System description Manual protection consists of manual firefighting actions and the systems necessary to support those actions such as extinguishers, pumps, valves, hoses, and the inspection, maintenance and testing of equipment to assure reliability and proper operation. Other manual actions that are elements of active fire protection include utility control, personnel control, and evacuation.

Preplanning and training by facility and emergency personnel ensures awareness of appropriate actions in fire response and possible hazards involved.

'Code of Federal Regulations, Chapter 10 part 20, U.S. Government Printing Office, 1982.

2 Dorsey, N.E., Properties of Ordinary Water-Substance, Reinhold Publishing Corp., New York p. 537.

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CHAPTER 9, AUXILIARY SYSTEMS - 1/2012 Automatic and manual protection systems in the building include several different type systems. In the academic wing of the building automatic protective actions are provided by a sprinkler system with heat sensitive discharge nozzles, detectors for heat and smoke, and ventilation systems, dampers. The reactor bay wing uses smoke detectors for areas outside the reactor bay that are radiation areas.' The reactor bay ventilation system has smoke detectors that provide a warning of problems within the reactor bay. Although not a strict safety requirement, a gaseous extinguisher system (halon) is installed to protect the reactor instrumentation, control and safety system.

Manual protection equipment includes a dry stand pipe system in each building stairwell.

Portable extinguishers such as C0 2, halon and- dry chemical are placed in specific locations throughout the building.

Elements of the passive fire protection include the structural construction system and the architectural separation into two separate buildings. Building structural materials are concrete cast in place for foundation, concrete walls, support columns and roof. Steel beam, metal and concrete deck comprises the reactor bay roof. A built-up composition roof with fire barrier materials completes the roof system. The building has pre-cast panels that are cast at the construction site cover 75% of the external perimeter. Metal paneling covers the other 25% of the perimeter. Design and installation of systems'and components are subject to the applicable building codes. . -"

The common wall between the academic win g.and the reactor bay wing of the building is a fire barrier. Doors between these two building sections and other penetrations such as HVAC chases will conform to applicable codes. Although a few metal stud and plaster board walls have been used in the reactor bay wing, the typical wall system is of concrete block construction. .

Design specifications are to meet life-safety requirements appropriate for the conditions. These specifications have requirements for emergency lighting, stairwells and railings, exit doors, and other building safety features.: An emergency shower -aid /yewash are available in the hallway adjacent to laboratory areas.,:

Eachof the three components of the fire protection program is applied to the design, operation and modification of the reactor facility and cornporients. Fir'eprevention is primarily a function of operation rather than design.

9.5.3 Operational analysis and safety function The University of Texas maintains an active fire protection system, with periodic testing and inspections to assure systems are prepared to respond.

The halon system automatically actuates if detectors in two control room ceiling units sense an initiating condition in coincidence. The haion system is equipped with a local alarm to prompt Page 9-14

THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR 0 01/2012 SAFETY ANALYSIS REPORT, CHAPTER 9 R'..flETI evacuation of the control room prior to system actuation; a manual override can defeat the system if the nature of the event does not require actuation 3f the system.

Fire suppression is used only in areas where there are no, significan't quan iites of radioactive materials or criticality concerns.

9.5.4 Instruments and controls A fire alarm panel transmits status and alarm information to the University of Texas Police Department dispatch station and a campus information network monitor.

9.5.5 Technical Specifications, bases, testing and surveillances There are no Technical Specifications associated with fire protection.

9.5 Communications systems A communication system of typical te!ephone equipment provides basic services between the building and other off-site points. Supplemental features to this system, such as intercom lines between terminals or points within the building and zone speakers for general announcements are to provide additional communication within the building.

9.5.1 Design basis Communications is required to support routine and emergency operations.

9.5.2 System description The telephone system is insta!led and maintained by the university. Connection of the main university telephone system is to standard commercial telephone network.. Telephones with intercom features are to be located at several locations in the building. Locations include the reactor control room, the reactor bay, and several offices. By use of the intercom feature, each of these locpations will be able to access public address speakers in one of several building zones.

A video camera system and a separate intercom system supplement the building telecommunication network. These two systems contribute to safe operation by enhancement of visual and audio communication between the operator and an experimenter. Each system has 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, as required. A building evacuation alarm system prcmpts personnel to initiate protective actions.

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CHAPTER 9, AUXILIARY SYSTEMS 01/2012 An emergency cell phone is maintained in the control room to compensate for loss of normal communications. A digital radio is kept in the control room to provide emergency communications on first responder and campus frequencies, and to compensate for loss of normal communications.

9.5.3 Operational analysis and safety function The control room has'adequate capabilities to initiate and coordinate emergency response.

There are multiple provisions specificallyto address failures on normal communications channels.

9.5.4 Instruments and controls As specified above 9.5.5 Technical Specifications, bases, testing and surveillances There are no specific Technical Specifications related to communications, but the reactor Emergency 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 supporting irradiation programs and a series of laboratories in the lab andoffice. wing.

9.6.1 Design basis The design;basisof the NETL laboratories is to allowthe safe arid controlled use of radioactive materials-. .

9.6.2 System Description Strategic lab and office wing rooms are equipped with fume hoods and ventilation to control the po.tential.for release of radioactive materials.: One rocm is equipped with two pneumatic transfer systems and a manual port. One system terminates in-a fume hood, monitored by a radiation detector. The other system delivers samples within the tube to a detector. The manual port allows samples to be transferred from the reactor bay to the lab without exiting the reactor bay through normal passageways. A more complete description of the associated laboratories is provided: in Chapter 10.

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0.

THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 9 9 MNtI I 01/2012 9.6.3 Operational analysis and safety function Engineered controls permit safe handling of radioactive materials.

9.6.4 Instruments and controls An installed radiation monitor ensures persornel handlir.g samples from the manual pneumatic sample transfer system are aware of the potential exposure.

9.6.5 Technical Specifications, bases, testing and surveillances There are no specific Technical Specifications related to the laboratories; all operations involved with potential radiation exposure at NETL are managed under the approved reactor Radiation Protection Program.

9.7 Control and storage of reusable components Several experiment facilities that are used in the core are designed to be removed and inserted as required to support various programs.

9.7.1 Design basis Management of experiment facilities is designed to minimize potential exposure to personnel.

9.7.2 System description The 3 element facility, 6 element faiity, pneumatic in-core terminals, and centra! thimble are described in chapter 10. Once irradiated, these facilities are maintained with activated portions in the pool, using pool water as shielding or in other locations typically within the reactor bay 9.7.3 Operational analysis and safety function Maintaining irradiated facilities under water minimizes potential exposure. Corncrete blocks provide temporar/ shielding as needed.

9.7.4 Instruments and controls Instruments and controls associated with specific facilities are addressed in Chapter 10.

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CHAPTER 9, AUXILIARY SYSTEMS 1 01/2012 9.7.5 Technical Specifications, bases, testing and surveillances The basis for Technical Specifications specific to the pool is in Chapter 5, the basis for experiment in Chapter 10.

9.8 Compressed gas systems There are two separate compressed air systems use at the UT facility. One system provides air for laboratories and service connections. One system provides control air.

9.8.1 Design basis Service air is provided to support laboratory and service operations with high capacity applications. Instrument air is intended to support HVAC and reactor operations.

9.8.2 System description One dual compressor system provides oil free compressed air for laboratories and services. The lab air compressor motor is rated at 30 hp. The other system also uses a dual compressor and motor, with 2-stage compressors. The instrument air compressor provides air to HAVC pneumatic controls, pool cooling flow controls. The laboratory air compressor provides aiur to shops and to the transient rod drive system.

9.8.3 Operational analysis and safety function The two systems have dual motors and compressors to provide maximum reliability. The two systems are connected through a manual shut off valve, providing maximum flexibility in the event of a system (or associated air dryer) failure.

Failure of the instrument air system will prevent air from supporting control systems. The pulse rod drive system requires air to couple the drive to the rod; a failure will cause the rod to fall into 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 water system. 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 fail closed, initiating confinement isolation. This is a fail-safe condition, assuring that there is no potential for inadvertent release of radioactive material into the environment in the absence of instrument air.

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THE UNIVERSITY OF TEXAS TRIGA II RESEARCH REACTOR SAFETY ANALYSIS REPORT, CHAPTER 9 0000 ETI 01/2012 9.8.4 Instruments and controls The 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 surveillances There are no Technical Specificatorns specifically associated with the compressed air systems.

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