ML20100L925
| ML20100L925 | |
| Person / Time | |
|---|---|
| Site: | University of Texas at Austin |
| Issue date: | 11/30/1984 |
| From: | TEXAS, UNIV. OF, AUSTIN, TX |
| To: | |
| Shared Package | |
| ML20100L920 | List: |
| References | |
| SAR-9-84, NUDOCS 8412120140 | |
| Download: ML20100L925 (261) | |
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SAR 9-84 TRIGA Reactor Facility Nuclear Engineering Teaching Laboratory The University.of Texas at Austin Submitted November 1984 h 0 602 PDR A
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SAR 9-84 Safety Analysis-Report Tablo of Contents
- 1. Introduction and' Summary ........................ 1-1
-1.1 Principal' Design Criteria 1-1 1.2 Design Highlights 1-1 1.3 Conclusions 1-3 References 1-5
- 2. Site Description................................. 2-1 2.I' General Location and Area 2-1 2.2 Population and Employment 2-6 2.3 Climatology 2-10 2.4. Geology -
2-13 2.5 Hydrology 2-19 2.6 Seismology 2-22 2.7 Historical 2-22 '
References 2-26
- 3. TRIGA Reactor ................................... 3-1 3.1 Design Bases ................................ 3-1 3.1.1 Reactor Fuel Temperature 3-2 3.1.1.1 Fuel and Clad Temperature 3-7 3.1.1.2 Finite Diffusion Rate 3-22 3.1.1.3 Summary 3-26 3.1.2 Prompt Negative Temperature Coefficient 3-27 3.1.2.1 Codes Used for Calculations 3-28 3.1.2.2 ZrH Model 3-30 3.1.2.3 Calculations 3-30 3.1.3 Steady-State Reactor Power 3-33 3.1.3.1 Entrance Loss 3-37 3.1.3.2 Exit Loss 3-37 3.1.3.3 Loss Through Portion of Channel Adjacent to Lower Reflector 3-38 3.1.3.4 Loss Through Portio,n of Channel Adjacent to Upper Reflector 3-38 3.1.3.5 Loss Through Each Increment of-the Channel Adjacent to the Fueled Portion of the Elements 3-38 3.1.3.6 Acceleration Term 3-38 i 3.1.3.7 Friction Term 3-39 3.1.3.8 Gravity Term 3-42 3.1.3.9 Nomenclature 3-47 3.2 Nuclear Design and Evaluation................ 3-50 3.2.1 Reactivity Effects 3-50 3.2.2 Evaluation of Nuclear Design 3-54
(- 3.3 Thermal and Hydraulic Design................. 3-55 l- 3.3.1 Design Bases 3-56 3.3.2 Thermal and Hydraulic Design Evaluation 3-57 l
SAR 9-84 3.4 Mechanical Design and Evaluation............. 3-58 3.4.1 General Description 3-58 3.4.2 Reflector Assembly. 58 3.4.3 Grid Plates 3-61 3.4.4 Safety Plate 3-63 3.4.5 Fuel-Moderator Elements 3-63 3.4.5.1 Evaluation of Fuel Element Design 3-65 3.4.6 Neutron Source Holder 3-65
.3.4.7 Graphite Dummy Elements 3-68 3.4.8 Control System Design 3-68 3.4.8.1 Control Rod Drive Assemblies 3-70 3.4.8.2 Transient Rod Drive Assembly 3-71 3.4.8.3. Evaluation of Control Rod System .
3-75 3.4.9 Experimental and Irradiation Facilities 3-76 3.4.9.1 Central Thimble 3-76 3.4.9.2-Rotary Specimen ~ Rack 3-76 3.4.9.3 Pneumatic Specimen Tube 3-76 3.4.9.4 Beam Tube Facilities 3 3.5 Safety Settings in Rslation to Safety Limits. 3-78 =
References 3-80
- 4. Instrumentation and Control System............... 4-1 4.1 Design Bases .
4-1 4.1.1'N-M 1000 Safety and Neutron Monitor Channel 4-2 4.1.2 CSC and Control Console 4-3 4.1.3 Reactor Operation Modes 4-7 4.1.4 Reactor Scram System 4-9 4.1.5 Logic Functions 4-10 4.1.6 Mechanical Hardware 4-12 4.2 Design Evaluation............................ 4-14 References 4-15
- 5. Water Coolant and Purification Systems .......... 5-1 5.1 Design Bases ................................ 5-1 5.1.1 Reactor Core Heat Removal 5-2 5.1.2 Reactor Pool Heat Removal 5-2 5.1.3 Heat Exchanger Design Basis 5-2 5.1.4 Water Purification Bases 5-5 5.2 System Design ............................... 5-5 5.2.1 Coolant System 5-5 5.2.2 Purification System 5-7 5.2.3 Water System Instrumentation 5-8 5.3 Water System Design Evaluation .............. 5-10 References 5-13
- 6. Facility Design ................................. 6-1 6.1 Design Bases ....................,............ 6-1 L _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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SAR 9-84 6.2 Reactor Bay and Control ~ Area ................ 6-7 6.2.1 Physical Design 6-7 6.2.2 Venti?tation' Design 6-8 6.2.3 Reactor Shield Structure 6-8
'6.3. Support Facilities .......................... 6-12 6.3.1 Radioactive Waste Control 6 ~- 12 6.3.2 Sampling Handling Laboratory 6-12 6.3.3 Health Physics Laboratory 6-15 ,
6.3.4 other Laboratories 6-15 f.3.5 Support Areas 6-15 6.4 Special Experimental Facilities ..............6-15 6.4.1 Reactor Core-Facilities 6-15 6.4.2 Beam Tubo Facilities 6-16 6.4.3 Cobalt-60 Irradiation Facility 6-16 6.4.4 Subcritical Reactor and Moderators 6-18 6.4.5.14 Mev Neutron Generator 6-18 6.5 Containment Design Evaluation ............... 6-18 6.5.1 Release of Argon-41 and Nitrogen-16 from Pool Water 6-19 6.5.1.1 Argon-41 Activity in Reactor Room 6-19 6.5.1.2 Nitrogen-16 Activity ~in Reactor Room 6-24 6.5.2 Activation of Air in the Experimental Facilities 6 References 6-31
- 7. Safety Analysis ................................. 7-1 7.! Reactivity Accident ......................... 7-1 7.1.1 Summary 7-1 .
L 7.1.2 Analysis of 2.8% insertion ~at 1 kW 7-2 7.1.3 Analysis of 2.8% insertion at 880 kW 7-8 7.2 Loss of Reactor Coolant ..................... 7-11 7.2.1 Summary 7-11 7.2.2 Fuel Temperature and Clad Integrity 7-13 ,
7.2.3 After-Heat Removal Coolant Loss 7-19 ;
7.2.4 Radiation Levels 7-22 7.3 Fission Product Release ..................... 7-26 7.3.1 Fission Product Inventory 7-26 L 7.3.2 Fission Product Release Fractions 7-26
! 7.3.3 Downwind Dose Calculation 7-29 i 7.3.4 Downwind Doses 7-30 ;
f References 7-33 I
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- 8. Facility Administration ......................... 8-1 i
8.1 Organization ................................ 8-1 f 8.1.1 Structure 8-1 l
, 8.1.2 Vice President for Research and Academic l Affairs 8-1 l 8.1.3 Director of Nuclear Engineering Teaching !
Laboratory 8-1 l
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SAR 9-84
- 8.1.4 Radiation Safety Committee 8-3 8.1.5 Radiation Safety Officer 8-3 8.1.6 Reactor Committee 8-3 i
8.1.7 Laboratory Supervisor 8-3 P
8.2 Qualifications ............................. 8-4 8.2.1 General 8-4 8.2.2 Academic Administration and Radiological Safety 8-4 8.2.3. Facility Director 8-4 8.2.4 Reactor Supervisor 8-4 8.2.5 Operators. Technicians, and others 8-4 8.3 Reactor Operations ......................... 8-5
'8.3.1 Staffing 8-5 8.3.2 Procedures 8-6 8.3.3 Experiments 8-6 8.4 Actions and Reports ......................... 8-6 8.4.1 Operating Reports 8-6 8.4.2 Safety Limit Violation 8-7 8.4.3 Release of Radioactivity 8-7 8.4.4 Other Reportable Occurrences 8-8 8.4.5 Other Reports 8-8 8.5 Records .................................... 8-9 8.5.1 Records to be Retained for the Lifetime 8-9 of the Reactor Facility 8.5.2 Records to be Retained for a Period of at Least Five Years or for the Life of the Component Involved. Whichever is Smaller 8-9 8.5.3 Records to be Retained for at Least One Training Cycle 8-9 References 8-10
- 9. Quality Assurance ............................... 9-1 9.1 Introduction ................................ 9-1 9-2 9.1.1 Purpose 9.1.2 Responsibility 9-2 9.1.3 Organization 9-3 9.1.4 Documentation 9-3 9.2 Quality Assurance Controls .................. 9-6 9.2.1 Design Controls 9-6 9.2.2 Procurement Controls 9-6 9.2.3 Document Control 9-7 9.2.4 Material control 9-7 9.2.5 Process Control 9-7 9.3 Inspection and Correction Actions .......... 9-7 9.3.1 Inspection Program 9-7 9.3.2 Test Program 9-8 9.3.3 Measuring and Test Equipment 9-8 9.3.4 Non-Conforming Material and Parts 9-8 9.3.5 Corrective Action 9-8 9.4 Records and Audits ........................,.9-8 9.4.1 Quality Assurance Records 9-8 9.4.2 Audits 9-9 References 9-11 l
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SAR 9-84
- 10. Radiological Protection ......................... 10-1 10.1 Radiological Management Organization ....... 10-1 10.1.1 Management and Policy 10-1 10.1.2 Responsibilities 10-1 10.1.3 Organizational Access 2 10.1.4 Equipment and Supplies 10-2
.10.1.5 Training and Safety. 10-2 10.2 Radioactive. Materials Control .............. 10 10.2.1 Reactor. Fuel 10-3 10.2.2 Reactor Components 10-3 10.2.3 Experiment Facilities 10-3 10.2.4 Activated Samples 10-3 10.2.5 Radioactive Waste 10-5 10.2.6 Other Materials 10-5 10.3 Radiation Monitoring ....................... 10-5 10.3.1 Minimum Procedures 10-6 10.3.2 Monitoring Techniques 10-7 10.3.3 Management Surveillance 10-7 10.3.4 Frequency and Accuracy 10-7 10.4 Instrumentation ............................ 10-8 10.4.1 Fixed Area Monitors 10-8 10.4.2 Airborne Radiotctivity Monitors 10-8 10.4.3 Laboratory Instrumentation 10-8 10.4.4 Liquid Effluents 10-8 10.4.5 Range and Spectral Response 10-9 10.4.6 Calibrations 10-9 10.5 Records .................................... 10-9 10.6 Emergency Plan and Radiological Program Review ..................................... 10-9 References 10-10
- 11. Fire Protection ................................. 11-1 11.1 Fire Protection Components 11-1 11.1.1 Passive Fire Protection Elements 11-1 11.1.2 Active Fire Protection Elements 11-2 11.1.3 Fire Prevention Elements 11-2 11.2 Fire Protection Controls 11-2 11.2.1 Facility Fire Protection Elements 11-2 11.2.2 Facility Fire Protection Control' 11-3 11.3 Fire Safety Assurance 11 References 11-5 l
- 12. Training and Certification of Operators ......... 12-1 12.1 Training Subjects 12-1 12.2 Training Experience 12-1 12.3 Evaluation 12-2 12.4 Records 12-2 References 12-4
SAR 9-84
- 13. Startup Program ................................. 13-1 13.1 Storage of Fuel and Acquisition of Components 13-1 13.2 Tests of Systems Before Core Londing 13-2 13.3 Core Load for Initial Criticality 13-2 13.4 Tests Subsequent to Core Criticality 13-2 13.5 Acceptance fo" ))3 ration 13-3
S Alt 9-84
- 1. i s t of Tables 1-1 PRINCIPAL DESIGN PARAMETERS l-2 2-1 TRAVIS COUNTY 1980 POPULATION DENSITY DISTRIBUTION 2-7 2-2 1982 METEOROLOGICAL DATA FOR AUSTIN TEXAS 2-14 2-3 HISTORICAL METEOROLOGICAL DATA FOR AUSTIN TEXAS 2-15 2-4 GROUND WATER ACTIVITY (GROSS BETA) 2-22 2-5 TANK SLUDGE SAMPLES 2-24 3-1 PHYSICAL PROPERTIES OF DELTA PHASE U-ZrH 3-2 3-2 HYDRAULIC FLOW PARAMETERS 3-34 3-3 TYPICAL TRIGA CORE NUCLEAR PARAMETERS 3-50 3-4 ESTIMATED CONTROL ROD NET WORTH 3-51 3-5 ESTIMATED FUEL ELEMENT REACTIVITY WORTH COMPARED WITH WATER AS A FUNCTION OF POSITION IN CORE 3-51 3-6 EXPECTED REACTIVITY EFFECTS ASSOCIATED WITH
, EXPERIMENTAL FACILITIES 3-54 3-7 COMPARISON OF REACTIVITY INSERTION EFFECTS 3-55 3-8 1000 KW (t) TRIGA HEAT TRANSFER AND HYPRAULIC PARAMETERS 3-57 3-9 THERMOCOUPLE SPECIFICATIONS 3-67 3-10
SUMMARY
OF FUEL ELEMENT SPECIFICATIONS 3-67 3-11
SUMMARY
OF CONTROL ROD DESIG3 PARAMETERS 3-70 3-12 TRIGA SAFETY SETTINGS 3-79 5-1 REACTOR COOLANT SYSTEM DESIGN 5-7
SUMMARY
5-2 HEAT EXCHANGER llEAT TRANSFER AND HYDRAULIC PARAMETERS 5-12 6-1 SATURATED ARGON CONCENTRATIONS IN WATER 6-20 6-2 VOLUMES AND THERMAL FLUXES OF FACILITIES 6-28 7-1 REACTIVITY TRANSIENT INPUT PARAMETERS 7-3 7-2 REACTIVITY TRANSIENT INPUT PARAMETERS 7-9 7-3 CALCULATED RADIATION DOSE RATES FOR LOSS OF SHIELD WATER 7-23 7-4 NOBLE GAS AND HALOGENS IN THE REACTOR 7-27 7-5 ASSUMED BREATHING RATES 7-31 7-6 AVERAGE GAMMA RAY ENERGY AND INTERNAL DOSE EFFECTIVITY FOR EACH FISSION PRODUCT ISOTOPE 7-31 7-7 DOSES FROM FISSION PRODUCT RELEASE 7-32 9-1 RESPONSIBILITIES AND KEY PERSONNEL 9-3 9-2 FORMAT FOR SAFETY RELATED QA CHECKS 9-5 9-3 QUALITY ASSURANCE PROGRAM AUDIT PROCEDURES 9-10 A
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i SAR 9-84 List of Figures 2-1 STATE,0F TEXAS COUNTIES 2-2 2-2 TRAVIS COUNTY 2-3 2-3 CITY OF-AUSTIN 2-4 2-4 BALCONES=RESEARCH CENTER 2-5 2-5 TRAVIS COUNTY 1980 CENSUS TRACT 2-8 BOUNDARIES 2-6 CITY OF AUSTIN CENSUS TRACT BOUNDARIES 2-9 7 AUSTIN CLIMATOLOGY DATA 2-11 2-8 AUSTIN WIND ROSE DATA 2-12 2-9 TEXAS TORNADO FREQUENCIES 2-16 2-10 TEXAS HURRICANE PATHS '
2-17 2-11 LOCAL FUNNEL CLOUD SITINGS 2-18 2-12 BALCONES FAULT ZONE 2-20 2-13 LOCAL WATER ACQUIFIERS 2-21 2-14 TEXAS. EARTHQUAKE DATA 2-23 2-15 BALCONES RESEARCH CENTER 1980 2-25 3-1 PHASE DIAGRAM OF THE ZIRCONIUM-HYDROGEN SYSTEM 3-3 3-2 EQUILIBRIUM HYDROGEN PRESSURE VERSUS TEMPERATURE FOR ZIRCONIUM-HYDROGEN 3-5 3-3 STRENGTH OF TYPE 304 STAINLESS STEEL AS A FUNCTION OF TEMPERATURE 3-6 3-4 STRENGTH AND APPLIED STRESS AS A FUNCTION OF TEMPERATURE, EQUILIBRIUM HYDROGEN DISSOCIATION PRESSURE 3-8 ,
3-5 RADIAL POWER DISTRIBUTION IN.THE U-ZrH FUEL
, ELEMENT 3-10 3-6 POWER DISTRIBUTION IN THE U-ZrH FUEL ELEMENT 3-11 3-7 .SUBC00 LED BOILING HEAT TRANSFER FOR WATER 3-12 3-8 CLAD TEMPERATURE AT MIDPOINT OF WELL-BONDED FUEL ELEMENT 3-14 3-9 FUEL BODY TEMPERATURES AT MIDPLANES OF WELL-BONDED FUEL ELEMENT AFTER A PULSE 3-15 3-10 SURFACE HEAT FLUX AT MIDPLANE OF WELL-BONDED FUEL ELEMENT AFTER A PULSE 3-16 3-11 SURFACE HEAT FLUX DISTRIBUTION FOR STANDARD NON-GAPPED (h Sap n 500) FUEL ELEMENT AFTER'A PULSE 3-18 3-12 SURFACE HEAT FLUX DISTRIBUTION FOR STANDARD NON-GAPPED (h =
375) FUEL ELEMENT AFTER A PULSE 83P 3-19 3-13 "URFACE HEAT FLUX DISTRIBUTION FOR STANDARD NON-GAPPED (h sap = 250) FUEL ELEMENT AFTER A PULSE 3-20 14 SURFACE HEAT FLUX AT MIDPOINT VERSUS TIME FOR STANDARD-NON-GAPPED FUEL ELEMENT.AFTEF A PULSE 3-21 3-15 TRANSPORT CROSS SECTION FOR HYDROGEN IN ZrH AND AVERAGE NEUTRON SPECTRA IN FUEL ELEMENT 3-29 3-16 A COMPARISON OF NEUTRON SPECTRA BETWEEN EXPERIMENTS AND SEVERAL HYDROGEN MODELS 3-31
SAR 9-84
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3-17 EFFECT 0F TEMPERATURE VARIATION ON'Z'IRCONIUM
' HYDRIDE NEUTRON SPECTRA 3-32 3 PROMPT NEGATIVE TEMPERATURE COEFFICIENT VERSUS-AVERAGE FUEL TEMPERATURE FOR TRIGA 3-35 3-19 GENERAL FUEL ELEMENT. CONFIGURATION FOR SINGLE COOLANT CHANNEL IN THE TRIGA 3-36 3-20 EXPERIMENTALLY DETERMINED VAPOR VOLUMES FOR SUBC00 LED BOILING IN A NARROW VERTICAL ANNULUS 3-40 3-21 CROSS' PLOT OF FIGURE 3-20 USED IN CALCULATIONS 3-41 3-22 PLOT FOR WHICH DNB RATIO IS 1.0 0F MAXIMUM HEAT
. FLUX VERSUS COOLANT TEMPERATURE 3-46
-! 3-23 ESTIMATED REACTIVITY LOSS VERSUS POWER 3-52 3 ESTIMATED MAXIMUM B RING AND AVERAGE CORE TEMPERATURE VERSUS POWER 3-53 3'-25' TYPICAL MARK I TRIGA REACTOR 3-59 3-26 REACTOR, REFLECTOR, AND SHIELDING 3-60 3 CORE ARRANGEMENT 3-62 3-28 TRIGA STAINLESS STEEL CLAD FUEL ELEMENT WITH END FITTINGS 3-64 3-29 INSTRUMENTED FUEL ELEMENT 3-66 3 FUEL FOLLOWED CONTROL ROD 3-69 3-31 RACK AND PINION CONTROL ROD DRIVE 3-72 t
3-32 ADJUSTABLE TRANSIENT ROD 3-73 3-33 TRANSIENT' ROD OPERATIONAL SCHEMATIC 3-74 4-1 NEUTRON CHANNEL OPERATING RANGES 4-4 4-2 LAYOUT OF THE REACTOR CONTROL CONSOLE 4-5 4-3 CONSOLE CONTROL PANELS 4-6 4-4 LGMIC DIAGRAM FOR CONTROL SYSTEM 4-11 4-5 LOCATION OF CONTROL SYSTEM COMPONENTS 4-13 5-1 COOLANT AND PURIFICATION SYSTEM LAYOUT 5-6 5-2 WATER SYSTEM INSTRUMENTATION 5-9 6-1 NETL SITE PLAN FOR BALCONES RESEARCH CENTER 6-2 6-2 NETL BUILDING FIRST. LEVEL 6-3 6-3 NETL BUILDING SECOND LEVEL 6-4 6-4 NETL BUILDING THIRD LEVEL 6-5 6-5 CROSS SECTION OF REACTOR FACILITY AREA 6-6 6-6 REACTOR BAY AREA 6-9 6-7 REACTOR BAY AIR VENTILATION SYSTEM 6-10 6-8 REACTOR BAY AUXILIARY EXHAUST SYSTEM 6-11 6-9 REACTOR SHIELD STRUCTURE 6-13 6-10 RADIOACTIVE EFFLUENT HANDLING SYSTEMS 6-14 6-11 SPECIAL EXPERIMENT FACILITIES 6-17 7-1 CALCULATED PEAK PULSE POWERS 7-3 7-2 FUEL TEMPERATURE DISTRIBUTION BEFORE AND AFTER PULSE 7-12 3 FUEL TEMPERATURE AND POWER DENSITY FOR ELEMENT COOLING TIMES 7-14 7-4 U-ZrH(1.6)-STRENGTH AND STRESS VERSUS TEMPERATURE 7-15 7-5 COOLING TIMES AFTER REACTOR SHUTDOWN TO LIMIT MAXIMUM FUEL TEMPERATURE VERSUS POWER DENSITY 7-16
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SAR 9-84 8-1 ADMINISTATIVE STRUCTURE 8-2 9-1 ACADEMIC ORGANIZATION 9-4 10-1 BUSINESS ADMINISTRATION 10-4 I
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SAR'9/84-Chapter 1.
1 INTRODUCTION'AND
SUMMARY
This report describes the TRIGA reactor and the g
-University. of Texas facility, and provides a safety
, evaluation which'shows that the reactor or facility does-not cause: undue risk to the health'and safety of the.public. A I
'fRIGA type reactor was first operated- in 1963 on the main campus of The University of Texas at Austin. Subsequent operation experience included-safe operation of the facility
, at. steady. state thermal power levels of 10 kW- and 250.kW, and pulse powers of 250 MW.
Safe operation of a TRIGA reactor at the Balcones Research Center of the University of.
Texas is expected for steady state power levels of 1.0 MW and pulse powers of 1400 MW.
Some values used in this report represent the latest design. parameters, or maximum values .as a means of evaluating the safety of the system. For-this reason, these L values may differ from.those quoted in other. documents or-4' from those that will be measured in the operating reactor
~
system. Safety analysis demonstrates safe operation at-power levels as high as 1.5 MW steady state.and 8400 MW peak pulse power.
1.1 PRINCIPAL DESIGN CRITERIA The reactor will be operated in-two modes: steady state and pulsing. Reactor power levels in the steady state mode will-range up to and include 1 MW(t). Pulsed mode operation will take place by step reactivity insertions -with .the reactor initially at a power level less than 1'kW. The maximum step reactivity insertion will be 2.1% 6k/k ($3.00) l which will produce a peak reactor power of approximately 1400 MW(t) with a prompt energy release of about 18 MW-sec.
A summary of principal design parameters for the reactor is given in-Table 1-1.
4
- 1. DESIGN HIGHLIGHTS
~
.The reactor will be located in a reactor pool structure.
Reactor cooling ~ will~be provided by natural circulation o f-pool water which is cooled and purified in external coolant
. circuits. . Reactor experiment facilities will include a rotary . specimen rack,- a pneumatic transfer system, core irradiation tubes, and horizontal and vertical beam tubes.
Manufactured:by GA. Technologies Inc.
1-1
SAR 9/84 Table 1-1
, PRINCIPAL. DESIGN PARAMETERS Reactor type TRIGA Mark II Steady-state power (maximum) 1 MW Pulse power (maximum) 2.1% 6k/k ($3.00)
Fuel element design Fuel-moderator material U-ZrH(a)
Uranium content 8.5 wt %
Uranium enrichment 19.7% U-235 Shape Cylindrical Length of fuel 38 cm (15 in.) overall Diameter of fuel ,
3.63 cm (1.43 in.) o.d.
Cladding material 304 stainless steel Cladding thickness 0.051 cm (0.020 in.)
Number of fuel elements Critical core '64 Operational core '90 Excess reactivity, maximum 4.9% 6k/k Number of control rods 4 transient I safety, shim (fuel followed) 2 regulating (fuel' followed) 1 Total reactivity worth of rods 8.7% 6k/k Reactor cooling Natural convection of pool water (a) The nominal'd/Zr ratio is 1.60, and the maximum value is 1.65.
1-2 .
SAR 9/84 The inherent safety of this TRIGA reactor has been demonstrated by the extensive experience acquired from similar TRIGA systems throughout the world. Forty-eight TRIGA reactors are now in operation throughout the world and of these 31 are pulsing. TRIGA reactors have more than 450 reactor -years. of operating experience, over 30,000 pulses, and'more than 15,000 fuel element years of operation. The safety arises from a large, prompt negative temperature coefficient that is characteristic of uranium zirconium hydride fuel-moderator elements used in TRIGA systems. As the fuel temperature increases, this coefficient immediately compensates for reactivity insertions. The result is that reactor power excursions are terminated quickly and safely.
The prompt shutdown mechanism has been demonstrated extensively in many thousands of transient tests performed on two prototype TRIGA reactors at the GA Technologies laboratory in San Diego, California, as well as other pulsing TRIGA reactors in operation. These tests included step reactivity insertions as large as 3.5% 6k/k with resulting peak reactor powers up to 8400 MW(t) on TRIGA cores containing similar fuel elements as are used in this TRIGA reactor.
Because the reactor fuel is similar, the previously cited experience and tests apply to this TRIGA system. As a result it has been possible to use accepted safety analysis techniques applied to other TRIGA facilities to update evaluations with regard to the characteristics of this facility (1-6].
1.3 CONCLUSION
S Past experience has shown that TRIGA systems can be designed, constructed, and safely operated in the steady state and pulsing modes of operation. This history of safety and the conservative design of the reactor have permitted TRIGA systems to be sited in urban areas using buildings without pressure type containment such as is normally associated with reactors of like power levels.
Results of this safety analysis indicate that the TRIGA Mark II reactor system proposed for construction and operation will pose no health or safety problem to the public when operated in either normal or abnormal conditions.
1-3
SAR 9/84 Abnormal or accident conditions considered in this analysis include:
- a. A step insertion of reactivity with the reactor at low and high power levels,
- b. Complete and instantaneous loss of coolant water in the' reactor pool.
-c. And fission product release from fuel element ruptured in air.
The insertion of excess reactivity may represent a normal reactor operating condition, while the loss of pool water.is expected to be an abnormal condition. . Conservative estimates of doses from fission product releases are made independent of accident scenarios.
In both these postulated conditions, fuel and clad temperatures remain at levels below those required to generate stress conditions which would cause loss of clad integrity. However, the results of a clad failure are analyzed and it is shown that such a failure will not cause excessive radiation exposures.
The loss of pool water has been examined from the standpoint of direct radiation to operating personnel as well as in terms of maintaining fuel integrity.
The effects of argon-41 and nitrogen-16 production during normal operation of the reactor have also been evaluated. Results of these analyses show that production of these radioactive gases will present no hazard to persons in the reactor room or to the general public.
1-4
SAR 9/84 Chapter'1 References
- 1. " Hazards Report for TRIGA. Mark II Pulsing Reactor",
General Atomic Division Report GA-1998, February 1961.
- 2. " Hazards Summary Report for a TRIGA-I Nuclear Reactor".
University of Texas Bureau of Engineering Research, October 1961.
- 3. " Safeguards Analysis Report for TRIGA Reactors using Aluminum-Clad Fuel", General Atomic Division Report GA-7860, March 1967.
- 4. " Safety Analysis Report for 250 Kilowatt Operation of a TRIGA Mark I Nuclear Reactor", University of Texas, College of Engineering August 1967.
- 5. " Safety Analysis Report for the TRIGA Mark II Retetor".
E-117-478, General Atomic Company, October, 1975.
4 1-5 i
p SAR 9/84 Chapter 2 SITE DESCRIPTION The facility containing the TRICA reactor will be situated on the east track of the Balcones Research Center
[1], a tract of land owned and operated by The University of Texas.
County The Research and the Center is located in northern Travis City of Austin about 11.6 kilometers north-northwest of The University Figures 2-1 thru 2-4 display the facility of Texas at Austin campus.
locations in relarion to surrounding areas. Located near the transition line between hill country and rolling plains, the site is situated about 7.4 kilometers from where the flo;d controlled Colorado river crosses the transition region and Balcones fault zone. The Balcones Research Center east and west tracts span part of the inactive fault zone. The east tract is within the transition region to rolling plains.
2.1 GENERAL LOCATION AND AREA State Maj or activities of The University of Texas at Austin, of Texas government, and City of Austin business district are centered at respective distances of 11.6, 12.6, and 12.9 air kilometers to the south-southwest. Distances to traffic landing facilities in the area are 6.2 kilometers for private aircraft, 9.7 kilometers for commercial aircraft and 20.8 kilometers for military aircraft.
A total area of 1.87 square kilometers is contained within the Research Center east and west tracts of land.
The east side of the Center is bounded by a State highway.
FM 1325, and the west side is bounded by a Federal highway, US 183. The two tracts are divided by a rail line, formerly the Missouri-Pacific, with 0.93 square kilometers in the east tract and 0.94 square kilometers in the west tract of land. Highway intersections of US 183 with FM 1325 and with Loops 1 and 360 are within two kilometers of the site. Both highway loops are planned for extension into the area associated with the west tract of land [2]. The reactor facility is sited near the center of the northern half of the east tract. Distance from the reactor site to nearest rail line or highway is approximately 250 meters.
Most areas adjacent to the Research Center are developed for mixed commercial and industrial activities including warehouses, manufacturing facilities, small business parks and a few undeveloped areas. Mixed 2-1 1
SAR 9/84
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SAR 9/84 commercial and industrial areas south and east of the Research Center are bounded by highway US 183, highway FM 1325, and the Texas New Orleans Railroad to the east.
Approximately 2.2 square kilometers of land are enclosed by the area. Much of the remaining area to the west of the Research Center is bounded by highway US 183 and the Missouri Pacific Railroad and is not developed. The area is planned for future road right-of-ways and includes the west tract area. Immediately north of the Balcones Research Center east tract is a 2.3 square kilometer complex operated by International Business Machines Corporation. Undeveloped areas around the Research Center are expected to develop as a mixed commercial, industrial, research or office park areas. Residential areas are located beyond adjoining areas around the Balcones Research Center with distances from the reactor facility site of 1.2 kilometers to 2.0 kilometers.
Few residential structures for either multifamily or single family units are located within a radius of 1.2 kilometers of the reactor site.
2.2 POPULATION AND EMPLOYMENT Austin is composed primarily of governmental, business, and professional persons with their families. The city has substantial light industry but practically no heavy industry. Many of the persons in the local labor force are related to activities of the City and its role as a State Capitol or the University and its educational and research programs. Population growth of Travis county between the 1970 and 1980 census was 42%. The substantial growth of the
" sun belt" areas, and Austin in particular, is expected to continue throughout the 1980's. Population data of the Travis county area is presented in Table 2-1 with supportive data in Figure 2-5 and Figure 2-6.
The 1980 population census listed the Austin city population at 345,496 and Travis county population at 419.573 with the Austin Standard Metropolitan Statistical Area population at 536,450 [3]. Three counties, Travis, Hays and Williamson, compose the Austin Standard Metropolitan Statistical Area. Population densities in Travis county range from 6.4 persons per 1000 square meters encompassing the cein university campus to less than 12 persons per 100 square meters in growth areas north of the research center site. Population census tracts adjacent to the site exhibit densities of 1.2 to 2.0 persons per 1000 square meters.
Approximately 800 persons were involved in activities on the east tract of the Balcones Research Center in 1981 with projected activities at the site to add an estimated
- 900 to 1000 persons by the late 1980's. On the west tract the Microelectronics and Computer Technology Corporation is 2-6 L
SAR 9/84 Table.2-1 TRAVIS COUNTY 1980 POPULATION DENSITY DISTRIBUTION 3:
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TRAVIS COUNTY 1980 CENSUS TRACT BOUNDARIES Figure 2-5 2-8
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SAR 9/84 expected.to. employ 1000 persons by 1985. Facilities north of the Research Center operated by International Business Machines Corporation are expected to employ 6500 persons in 1985.
Research activities at the Balcones Research Center are diverse, consisting of many different research organizations of .the university science and engineering colleges.
Research organizations located adjacent to the site for the TRIGA reactor are the Center for Energy Studies, Center for Electromechanics, Bureau of Economic Geology and Water Resources Center.
2.3 CLIMATOLOGY Austin, capital of Texas, is located on the Colorado River. where the stream crosses the Balcones Escarpment separating the Texas Hill Country from the Blackland Prairies to the east. Elevations within the City vary from 120 meters to 275 meters above sea level. Native trees include cedar, oak, walnut, mesquite, and pecan.
The climate [4] of Austin is humid subtropical-with hot summers. Winters-are mild, with below freezing temperatures occurring on an average of less than twenty-five days each year. Rather strong northerly winds, accompanied by sharp drops in temperature, occasionally occur during the winter months in connection with cold fronts, but cold periods are usually of ehort duration, rarely lasting more than two days. Daytime temperatures in summer are hot, but summer nights are usually pleasant with average daily minima in the low seventies.
Precipitation is fairly evenly distributed throughout the year, with heaviest amounts occurring in late spring. A secondary rainfall peak' occurs in September. Precipitation.
from April through September usually results from thundershowers, with fairly large amounts falling within short periods of time. While thunderstorms and heavy rains have occurred in all months of the year, most of the winter precipitation occurs as light rain. Snow is insignificant as a source of moisture, and usually melts as rapidly as it falls. The City may experience several seasons in succession with no measurable rain fall.
Prevailing winds are southerly throughout the year.
Northerly winds accompanying the colder air masses in winter soon shift to southerly as these air masses move out over the Gulf of Mexico.
Climatology data is summarized in Figure 2-7. Typical Austin wind data are presented in Figure 2-8 [5].
2-10
SAR 9/84 a
l CLIMATOLOGICAL DATA TEMPERATURE (oEGREEs FAMRENNE T) 100 _ _ i 80 ,
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r SAR 9/84 The -average length .of the warm season (freeze-free
< period) _is 270 days. Based on data from 1943-1961, 'the average date of the last occurrence'of freezing-or below has occurred as late as April.13 (1940), and as early as_0ctober
~26 (1924). _ Meteorological data - is _. tabulated in Table 2-2 and Table.2-3 [4].
Destructive winds and damaging hailstorms are infrequent. On rare occa_sions, dissipating tropical storms effect .the City with strong winds and heavy rains. The frequency of tornado type activity is illustrated in Figure 2-9 [6). Recent tropic storm paths and local sitings' of tornadoes and funnel clouds are presented in Figure 2-10 and Figure 2-11 [7,8].
2.4 GEOLOGY The northwestern half of Travis county is part of the'
- physiographic province of Texas known as the Edwards Plateau. In Travis County, this is a highly dissected plateau with wooden hills rising in some places more than
' 150 meters above the drainageways. In marked contrast, the southeastern half of the _ county _ is gently rolling prairie land which is part of the physiographic province known as the Gulf Coastal Plain. These provinces are separated by the scarp of the Balcones fault zone, which rises 30 to 90 meters above the Coastal Plain. The scarp, however, is not
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a vertical cliff; it_ is an indented line of sloping hills leading up from the lower plain to the plateau summit.
The rocks that outcrop in Travis County are primarily of sedimentary origin and of Mesozoic (Cretaceous) and Cenozoic age. They consist largely of limestone, clay, and sand strata which dip southeastward toward the Gulf of Mexico at an angle slightly greater than. the slope of the land surface. Therefore, in going from southeast to northwest the outcrops of progressively older formations are encountered, and the rocks lowest- in the geologic column have the highest topographic exposure.
At the reactor facility site on the - east tract, the geology is of the Austin Group defined as chalk, marly limestone, and limestone with light gray, soft to hard, thin to thick bed, and massive to slightly nodular character. On the west tract, the geology changes to the Edwards Formation of limestone and dolomite with light gray to tan, hard to soft, thin to thick bed, and fine to medium grain character.
The separate formations are, respectively, the up and down side of a segment of the Mount Bonnell Fault that passes approximately along the boundary of the east and west Balcones Research Center tracts. Distance to the fault is about 500 meters from the reactor' facility site.
2-13
SAR 9/84 Table 2-2 1982 METEOROLOGICAL DATA FOR AUSTIN TEXAS
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SAR 9/84
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f' Table.2-3 HISTORICAL METEOROLOGICAL DATA FOR AUSTIN TEXAS Average Temperature Heating Depec Dayr .
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SAR'9/84 4
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s SAR 9/84 4
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Tornado and funnel cloud occurences in a 50 mile radius of Austin for the years 1975-1983; # Tornados (# Funnel Clouds).
LOCAL FUNNEL CLOUD'51 TINGS Figure 2-11 2-18
- .. . . _ . - _ . , _ _ . . - _ . . _ , - ~ . _ - . _ . _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ . . . . _ . . . . _ _ . _ _ _ _ _ _ _ . . _ _ . . . _ _ . - . - _ . . _ _ _ . . _ . _ _ _ . _ _ _ -
c-SAR 9/H4 i
The Balcones fault zone, which extends from Williamson County to'Uvalde County, extends the full length of Travis County on a line passing through Manchaca, Austin, and McNeil'. Here the orderly sequence of formations is replaced by an outcrop pattern controlled by the faults, most of which are normal faults with the down-thrown side toward the coast.- Most of the movement of the Balcones Fault zone occurred during the Miocene period. Since no movement has been detected during modern times, this fault is no longer considered active [9]. Location of the Balcones Fault zone and formations in the Austin area are depicted in Figure 2-12.
2.5 HYDROLOGY Almost the entire county is drained by the Colorado River and its tributaries. Lake Travis, which is formed by the Mansfield Dam on the Colorado River, is part of the power, flood-control, water conservation, and recreation project of the Lower Colorado River Authority. Other lakes are also operated by the Authority, such as Town Lake and Lake Austin, and are created by Longhorn and Tom Miller dams, respectively. Low level alluvial deposits of the river are commonly saturated with water at relatively shallow depths. Recharge is primarily from the river and local surface contaminations are easily transmitted to this shallow water table.
Ground water from subsurface formation is found in basal. Cretaceous sands referred to as the " Trinity" sands.
Elevations of the Trinity aquifer range from depths commonly less- than 300 meters east of the Balcones Fault Zone to greater than 450 meters to the west of the zone. East of the Mount Bonnell Fault,- dolomite and dolomite limestones provide a source of ground water at shallower depths.
Access to the Edwards aquifer ranges from 30 meters to 300 meters with natural springs occurring in areas near the Colorado River. Minor aquifers associated with the Glenn Rose Formation supplies small quantities of water west of the Balconcs Fault Zone. Water bearing areas in the formation are at varying depths and literally discontinuous.
On the Balcones Research Center cast tract, wells drilled for environmental monitoring have produced ground water at depths of less than 15 meters. Figure 2-13 shows the location of the ground water aquifers.
Water supply for the research center and wastewater treatment is provided by the City of Austin. Although wells into the aquifers provide substantial water the city supply is filtered ' river water. Other area municipalities and organizations utilize aquifer water. Control of private wells is the f urs t ion of county aad state Health 2-19
SAR 9/84
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F SAR 9/84 EDWARD $ AQU!FER
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SAR 9/84 Departments. Gross beta radioactivity of city water has been measured and is reported in Table 2-4. Table 2-4 GROUND WATER ACTIVITY (gross beta) Travis County 6 x 10 -9 Balcones Research Center 8x 10~9 uCi/ml pCi/ml 2.6' SEISMOLOGY Thirty three earthquakes of intensity IV or greater have had epicenters in Texas since 1873 [10,11]. The earthquakes were characterized using the Modified Mercalli Scale of 1931. 'The scale has a range of I thru XII, on which an intensity of I is not felt, an intensity of III is a vibration similar to that due to the passing of lightly loaded trucks, and intensity of VII is noticed by all as shaking trees, waves on ponds, and quivering suspended objects but causes negligible damage to buildings of good design and construction, and an intensity of XII results in practically all works of construction being severely damaged or destroyed. The strongest earthquake, a maximum intensity of VIII, was in western Texas in 1931 and was felt over 1,165,000 square kilometers. Figure 2-14 shows the locations and intensities of all earthquakes in Texas since 1873. Of these, some are known to have been felt in Austin, but no damage has ever occurred to local buildings. 2.7 HISTORICAL Relocation of the UT TRIGA reactor and related facilities to the Balcones Research Center site is to help accommodate growth of programs both at the University main campus and at the Research Center site. The actual facility location at the Research Center is to replace a concrete and brick lined, 50 meter diameter, tank structure remaining from a magnesium manufacturing plant. The Balcones Research Center site was operated as a magnesium manufacturing plant by the federal government prior to the University's leasing in 1947 and eventual acquisition. Activities at the site were not fully developed by the early 1980's. University functions or research activities were moved to the site when required accommodations were not available on the main campus. A few functions of the University at the site have resulted in the construction of major facilities suitable for long term use. 2-22
SAR 9/84 f f Feb. 1974 March 1948 h- -kh- March 1917 June 1951 C_ mT -l- July 1925 June 1936
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SAR 9/84 3 Other activities at the site have utilized existing structures or other buildings not suited for long term use. In the 1980's, a major program was established to develop the . Balcones Research Center site activities.- As part of the first phase of development, several major research programs associated with energy and engineering were moved to facilities- constructed at the site. Features of the site, before the development activities in the 1980's, are illustrated in Figure 2-15. Several activities at the Research Center prior to 1980 had been associated with radioactive materials. These activities ranged from the burial of low level radioactive waste materials such as tritium and carbon-14 in the northwest corner of the site, to water transport studies performed in 30 meter diameter tanks adjacent and south of the TRIGA facility site. Isotopes of cesium-137, cesium-134, and cobalt-60 were present in sludge samples of the west tank, and are reported in Table 2-5. Gross beta activity in the samples of the west tank measured 22 microcuries per milliliter (1979). Table 2-5 TANK SLUDGE SAMPLES West tank (1979) 22 pCi/ml (gross S) Cs 137
-3 pC1/mi 1.3 x 10 Cs I30 2.5 x 10~0 pCi/mi Co 60 5.7 x 10-5 pCi/mi Radioactive materials at the Research Center site are part of the University broad license for radioactive materials which is managed by the University Safety Office and issued by the Texas Department of Health.
2-24
SAR 9/84 EXISTING CONDITIONS
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} LEGEND Yuw BALCONES RESEARCH CENTER 1980 Figure 2-15 2-25
SAR 9/84 Chapter 2 References
- 1. "Balcones Research Center Project Analysis", Volume I, The University of Texas, 1981.
2.- " Basic Data", City of Austin Planning. Department, June 1980.
- 3. "1980 - Census of Population", Department of Commerce, Bureau of Census, City of Austin Planning Department.
4.- " Local Climatological Data; Annual Summary with Comparative Data 1982", National Oceanic and Atmospheric Administration, Environmental Data and Information Service, National Climatic Center Asherville, N.C.
- 5. "Climatography of Texas; Wind Rose-Austin,. Texas",
National Weather Service, Austin.. Texas.
- 6. " Texas Annual Tornado Density", National Weather Service, Austin, Texas.
- 7. George W. Bomar, " Texas Weather", 1983.
- 8. " Storm Data", 1975-1983, National Oceanic and-Atmospheric Administration, Environmental Data and Information Service, National Climatic Center, Asherv111e, N.C.
- 9. L.E. Carner and K.P. Young, " Environmental Geology of the Austin Area: An Aid to Urban Planning", Report of Investigations No. 86.
- 10. " Earthquake Information Bulletin," May-June 1977 Vol. 9 No. 3, U.S. Department of the Interior Ceological Survey.
- 11. Steven M. Caulson, " Investigations of Recent and Historical Seismicity in East Texas", Masters Thesis May 1984, University of Texas.
- 12. " Project Analysis". Vol. 1, Balcones Research Center.
The University of Texas at Austin, August 1981. i o i 2-26
r SAR 9/84 Chapter 3 TRIGA REACTOR 3.1. DESIGN BASES The reactor design bases are predicted on the maximum operational capability for the fuel elements and configuration described in this report. The TRIGA reactor system has three major areas which are used to define the reactor design bases:
- a. Fuel temperature,
- b. Prompt negative temperature coefficient,
- c. Reactor power.
Of these three only one, fuel temperature, is a real limitation. A summary is presented below of the conclusions obtained from the reactor design bases described in this section. Fuel Temperature The fuel temperature is a limit in both steady-state and pulse mode operation. This limit stems from the out-gassing of hydrogen from U-ZrH (H/Zr ; x) fuel and the subsequent stress produced in the fuel element clad material. The strength of the clad as a function of temperature can set the upper limit on the fuel temperature. Fuel temperature limits of 1150*C (with clad < 500*C) and 970*C (with clad > 500*C) for U-ZrH (H/Zr ; 1.65) have been set to preclude the loss of clad integrity. Prompt Negative Temperature Coefficient The basic parameter which provides the TRICA system with a large safety factor in steady-state operation and under transient conditions is the prompt negative temperature coefficient which is rather constant with temperature (~0.01% 6k/k'C), as described later. This coefficient is a function of the fuel composition and core geometry. Reactor Power Fuel and clad temperature limit the operation of the reactor. However, it is more convenient to set a power level limit which is based on temperature. The design bases 3-1 km. _
n-SAR 9/84 analysis indicates that operation at up to 1900 kW (with an 85 element core and 120*F fnlet water tempccature) with natural convective flow will not allow film boiling, and therefore high fuel and clad temperatuces which could cause loss of clad integrity could not occur. 3.1.1 Reactor Fuel Tempercture The. basic safety limit tor the TRIGA reactor system is the fuel temperature; this applies for both the steady-state and pulsed mode of operation. Two limiting temperatures are of interest, dapending on the type of TRIGA fuel used. The TRIGA fuel which in considered low hydride, that with an H/Zr ratio of less than 1.5, has a lower temperature limit than fuel with a higher H/Zr ratio. Figure 3-1 indicates that the higher hyd ride 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 [1] that the higher hydrides lack any significant thermal diffusion of hydrogen. These two facts preclude concomitant volume changes. The important properties-of delta phase U-ZrH are given in Table 3-1. Table 3-1 PHYSICAL PROPERTIES OF DELTA PHASE U-ZrH Thermal conductivity (93*C - 650*C) 13 Btu /hr-ft 2_.y 6 Elastic modulus: 20*C 9.1 x 10 p,f 650*C 6.0 x 10 psi Ultimate tensile strength (to 650*C) 24,000 psi Compressive strength (20*C) 60,000 psi Compressive yield (20*C) 35,000 psi Heat of formation (6H 298'C) 37.72 kcal/g-mole Among the chemical properties of U-ZrH and ZrH, the reaction rate of the hydride with water is of particular interest. Since the hydriding reaction is exothermic, water will react more readily with zirconium than with zirconium ; hydride systems. Zirconium is frequently used in contact 3-2
SAR 9/84 "O i i i e i i e i i e50 - k 750 - ZrIA) 2r (p)
+
3-HYOR10E ' . P - w" 650 Zr(a) *
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SAR 9/84 with water in reactors, and the zirconium-water reaction 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- Z r il 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 the fuel 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 unciad U-Zril fuel. For the rest of the discussion of fuct temperatures, we will concern ourselves with the higher hydride (H/Zr > l.5) TRICA fuel clad with 304 stainless steel 0.020 in. (0.508mm) thick, or a cladding material equivalent in strength at the temperatures discussed. 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 either radial or normal to the contral axis and do not interfere with radial heat flow. Since the segments tend to be orthogonal, their relative positions appear to be quite stable. The limiting effect of fuel temperature then is the hydrogen gas over pressure. Figure 3-2 relates equilibrium hydrogen pressure over the fuel as a function of temperature for material with three different H/Zr ratios. The hydrogen gas over pressure is not in itself detrimental but if the str se produced by the gas pressure within the fuel can exceedc the ultimate strength of the clad material, a rupture of the fuel clad could occur. While the final conditions of fuel temperature and hydrogen pressure in which such an occurrence could come about are of interest, the mechanisms in obtaining temperatures and pressures of concern are different in the pulsing and steady-state mode of operation, and each mechanism will be discussed independently of the other. In this discussion it will be assumed that the fuel consists of U-Zrli (H/Zr ; 1.65) with the uranium being 8.5 we. % and further that the cladding can is 304 stainless , steel. The clad thickness is 0.020 in. (0.508 mm) with an l inside clad diameter of 1.43 in. (3.63 cm). These fuel parameters have been chosen since they represent the nominal specifications for TRICA fuel elements. Figure 3-3 shows 3-4
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,0 0 600 700 800 900 1000 1100 1200 1300 TEMPERATURE ('C)
EQUILIBRIUM llYDROCEN PRESSURE VERSUS TEMPERATURE FOR ZIRCONIUM-IlYDR00EN Figure 3-2 3-5
SAR 9/84 10 5 ULTIMATE TENSILE 0.2% YlELD C E 5 10 - O
= -
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REFERENCE:
CARPENTER AND CRUClBLE STEEL 10 3 400 500 600 700 800 900 1000 1100 TEMPERATURE (*C) STRENGTH OF TYPE 304 STAINLESS STEEL AS A FUNCTION OF TEMPERATURE t Figure 3-3 3-6
SAR 9/H4 the characteristic of 304 stainless steel with regard to yield and ultimate strengths as a function of temperature. In determining the stress applied to the cladding from the internal hydrogen gas pressure the equation S = P r/t , (1) applies where S= stress in psi, P = internal pressure in pai, r= radius of the cladding can, t = wall thickness of the clad. Then for the cladding we have approximately S = 36.7 P , (2) or the stress applied to the clad is approximately 36.7 times the internal pressure. It is of interest to relate the strength of the cind material at its operating temperature to the stress applied to the clad from the internal gas pressure associated with the fuel temperature. Figure 3-4 gives information as to the ultimate clad strength as a function of temperature and also describes the stress applied to the clad as a result of hydrogen dissociation for fuel having a ll/Zr ratio of 1.65 as a function of temperature. There are several reasons why the gas prennure 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; reabsorption takes place concurrently on the cooler hydride surfacen away from the hot spot; there is evidence for a low permeability oxide film on the fuel surface; and some local heat transfer does take place during the pulse time to cause a less than adiabatic true surface temperature. 3.1.1.1 Fuel and Clad Temperature. The following discussion relates the element cind temperature and the maximum fuel temperature during a short time after a pulse. The radial temperature distribution in the fuel element immediately following a pulse in very similar to the power distribution shown in Figure 3-5. This initial stoop thermal gradient at the fuel surface resultn in some hont 3-7
w-SAR 9/84 i 5 10 ULTIMATE STRENGTH 304 $$ ZrHl .65 10*- 2 E b 103 - 2 10 , , , , , , 500 600 700- 800 Soo 1000 1100 TEMPERATURE (*C) STRENGTH AND APPLIED STRESS AS A FUNCTION OF TEMPERATURE, EQUILIBRIUM llYDR00EN DISSOCIATION PRESSURE Figure 3-4 3-8
7 n SAR 9/84- . transfer during the time of the pulac so that the true peak
. temperature does not quite reach the adiabatic peak temperature. A large temperature gradient is also impressed upon the clad which can result in a high heat flux from the clad into the water. If the heat flux is sufficiently high, ;
film boiling may occur and form an insulating jacket of steam .around the fuel elements permitting the clad temperature to tend to approach the fuel temperature. Evidence has been obtained experimentally which shows that film boiling has occurred occasionally for some fuel elements in the Advanced TRIGA Prototype Reactor located at GA Technologies [2]. The consequence of this film boiling. was discoloration of the clad surface. Thermal transient calculations were made using the RAT computer code. RAT is a 2D transient heat transport code-developed to account for fluid flow and temperature dependent material properties. Calculations show that if film boiling occurs after a pulse it may take place either at the time of maximum heat flux from the clad, before the bulk temperature of the coolant has changed appreciably, or it may take place at. a much later time when the bulk temperature of the coolant has approached the saturation temperature, resulting - in a markedly reduced threshold for film boiling. Data obtained by Johnson et al. [3] for transient heating of ribbons in 100'F water, showed burnout fluxes of 0.9'to 2.0 MBtu/ft 2-hr for e-folding periods from 5 to 90 milliseconds. On the other hand, sufficient bulk heating of the coolant channeled between fuel elements can take place in several tenths of a second to lower the departure from nucleate boiling (DNB) point to approximately 0.4 MBtu/ft 2-hr. It is shown, on the basis of the following analysis, that the second mode is the most likely; i.e., when film boiling occurs it takes place under essentially steady-state conditions at local water temperatures near saturation. A value for the temperature that may be reached by the clad if film boiling occurs was obtained in the following i manner. A transient thermal calculation was performed using ! the radial and axial power distributions in Figure 3-5 and l Figure 3-6, respectively, under the assumption that the , thermal resistance at the fuel-clad interface was : nonexistent. A boiling heat transfer model, as shown in ! Figure 3-7, was used in order to obtain an upper limit for the clad temperature rise, the model used the data of ! McAdams [4] for the subcooled boiling and the work of Sparrow and Cess (5) for the film boiling regime. A conservative estimate was obtained for the minimum heat flux in film boiling by using the correintions of Speigler et al. [6], Zuber [7], and Rohsenow and Choi (8] to find the ! minimum temperature point at which film boiling could occur. This calculation gave an upper limit of 760*C clad temperature for a peak initial fuel temperature of 1000*C, 3-9
,c < F SAR 9/84 i l: l p .- i i 1.3 i i e i i i i I-- j , 1.2 - , r j-1.1 - - [ I 2 e 1.0 - _ 0.9 i. 0.8 O 0.1 0.2 0.3 0. fo 0.5 0.6 0.7 0.8 RADIUS (IN.) RADIAL POWER DISTRIBUTION IN THE U-ZrH FUEL ELEMENT Figure 3-5 3-10
r-SAR 9/84 i ..
'l 1 i i i i i i i 1.0 -
0.9 - - n O w
.0.8 - -
0.7 - - 0.6 - - 0.5 I ' ' ' ' ' ' O I 2 3 4 5 6 7 8 AXIAL OlSTANCE FROM M10-PLANE OF FUEL ELEMENT (IN.) AXIAL POWER DISTRIBUTION IN THE U-ZrH FUEL ELEMENT Figure 3-6 3-11 1
SAR 9/84 6 i i i 10 i i i i i i i l j f%
~
~
\ CURVE BASED ON
\ DATA 0F ELLION N
0 \ - N[ 10 s -
- u. - s
- % _,/
= -
S x - o - CURVE USED d IN ANALYSIS g - 4 5x 10 -
~
' ' ' ' I ' ' ' ' ' ' ' '
3 10 2 10 3 ,g 4 10 10 Tg -T SAT I II SUBC00 LED BOILING HEAT TRANSFER FOR WATER Figure 3-7 3-12
SAR 9/84
.as shown in Figure 3-8. Fuel temperature distributions for this case are shown in Figure 3-9 and the heat flux into the water from the clad is shown in Figure 3-10 In this limiting case, DNB occurred only 13 milliseconds after the pulse, conservatively calculated assuming a steady-state DNB correlation. Subsequently, experimental transition and film boiling data were found to have been-reported by Ellion [9]
for water conditions similar to those for the TRIGA system. The Ellion data show the minimum heat flux, used in the limiting calculation described above, was conservative by a factor of 5. An appropriate correction was made which resulted in a more realistic estimate of 470*C as the maximum clad temperature expected if film boiling occurs. This result is in agreement with experimental evidence obtained for clad temperatures of 400*C to 500*C for TRIGA Mark F fuel elements which have been operated under film boiling conditions [10]. The preceding analysis assessing the maximum clad temperatures associated with film boiling assumed no thermal resistance at fuel-clad interface. Measurements of fuel temperatures as a fun tion of steady-state power icvel provide evidence that after operating at high fuel temperatures, a permanent gap is produced between the fuel body and the clad by fuel expansion. This gap exists at all temperatures below the maximum operating temperature. (See, for example, Figure 16 in Reference 10.) The gap thickness varies with fuel temperature and clad temperature so that cooling of the fuel or overheating of the clad tends to widen the gap and decrease the heat transfer rate. Additional thermal resistance due to oxide and other films j on the fuel and clad surfaces is expected. Experimental and j theoretical studies of thermal contact resistance have been i reported [11-13] which provide insight into the mechanisms ; involved. They do not, however, permit quantitative prediction of this application because the basic data , j required for input are presently not fully known. Instead, several transient thermal computations were made using the RAT code. Each of these was made with an assumed value for the effective gap conductance, in order to determine the effective gap coefficient for which departure from nucleate boiling is incipient. These results were then compared with the incipient film boiling conditions of the 1000*C peak fuel temperature case. For convenience, the calculations were made using the same initial temperature distribution as was used for the preceding calculation. The calculations assumed a coolant flow velocity of 1 ft per second, which is within the range of flow velocities computed for natural convection under various steady-state conditions for these reactors. The calculations did not use a complete boiling curve heat transfer model, but instead, included a convection cooled region (no boiling) and a subcooled nucleate boiling region 3-13
~ --- , .. . . . . .. . -- --
'SAR 9/84
.t
.i PEAK HEAT ONStf 0F NUCLEAft
[ FLUX - .
- ON$tt OF STA$lt ~
/ FILM BOLLING a 1000 - CLA0 INNER -
.' ~ $UAFACE TEMP W - _
?
3 - - y
~
ioo - -
" CLA0 OUTER -
- SUAFACE TEMP i . ,,! , , . .I . ,.,I . . ..I i , ,,
ia o.001 c.oi o.: i.o lo 100 ELAPSEO TIME FROM (N0 0F PULSt (SCC) i j CLAD TEMPERATURE AT MIDPOINT OF WELL-BONDED FUEL ELEMENT Figure 3-8 3-14
SAR 9/84 i i i i i i i 1800 - 1700 ELAPSED TIME FROM END OF PULSE 0.10 SEC , 1600 - 0 SEC p 1.0 SEC [ - E-1500 - a r b - 1400 ,, 3 1300 100 SEC , 1200 - i t i Il I i i 0.5 0.6 0.7 0.8 0 0.1 0.2 0.3 0.4 RADIUS (IN.) FUEL BODY TEMPERATURES AT MIDPLANE OF WELL-BONDED FUEL ELEMENT AFTER A PULSE Figure 3-9 3-15
n :- . , . SAR 9/84 ss 10 6 , , , , , , ,, , , ,, , , , , ,
, ,,~
U ONSET OF PEAK HEAT FLUX -
~
NUCLEATE BOLLING ,
~
ee . C 105
- 'l 3 .
ll ~ O ~ ONSET OF STABLE x - FILM SOILING - 3 w g- - x ^ 1 - b - w y 10 . i . E . e 103 0.1' l.0 10 100 O.001 0.01 [ LAP $[D TIME FROM [ND OF PULSE (SEC) SURFACE HEAT FLUX AT MIDPLANE OF WELL-BONDED FUEL ELEMENT AFTER A PULSE Figure 3-10 3-16
SAR 9/84 without employing an upper DNB limit. The results were analyzed by inspection using the extended steady-state correlation of Bernath [14] which has been reported by Spano [15] to give agreement with SPERT II burnout results within
~
the experimental uncertainties in flow rate. The transient thermal calculations were performed using effective 2
- gap conductances of 500, 375, and 250 Btu /hr-ft F. The resulting wall temperature distributions were inspected to determine the axial wall position and time after the pulse which gave the closest approach between the local computed surface heat flux and the DNB heat flux according to Bernath. The axial distribution of the computed and critical heat fluxes for each of the three cases at the time of closest approach is given in Figures 3-11 thru 3-13. If the minimum approach to DNB is corrected to TRICA Mark F conditions and cross-plotted, an estimate of the effective gap conductance of 450 Btu /hr-ft 2*F is obtained for incipient burnout so that the case using 500 is thought to be representative of standard TRICA fuel.
The surface heat flux at the midplane of the element is shown in Figure 3-14 with gap conductance as a parameter. It may be observed that the maximum heat flux is approximately proportional to the heat transfer coefficient of the gap, and the time lag after the pulse for which the peak occurs is also increased by about the same factor. The closest approach to DNB in these calculations did not necessarily occur at these times and places, however, as indicated on the curves of Figures 3-11 thru 3-13. The initial DNB point occurred near the core outlet for a local heat flux of about 340 kBeu/hr-ft 2 *F according to the more conservative Bernath correlations at a local water temperature approaching saturation. From the foregoing analysis, a maximum temperature for the clad during a pulse which gives a peak adiabatic fuel temperature for the clad during a pulse which gives a peak adiabatic fuel temperature of 1000*C is conservatively estimated to be 470*C. As can be seen from Figure 3-3, the ultimate strength of the clad at a temperature of 470*C is 59,000 psi. If the stress produced by the hydrogen over pressure in the can is less than 59.000 psi, the fuel element will not undergo loss of containment. Referring to Figure 3-4, and considering U-ZrH fuel with a peak temperature of 1000*C, one finds the stress on the clad to be 12,600 psi. Further studies show that the hydrogen pressure which would result from a transient for which the peak fuel temperature in 1150*C would not produce a stress in the clad in excess of its ultimate strength. TRIGA fuel with a hydrogen to zirconium ratio of at least 1.65 has been pulsed to temperatures of about 1150*C without damage to the clad (16]. 3-17
~SAR 9/84 7 i i i i i f ELAPSED TIME FROM
[ 6 - END OF PULSE - 0.247 SEC - E E S ACTUAL HEAT FLUX T 5 - _ S M 5 d 4 _ CRITICAL HEAT FLUX _ 4 E I I I i i 3 7 8 9 10 11 12 13 { DISTANCE FROM BOTTOM OF FUEL (IN.) I I i' SURFACE IIEAT FLUX DISTRIBUTION FOR STANDARD NON-CAPPED (h =500) FUEL ELEMENT.AFTER A PULSE I SnP Figure 3-11' i > f 3-18
. g p_
- m. - . _ _
. . ~ . -
, SAR 9/84 h. p- / i: I- ! :: e
?
- a h
+
f a , , , , 7 CRITICAL HEAT FLUX
.; y -
C. 6 -
'E 3
To 5
;' ACTUAL HEAT FLUX 5
e i. - i - _ ELAP5ED TIME FROM 3 END OF PULSE 15 0.314 SEC A 2 - I f I I f f I 12 13 15 9 10 11
? 8 DIST ANCE FROM BOTTOM OF FUEL (IN.)
f e 4 l- SURFACE ilEAT FLUX DISTRIBUTION FOR STANDARD NON-GAPPED (h gap =375) FUEL ELEMENT AFTER A PULSE Figure 3-12 l 3-19 i
- h
--A,_ _ _ _ _ . . _ _ _ _ _ _ _ _ _ ________m_m_ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _
l SAR 9/84 I l
^
i
.8 i i i i i i
~
7 - CRITICAL HEAT FLUX cT C6 - - a E R e5 - m
'o ~
ELAPSED TIME FROM END OF PULSE IS 0.440 SEC - x 4 - 5 if
;3 . .E ACTUAL HEAT FLUX 2 -
D y i t t i I I I
-7 8 9 10 11 12 13 14 15 DISTANCE FROM BOTTOM OF FUEL (IN.)'
LC78709 .. SURFACE HEAT FLUX DISTRIBUTION FOR STANDARD, NON-GAPPED (h gap =250) FUEL ELEMENT AFTER A PULSE Figure 3-13 , 3-20 s
SAR 9/84 jo6 , , , , , , , , EFFECTIVE HEAT TRANSFER --
- COEFFICIENT IN GAP, 4
BTU /HR-FT2 ..y 500 375 250 _ J N[ 105 _ t _ E _ s, _ m _ E d a _ y _ k-w N in 0 D
- 10 FLOW VELOCITY = 1 FT/SEC _
GAP THERMAL RESISTANCES ARE REPRESENTATIVE OF CONDITIONS AT - END OF PULSE (1.E. TIME - ZERO) 10 3 1.0 O.01 0.1 ELAPSED TIME FROM END OF PULSE (SEC) SURFACE HEAT FLUX AT MIDPOINT VERSUS TIME FOR STANDARD NON-GAPPED FUEL' ELEMENT AFTER A PULSE Figure 3-14 3-21
SAR19/84' 3.1.1.2 Finite Diffusion Rate. To-assess the effect of the finite diffusion rate.and the rehydridingfat the cooler
; _ surfaces, the following analysis is presented.
As' hydrogen is ' released from the hot fuel r e g i o n n ~, it is taken up11n.the cooler regions and the. equilibrium that is obtained is characteristic of some-temperature lower than thefmaximum. To evaluate this reduced, pressure, we will use
. diffusion theory to calculate the rate at which hydrogen is evolved and. reabsorbed at the fuel surface.
Ordinary diffusion . theory provides an expression for
- describing the time dependent loss of gas from a_ cylinder:
~
C t f- = f exp - n , (3) 2 2 "i ~ "f n= 1; Z r i. n 0 where E, e t, c g, = the average, the-initial, and the final gas concentration in the cylinder, respectively, ! 1 Z = the roots'of the Equation J0(x) = 0, D '= the diffusion coefficient for the gas in the cylinder, r - = the radius of the cylinder,- 0 t = time. Setting the term on the right-hand side of Equation 3 equal to e, one can rewrite Equation 3 as: E/c t = eg/c1 +. (1 - c /ct) f e , (4) and the derivative in time is given by d(c/et) = (1 - c g/ct) de . (5)
-dt dt This represents theffractional release rate of hydrogen,from p the cylinder,- f(t). 'The derivative of the series i n' the right-hand-side of Equation 3 was. approximated by E = -
(7.339e -8.34c + 29.88e-249c) Jjc i , (6) de d t- , 2
'where'c = Dt/r ,
e
- l. 3-22 F'
- ~ ~
SAR 9/84 V
-The diffusion coefficient for hydrogen' in zirconium r ;
'~
-hydridejin'which the H/Zr ratio is between 1.56 and 1.86 is-
;giv en. . by '
~
DL= 0.25.e-17800/R(T+273) (7) where R.= the-gasJcon'stant and, T'= the zirconium hydride temperature in *C. EquationE3. describes-the escape of gas from a cylinder through diffusion until' some . final concentration -is achieved. Actually, in _ the closed system considered here, not on'ly does the hydrogen diffuse into the- 'f uel-clad o gap , bu t. .also it-diffuses back into the fuel in the regions of
' lower _ f uel temperature. The gas also diffuses through the
_ , clad'at a. rate dependent on the clad temperature. Although E
^
.this tends - to reduce _ the hydrogen pressure,- it is not considered in' this . analysis. When the diffusion rates-are equal, an equilibrium condition will exist. To account for this, . Equation 5 'was _ modified by substituting- for the.
concentration ratios the ratio of.the hydrogen' pressure in
.the' gap _to the' equilibrium hydrogen pressure, P /P,. Thus,
'f(t) = (c/cg) =
(1 - P (t)/P,) d5, (8) dt. dt 3where.Ph(t) = the hydrogen pressure, a function of time and P, = the equilibrium hydrogen p^r' essure over the zirconium hydride which-is a function'of the fuel-temperature. The rate of-change of the internal hydrogen pressure, in psi, inside the fuel element cladding is' dP h - *
- h. 22.4 i+273 , (9) dt y 6~.02 x 10 23 &
273 where N h= the number of molecules of H 2 I" th* f"*1' T= the gas temperature (*C), e
=
i f(t) 'the fractional loss rate'from Equation 8 ~, j V = the free; volume ~inside the fuel clad (liters). g As the atom density of hydrogen in ZrH (H/Zr ; 1.65) is-about 0.1-moles and'the fuel volume is 400 cubic cm., Nh is ~ 19.9fmoles (H2 ). The free volume is assumed to consist of a
' cylindrical volume,_ at the' top of'the element, 1/8-in. high
' ' with a diameter:cof 1.43 in. for a total of 3.3 cubic cm. Also~, the temperature of1the hydrogen in the gap was assumed to be the .-temperature of the clad. The effect of changing f
. 3-23
[ -_b-__'_.'._-_a___.____mmm.x.__.__
e
" 1 SAR 9/84
, ' these Lewo- assump'tions f was teste'd : by f ealculat. ions . in which the-gapLvolume was decreased by 90% and _ the. t empe ra tu re. of
- the' hydrogen.in'the gap'was set.up equal.to the maximum fuel
. temperature. . Neither of these changes resulted- in _ maximum-
_ pressures- different from._ those based on the. . original
. assumptions although1 the initial- ~ rate - of pressure increase wasigreater. For these conditions Pg = : 7. 29 x 10 (T +'273) f ( t ) - d't . (10)
The -fuel temperature used in Equation 7:to. evaluate the
. diffusion ~ coefficient is expressed as
'T(z): = T O ;t 0 , ..
(11) T(z) =
~ TO + 'I m' 0) c s L [ 2.'4 5 (z-0. 5) ] ; t 2 0 ,
Ewhere T,_= the, peak fuel temperature-(*C), T0;=.the clad temperature (*C), z = the. axial distance expressed as - a fraction of the fuel length, t = the time after step increase in p o we'r .- It was assumed that'the fuel-temperature was invariant-with -radius. The hydrogan pressure over the zirconium
. hydride surface when equilibrium Lprevails- is strongly. l temperature dependent as shown in Figure ' 3-2 and , f or _ ZrH _
.H/Zr
( ; 1.65), can be~ expressed by P, = 2.07 x 10 9 e
-1.974x100/(T+273) ,
(12). E The coefficients have been-derived-from data, developed by Johnson. .-The . rate at which - . hydrogen is. released or reabsorbed. takes the. form x g(t,z)' = [P,(z) - Ph(E)I f(t,z)- , (13) P,(z) where f(t,z) = the derivative given in Equation 8 with' respect to time evaluated at the-axial
-. position z, Ph(t) = the hydrogen pressure in the gap at time t,
( P,(z) = the-equilibrium hydrogen pressure at the ZrH temperature.at position z. The internal / hydrogen pressure is then 1 Y 3-24
~.y -
s SAR 9/84 l-t 1 . 3 P - h ( t')7 = ? 7. 2 9 : x 10 - (TO + 273) <0 g(t,z) dz dt.(14)
<0 This :. Eq'ua t ion, was approximated by 3
Pht{)-=7.29x10 (T0+ 273) x { { gg_ PH( i-1) i=i j=1. E(*j) x f'(tj,z )'6z 6t , ,(14) where'the internal summation is over the fuel element length increments and the external' summation is over time. For' the case in which the maximum fuel temperature is l 1150*C, the equilibrium hydrogen pressure in ZrH (H/Zr ; 1.65) is 2000 psi. Calculations indicate, however, that the internal pressure increases to a peak at about 0.3 sec, at-
-which ' time the pressure. is. -about one-fifth of .the equilibrium value or- abut 400. psi. After this ' time, .the pressure - slowly decreases as the hydrogen continues :to ' be redistributed along. the . . length of-the element from the hot y ^ regions to-the cooler regions.-
Calculations have also been made for step increases in power to peak fuel temperatures greater _than:1150*C. -Over a 200*C range, the time to the peak pressure and the fraction
~
o f -. the equilibrium pressure value achieved were approximately the same as for the 1150*C case. Thus, if'the ~ clad remains below . about 500*C, the internal pressure. that would produce the. yield stress in the clad (35,000 psi) is about- 1000 psi and the corresponding equilibrium - hydrogen , l . pressure - corresponds to a' maximum > fuel temperature of about
-1250*C in ZrH. (H/Zr ;_ 1.65)..
Similarly, an internal pressure of- 1600 psi would . p roduce' a stress equal to'_ ' the j' ultimate. clad' strength (over 59,000 psi). This corresponds l to' an equilibrium hydrogen pressure of 5 x.1600 or 8000 psi L. . and: a fuel temperature of about 1300*C.
-Measurements of hydrogen pressure in -TRICA -fuel p elements during': s teady-s' tate . operation have not been made.
However, measurements have been- made during transient L, operations . and ' compared with . the - results .of' an analysis [ .similar- to :that described here. These. measurements , ~ indicated that in a pulse-in which the . maximum temperature; F :in the' fuel was greater than 1000*C the maximum pressure was
'only about 6% of'.the equilibrium value evaluated at'the peak i;
- temperature. Calculations of the . pressure resulting from I such
- a pulse using the methods described above gave calculated pressure values about three times greater than the measured. values.
[L L r
~
b 3-25
- e. -
SAR 9/84 An instantaneous increase in fuel temperature vill produce the most severe pressure conditions. When a peak fuel. temperature of 1150*C is reached by increasing the power over a finite period of time, the resulting pressure will be no greater than that for the step change in power analyzed above. As.the temperature rise times become long compared with the diffusion time of hydrogen, the pressure will become increasingly less than for the case of a step change in power. The reason for this is that the pressure in the clad element results from the hot fuel dehydriding faster than the cooler fuel rehydrides (takes up the excess hydrogen to reach an equilibrium with the hydrogen over pressure in the can). The slower the rise to peak temperature, the lower the pressure because of the additional time available for rehydriding. 3.~1.1.3 Summary. The foregoing analysis gives a strong indication that the clad will not be ruptured if fuel temperatures are never greater than in the range of 1200*C to 1250*C, providing that the clad temperature is less than about 500*C. However, a conservative safety limit of 1150*C has been chosen for this condition. As a result, at this safety limit temperature the pressure is about a factor of 4 lower than would be necessary for clad failure. This factor of 4 is more than adequate to account for uncertainties in clad strength and manufacturing tolerances. Under any condition in which the clad temperature increases above 500*C, the temperature safety limit must be decreased as the clad material loses strength at elevated temperatures. To establish this limit, it is assumed that the fuel and the clad are at the same temperature. There are no conceivable circumstances that could give rise to a situation in which the clad temperatures was higher than the fuel temperature. In Figure 3-4 there is plotted the stress imposed on the clad by the equilibrium hydrogen pressure as a function of the fuel temperature, again assuming a clad radius of 0.73 in. and a thickness of 0.02 in. Also shown is the ultimate strength of 304 stainless steel at the same temperatures. The use of these data for establishing the safety limit is justified as
- a. the method used to measure ultimate strength requires the imposition of the stress over a longer time than would be imposed for accident conditions,
- b. the stress is not applied biaxially in the ultimate e strength measurements as it is in the fuel clad.
The point at which the two curves in Figure 3-4 intersect is the safety limit, that is, 970*C. At that temperature the 3-26 w
7
- g. SAR 9/84 m:
equilibrium ~ hydrogen pressure would impose a ' stress on the
- clad equal to'the ultimate strength of the clad.
The same argument about the redistribution of the hydrogen within the fuel presented earlier is valid for this
. case _ also. In addition, at elevated temperatures the clad becomes quite permeable to hydrogen. Thus, not only will
' hydrogen redistribute itself within the fuel to reduce the pressure, but also some hydrogen will escape from the system entirely.
The use of the ultimate strength of the clad material in the. establishment of the safety limit under these conditions is justified because of the transient nature of such accidents. Although the high clad temperatures imply sharply reduced heat transfer rates to_the surroundings (and consequently longer cooling times), only slight reductions in the fuel temperature are necessary to reduce the stress sharp 1y. A 50*C decrease in temperature from 970*C to 920*C will , reduce the stress by a factor of 2. As a safety limit, the peak adiabatic fuel temperature to be allowed during transient conditions is considered to be 1150*C for U-ZrH 1.65* 3.l.2. Prompt Negative Temperature Coefficient The basic parameter which allows the TRIGA reactor system to operate safely with large step insertions of reactivity is the prompt negative temperature coefficient associated with the TRIGA fuel and core design. This temperature coefficient (a) also allows a greater freedom in ! steady-state operation as the effect- of accidental reactivity changes occurring from the experimental devices in the core is greatly _ reduced. 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 a 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 neutron in the fuel element 3-27 i
'SAR 9/84 'will gain energy from an excited state of an oscillating hydrogen atom in the lattice. As the neutrons gain energy from the ZrH, their mean free path is' increased appreciably.
This. .i s , shown , qualitatively in Figure 3-15. Since the average chord length in the fuel element is comparable with a mean free path, the probability of escape from the fuel element before capture is increased. In 'the water the neutrons are rapidly rethermalized so that the.. cap ture 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 spectrum to harden 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. 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.
3.1.2.1. Codes Used for Calculations. Calculational ' work on~ the temperature coefficient made use of a group of ' codes developed by GA- Te'chnologies: GGC-3 [17], GAZE-2 [-18), and GAMBLE-5 [19), as well as DTF-IV [20], an S multigroup transport code written at Los Alamos. Neutron cross- sections for energies above thermal (>l 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 B equations for.cach discrete reactor region composition. Yhis 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 Nordheim ( 21] - 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 3-28 c -
SAR 9/84 10 0 g 80 - e p 400'c u r E Go - 23*C y h3
" n: +
O w 40 - ab E5 5E 20 - 0 m i I i , . . I , , , , 0.01 0,1 g,o NEUTRON ENERGY (eV) TRANSPORT CROSS SECTION FOR HYDROGEN IN ZrH AND AVERAGE NEUTRON SPECTRA IN FUEL ELEMENT Figure 3-15 3-29
.. . .-. . . . . . . _ - - .. .-. , _ . - . - . ~ . . -- . -.
m SAR 9/84 the neutrons with the chemically bound moderator atoms. The bound hydrogen kernels used for hydrogen in the water were generated by the THERMIDOR code [22] using thermalization
. work of Nelkin -[23]. Early thermalization work by McReynolds et al [24] on zirconium hydride has been greatly extended at GA Technologies [25], and work by Parks resulted in the SUMMIT [26] 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 Figure 3-16 and. Figure 3-17 [27].
3.1.2.2. ZrH Model. 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 ~0.14 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 eV, and whose width is adjusted to fit the cross section data of Woods al. [28]. The low frequency acoustical modes are assumed e_ t_ 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 the transition 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 1attice 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. 3.1.2.3. Calculations. Calculations of the temperature coefficient were done in the following steps: 3-30 L
SAR 9/84 6 , ,, 10 _ , ,, , , , , , , , , ,, , T= 316 *C
.-:::::a T.c;;.;~.=.;--~ . . . . . , _
/, - s. - .. .
10 5 - ---- W#! T;i"""" T
, /.hT
., =-23,2*C . - - . -
- --u - - - - .
m
/
/ / .,/ - m. .
t- f / .-* ' T = 150'C _ Z
~
/ r~g D 4 / / .=- %
10 / ' ,'T= 30*C \ 7 3 rf m - / -% 4 / / ~.
~ -
m
/ ..:. . ~
l- //. s 5 -
/ '...
x / .
< 103 -
n w - -
-6 N. -
z ,. .. - io _ l. 3 ' ' I ' ' I ' ' ' ' ' ' 10 O.OOI O.01 0.1 1.0 10.0 10 0 NEUTRON ENERGY (EV) A COMPARISON OF NEUTRON SPECTRA BETWEEN EXPERIMENTS AND SEVERAL HYDROGEN MODELS l Figure 3-16 - 3-31
SAR 9/84 lo s , , ,,,,g, , , _ , , , , , , , , , ,,,,n, , , , on;
- ROOM -
TEMPERATURE - 5 10 ;_ , 150*C 10 4 - m : . : I-- - $. : 2 o s - s x . 316'C . '. . s m go3 _ H : . 55 : \.~ : x .
~
4 . S., .- G _' 468'C .a
- ~
., [
g .
-s-102__
/ .
. g -
. 3 e, .
ZrH.75l BORON POISONED
- 3.4 B ARNS/ HYOROGEN ATOM 10' :--
- . ., * ~
***** OATA _-
~
~
EINSTEIN OSCILLATOR . MODEL INCLUDING ACOUSTICAL , TRANSITIONS ., O . ..4...i . . . . ..I - i....I i . . . . . . . O.0 01 0.01 0.1 1.0 10.0 NEUTRON ENERGY (EV) EFFECT OF TEMPERATURE VARIATION ON ZIRCONIUM HYDRIDE NEUTRON SPECTRA Figure 3-17 3-32
SAR 9/84 Multigroup cross sections were generated'by the a. 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 U-238' 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 P g approximation. . Comparisons have shown Sg and S a results to be nearly identical. Group dependent disadvantage factors were calculated for each cell region (fuel, clad, and water) where the disadvantage factor is defined as the ratio: 4 # / 4" (region / cell). E 5
- 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.
- d. The cross sections for an equivalent homogenized core were used in a full reactor calculation to
~
determine the contribution to the temperature
. coefficient due to the increased leakage of thermal neutrons into the reflector with increasing hydride temperature. This calculation'still requires several-thermal groups, but transport effects are no longer of major concern. Thus, reactivity calculations as a function of fuel 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 U-238 resonances and temperature dependent leakage from the. core. These effects produce a temperature coefficient of -0.01%/*C, which is rather constant with temperature. The temperature coefficient is shown in Figure 3-18 for the high-hydride core of this TRIGA. 3.1.3. Steady-State Reactor Power
'The following evaluation has been made for a TRIGA system operating with cooling from natural convection flow {
around the fuel elements. This analysis investigates the i limits to which such a< system may be' operated. i 3-33
SAR 9/84 The analysis was conducted by considering the hydraulic characteristics of the flow channel from which the heat rejection rate is maximum. The geometrical data from this channel are given in Table 3-2. All symbols in Equation 16 through 45 are defined in the list of nomenclature in Section 3.1.3.9. Table 3-2 HYDRAULIC FLOW PARAMETERS Flow area (ft /elem.) 0.00580 Wetted perimeter (ft/elem.) 0.3861 Hydraulic diameter (ft) 0.0601 Fuel element diameter (ft) 0.1229 Fuel surface area (ft 2) 0.4826 The heat generation rate in the fuel element is distributed axially in a cosine distribution chopped at the end such that the peak-to-average ratio is 1.25. The number of fuel elements in the core is assumed for 1 MW operation, but .the departure from nucleate boiling (DNB) ratio is conservatively evaluated on the basis of 85 elements. The driving force is supplied by the buoyance of the heated water in the core. Countering this force are the contraction and expansion losses at the entrance and exits to the channel, and the acceleration and potential energy losses and friction losses in the cooling channel itself. Figure 3-19 illustrates schematically the natural convection system established by the fuel elements bounding 1 one flow channel in the core. The system shown is general ' and does not represent any specific configuration. Steady-state flow is governed by the Equation n 6p + 6p, + 6p1 + 6P u + (16) i [ wPj " 2 t /V o , j=1 where the inft-hand member represents the pressure drops through the flow channel due to entrance, exit, friction, acceleration, and gravity losses and the right-hand member 4 represents the driving pressure due to the static head in the pool. The pressure drops through the flow channel are
?
3-34
SAR 9/84 4 i' J l -14 h STAINLESS STEEL CLAD g -12 - 8.5 WT-% U-ZrH .60 l CORE
,9 5
e _io Z
- i. E
. u E -8 W O O y -6 - 3 k E E -4 - 2 W b-t -2 - 2 O i E Q. I I I I I O 0 200 400 soo 800 1000 /200 TEMPERATURE (*C) i PROMPT NEGATIVE TEMPERATURE COEFFICIENT VERSUS AVERAGE FUEL TEMPERATURE FOR TRIGA Figure 3-18 3-35
- - _ . - , ~ . . . . . - - - . - . . . - _ . . - , . . . - - _ _ . , . . . - - - . - - . - - - - _ - , , __
c - SAR-9/84
, CHANNEL' SURFACE TO
- VOLUME RATIO S/V CALCULATED FLOW AREA, Af -
FROM GIVEN . HEATED PERIMETER, P DIMENSIONS EQUIVALENT DIAMETER, D e '-
/.
/ \
FREE SURFACE OF POOL
, VIEW A-A /
n\/n b z = zt #
- 0
- A.l a u H z =z n+2 - { - -- - -
gp p= +Pamb k = n+1 j=n p-T sat
, L o
z P0OL AT CONSTANT ! q" = q"(z) TEMPERATURE, T, j (GIVEN) l I j=1 -[ k=1
/
z = zt opg CROSS FLOy, OR i FLOW BETWEEN z=0 ADJACENT CHANNELS, l q a { apia- ,' _ IS IGNORED M A api * [// (COOLANT INLET H0LE OF AREA A I EL-0581 j GENERAL FUEL ELEMENT CONFIGURATION FOR ! SINGLE COOLANT CHANNEL IN THE TRIGA , Figure 3-19 l s' 3-36
Wn -
, .S AR ,9 / 84-
' dependent '.o'n .Le h e'. flow- rate; while th'e available stat'ic driving: pressure is. fixed for a'. known : core height and pool temperature. .The analysis, therefore, ,becomes an iterative-1
-one sin:which th'e left-hand side of Equation 16 is evaluated on the' basis of.an' assumed flow' rate and; compared with the known =right-hand side until equality is achieved. The
, . method has . been programmed for digital. . computer solution.
~ The; methods -of evaluating;each of the 6p terms in Equation 16' for known _ power. distribution .and flow geometry and assumed flow-rates are_ discussed below.
3 .1. 3 .1 ~. Entrance-Loss, 6p . The entrance loss. 6 p g ,' s may be evaluated in the ; usual way -as a f raction of tRe-velocity-head in the lower: grid place hole: k -+ k y - 1 1
- 6p g = 1 2 (NW) ,
(17) 2g A
.where N =_the~ number of channels which receive their flow from a single hole in'the-. lower grid plate, k
i t = the loss factor for the entrance to _the hole in thezlower grid plate. For even slight rounding.of the entrance, k gg will-be no' greater than O.30, k 12 - the loss factor covering transfer of the flow from the hole in the' lower. grid _ plate to the coolant channels. In most cases' this can be satisfactorily approximated-as a sudden expansion using~kg= 1.0. 3 .1 ~. 3 . 2 . Exit Los's, 6p,. The. exit loss is expressed in terms of a- coef ficient K which is the fraction of. the velocity head in the flow channel which is'not recovered: E Y 6p, = e n+1 W2 . (18) i 2 2gA The term v station - a l,o# ng 1 is tihe specific volume at the highest axial 4 the heated length of the core. It is evaluated f rom the temperature T, g which is obtained from an overall.-heat balance: T,,g = 9tM & T, , (19) .l
*n+1 where q _e
= P q"(z) dz .
z g
+
F )- i 3-37 i y, ,
a --4 4 SAR 9/84 s
- 3.1.~3.3. Loss Through Portion of Channel Adjacent to.
Lower Reflector,- 6p flow -is isothermal ~.at the bulk
~ -pool-temperature so k..~The hat I V 08 2 6z l 6py -- m' o 1W , . (20) s2g D, A v 9
f, isl evaluated, from the Moody chart (assuming ' smooth surface) on.the. basis of a Reynolds number _which is Y R f= e-oW . (21) Ag v o 3.1.3;4. Loss Through-Portion of Channel Adjacent to Upper Reflector, 6p . The ' flow is is o th e rmal- ..at- T n+1 where T,,y is determine 8 by Equation 19 : m n 0* uW2 , 6z u I V ' 6p, = . (22) 2g De A$ vn is.again evaluated from the Moody chart, assuming smooth f, surface, on the basis--of a Reynolds number which is R, = D , v n y-. (23) Ag v 3.1.'3.5. Loss Through Each Increment of the Channel' Adj acent to- the Fueled Portion of the Elements, 6p,. For the present, let us assume that the entire heated portion of
-the channel is in subcooled boiling. This implies that the wall temperatures _ calculated from subcooled boiling correlations-are lower than those calculated for convection c alone and that the liquid is below its saturation' -
temperature at all locations. The pressure drop through an
' increment'is given by V V 6z m -V m 2 'b 5 m 2* 6z 6pn -(n+1) ' I # 5 - (
2 2 Y gA 2g A D' m) (acceleration) (friction) (gravity), 3.1.3.6. Acceleration Term. v denotes the mean specific volume and is larger than %he liquid specific c volume, because of the vapor voidage:
'v, = v/(1-a)- .
(25) a is the void fraction or the fraction of a channel cross section- which is _ occupied by vapor, a may be calculated
.f rom the. vapor volume (cubic in, vapor / square in. heating
_I 3-38 i.
SAR 9/84^ surface) and the flow channel: geometry. Denoting the vapor volume as (, a =( (S/V) (26) where S/V is- the surface to volume ratio of the coolant channel. The parameter (, is_ dependent on the surface heat flux. -the subcooling of liquid and the velocity of the liquid. It can be evaluated only by experiment. Data-given by Jordan and . Leppert -[29] were used to estimate (; these data are plotted. in ' Figures 3-20 and . 3-21. Most of this represents a flow velocity of.4 ft/sec and appears to be the only.~available data. applicable under the thermal conditions encountered in TRIGA type reactors _ Extrapolations' from these data are made for flow velocities different from 4 ft/sec. The extrapolations were based on a small amount-of data given for flow velocities other than 4 ft/sec. The liquid temperature at a station, T k, may be calculated from a z P q"(z)dZ T <
+
T, (27) k= 1 WC Therefore, one finds ( (Figure 3-21) from T,,g -Tk ""d 9k ' where T,,g and qk are known. Since- a k"Ek (8/Y) and v k Equation 25. is a function of T gg v, may be evaluated from 3.1.3.7. Friction Term. v denotes a linear average of. the mean specific volumes"d at the upper and lower boundaries.of an increment. The approximate mean value is assumed to apply over the entire increment so that v v, , m k + y"k+1 . (28) j 2 A friction factor f b is applied locally to calculate the friction pressure droh over the increment in subcooled boiling. Jordan and Leppert develop the correlation 8 9" fb=8St = p CV b= p CV (T y - T) (29) and provide experimental verification near atmospheric pressure in the range 0.0015 < S < 0.0050. This is simply t. an extension of Reynolds' analogy to the case of subcooled boiling. The . Equation of continuity is used to write Equation'29 as ! 3-39 i
SAR 9/84 to i i i i i i
~
SUSC00 LING, (T sat - T)= 18'F 48'F 10-2 _ g - - f - 18*F - j 108'F O . - u w E o, . 10'3 -
~
PRESSURE = 16.4 PSIA
~
FLOW VELOCITY = 4 FT/SEC 10' I I . I I f 0 2 4 6 8 to 12 14 2 , HEAT FLUX, q" (81U/HR.FT X 10 53 4 EXPERIMENTALLY DETERMINED VAPOR VOLUMES FOR SUBC00 LED BOILING IN A NARROW VERTICAL ANNULUS Figure 3-20 3-40
- ~ . . . . . ..
'SAR 9/84 '
-10*I e i i e i i e- ~
~
k 10 2 - g - 5 u s qrt X 10* 5
% 5.00
' 4.75 yg.3 _. 4.50 -
- 4.25 ~
4.00 3.75 . ' 3.50 i
- 3.25 ~
3.00 2.75 2.50
~
2.25 ~ 2.00 1.75 1.50
- I I t I 1.25 10,~ 4 i 0 20 40 60 80 100 t20 I40 SUBC00 LING, T,,, . T CROSS PLOT OF Figure 3-20 USED IN CALCULATIONS Figure 3-21 3-41
- _-, _. _ ~ . . . - . . . - . _ , . - , , . . - . . , , , _ - - , - - . . - . - . - . . - - . - - . . . - . . . . - . . . . . - . . _ . . - - . - . - . , -
SAR 9/84 A f b
= f (30)
WC (T - T) which may be evaluated if T is known. For subcooleo boiling... the heat transfer Ys usually defined by an experimentally determined correlation of q" vs (T - T which has been obtained over a given range of f low" velo $ f E y) and. pressure. McAdams [30] gives such a correlation for pressures between 2_ and 6: atmospheres and flow velocities , _between 1 and 12 ft/sec. This correlation will be used to determine T y for use in Equation 30. Approximate mean values are assumed to apply over the entire increment so that. f b l/2 ^f k + 9 k+1 j , (31) [ -T wk k -T Wk+1 k+1 and T -T sat
= '( "k} '(9 k+1} ,
where 4(q") is the correlation of McAdams previously cited. 3.1.3.8. Gravity Term. The gravity term is evaluated from=v) calcuated from Equation 28. As implied in Section 3.1.3.5., each increment must be checked to determine whether heat is being transferred by subcooled boiling or by convection. T is evaluated at the lower . boundary of the increment on the basis of both the correlation from McAdams for subcooled boiling and a standard correlation for convection (Dittus-Boelter). If the T calculated from. convection correlations is less than that Ebtained for subcooled boiling, boiling is assumed not to be present in the increment. Equation 24 still applies, but since there is no boiling and hence no vapor void, v becomes y and f
- 3 becomes f,.
In the foregoing analysis an assumption was made that all of the vapor formed on the surface _of the fuel element detaches and adds to the fluid buoyancy. This is not a conservative assumption. The position where vapor bubbles first leave the heated surface is obtained from two considerations; first, the balance of the forces exerted on
.the vapor bubble while it is in contact with the wall (buoyancy, surface tension, and friction), and, second, the temperature distribution in the single phase liquid away ,
from the walls. Determinitation of the buoyance forces resulting from l the formation and subsequent detachment of vapor bubbles is complicated by the difficulty in predicting the point at ! which the vapor detaches, and the fraction of that vapor l l 3-42 L
SAR 9/84
'which subsequently condenses. The problem was simplified by making _ use of an analysis performed by Levy [31] to determine-the position at which the vapor detaches from the wall, assuming that at that point all of the vapor detaches -and, finally, that there is no recombination of the vapor with subcooled fluid.
According to Levy the position at which the- vapor leaves the-surface is obtained from considering the balance of forces exerted on the vapor bubble while it is in contact with the ' wall, and -the temperature distribution in the single phase liquid away from the wall. The forces acting on the bubble in the vertical direction consist of a buoyant force, F a frictional force, F y; exerted by the liquid on the B; bubble; and a vertical component of the surface tension force, F g. The buoyant force, F B, is given by 3 F " b # ' ( } B E c where r is the bubble radius, CB is a proportionality B constant, p and p t y are the liquid and vapor density, g is the acceleration due to gravity and g is a conversion ratio from lb-force to lb-mass. The frictional torce, Fp, -is related to the liquid frictional pressure drop per unit length, The pressure differential across the bubble is(-dp/dz)k. propor ional to the pressure differential times the bubble radius and it acts across an area proportional to the square of the bubble radius. Relating the pressure differential to the wall shear stress T, by
-(dp/dz), = 4 T /D H , (33) there results for F 7:
Fp=Cp [w r B D H where C is a constant of proportionality and D H is the hydrau12c diameter (four times the cross-sectional area divided by the wetted perimeter). The surface tension force, F 3, is given by F3=C3 r B o , (35) where C 3 is a proportionality constant and o is the surface tension. Assuming upward flow the balance of these forces ; results in the following solutions for the bubble radius: 3-43
o , it SAR 9/84 C o 1/2 r 3 =g s
)
(36) C + " B E ("L -E v F E D c H Assuming-that the distance from the' wall to the tip of- the
~ bubble is proportional ,to the bubble radius, a non-dimensional distance corresponding to this real distance can be given by y- . (o s e D H -O L)1/2 [1 +C 'E (P L -Ov) H]~I!2 (37)
B C "L E c T w where C and C' are appropriate constants. For those cases where the fluid forces are considerably greater than the buoyant forces this expression reduces to Y =C (a g, DH 3 -8L) 1/"L - (38) For the bubble to detach, the fluid temperature at the tip of the bubble must exceed the saturation temperature by an amount such that the pressure differential acting across the interface at the'tip of the bubble balances the surface _ tension forces at the same position. By using the Clausius-Clapeyron solution of this ' pressure differential one finds that the fluid temperature-saturation temperature difference can be assumed to be zero. The temperature at the tip of the bubble can also be specified f rom ~ existing solutions for the fluid temperature distribution. Thus, if the flow is assumed to be turbulent, and using the solution proposed by Martinelli, we have T, - TB = 0P r Y B
; O 5Y B 5 5 (39)
= 50 (Pr + In [1 + Pr (YB /5 - 1)]) ; 5 5Y B 5 30 g
= 50 (Pr + in [1 + 5 Pr] + 0.5 in [Yg/30]} ;
YB> 30. The parameter 0 is a non-dimensional term defined through the heat flux and liquid specific heat, that is, 0= q/A
. (40) og C (T,g /p )1/2 Levy obtained values for the constants C and C' by correlation with available experimental data. Using the accepted heat-transfer relation from Dittus-Boelter, one obtains i hDH /kg = 0.023 (WDH/NL)
(Pr)0.4 . (41) 3-44 ~__ . _ _ _ _ _ _ _ - _ _ - - -
l{; . SAR,9/84-4
-Calculating the friction' factor from-6 p{)]I! },-(42) 4 ,
f = 0.0055(1 +~[20,'000(c/Dy) + 10 /(WD H
. .weiare able to find the wall shear stress from T, = (f/8) (W 2j 8) , (43) c
' The - correlation' with experiment . yielded values for the constants of
;C = 0.015 , (44)
C' =0 . Finally, from the definition of the heat transfer-7 coefficient, one-obtains T,- T = q"/h , (45)-
- and . setting the- bubble tip temperature, T B' '9"81
- th' saturation' temperature, T **, we can express the relationship .between t h e . s a t'u r a t i o n temperature, ,the wall.
temperature, and the fluid temperature at which the. bubble would detach from the wall by (T,-T,,g)/(T,-T) = 0.023 (WD H/p{) (Pr)-0.6(f/8)-0.5 g, ,
'(46) where.0 = Pr Y B ; O
<YB< 5
=5 (Pr + in [1 + Pr (0.2 Y B
- 1)]) ; 55Y B
< 30
-5 (Pr + In (1 + SPr) + 0.5 in (Y /30)} ; Y B 2 30..
B The solution of the force balance equation with void detachment was accomplished by iterating on the. void detachment point to find where the right and left sides of' Equation 46 were equal. The point at which' the : void 'was assumed to seperate from the surface.was taken as the point at which equality obtained. The peak heat flux, that is, the heat flux at .which there -is a departure from nucleate boiling and. :the
- . transition to film boiling begins, was determined by two.
correlations. The first, given by McAdams .[32) , indicates that the peak heat flux is a function of the fluid velocity.- and the fluid only. The second correlation is due to Bernath [33]. It encompasses a wider range . of ' variables o v e r.. which the correlation was made and it takes into account the effect of different flow geometries. It genera 11y'gives a-lower value for the peak heat flux and11s the value used here for determining the minimum DNB ratio, that is, the minimum ratio of the local allowable heat flux 3-45 L
r , 3 , t , SAR:9/84
? , , to thel actual- heat flux. In general, the McAdams correlation gives a DNB ratio 50% to 80% higher than the Bernath correlation.
Figure'3-22 shows the results - of - this. analysis. 'Here we have plotted the maximum channeliheat flux for which the. DNB - ratio. is 1, with bulk pool water ' temperature as a
-parameter. It. is assumed that . all ' the vapor above the
-detachment point seperates f rom the heated, surf ace. From the figure-it_can be.seen that-with the design cooling. water temperature at the. core inlet (120*F) the maximum' heat flux
.- i s 325 kBTU/hr-ft 2 For a 85 element. core with~an overall peak-to-average' _ power. density _ ratio of 2.0, this heat _ flux corresponds'to a maximum reactor power.of 1900 kW.
- 4. 5 -
4.0 g
- 3.5 *
.]
.n m
/
?e.
- 3. 0
. t 2.S i 80 90 100 110! 120 130 ;; 'l40 Coolant inlet tem'perature ('F) 4 t
PLOT FOR WHICH DNB RATIO IS 1.0 0F MAXIMUM HEAT FLUX VERSUS COOLANT TEMPERATURE
- t. Figure 3-22 3-46
SAR 9/84-3.1.3.9. Nomenclature 2-A cross-sectional area, ft A: g channelifree flow area, ft 2 C coolant specific heat, . Btu /lb *F d diameter, in. D,. channel ~ equivalent diameter, ft D H hydraulic diameter, ft f b fricti n factor with subcooled boiling, dimensionless f, friction factor without boiling, dimensionless F forces acting on vapor bubble g constant. 4.'18 x 10 8 ft/hr 2 h b heat trangfer coefficient with subcooled boiling, Beu/hr-ft -F' H distance'from midplane of heated channel to free surface of pool, ft K pressure loss factor at channel inlet or exit, dimensionless n number.of equal axial increments into which hented length of core is subdivided N Number of channels which receive their flow from a single opening in t'ae lower grid plate p absolute pressure, 1b/ft 2 P heated perimeter of channel, ft
-P r .Prandt1 number 6p- pressure loss, Ib/ft 2 q heat load, Btu /hr q total heat load to channel, Btu /hr t
q" heat flux, Btu /hr-ft 2 q" p peak or " burnout" heat flux, Btu /hr-ft r B bubble radius 3-47 u -:
7-
,. SAR 9/84 s
1 m ., R'.e
.Reynolds number, dimensionless *
'S/V channel surface to volume ratio, iri+1 1 T '. coolant temperature. *F- s T,,g;' coolant. saturation temperature, *F 4-v- specific volume, ft 3 /lb - >
V . flow velocity',~ft/hr' [' W mass 1 flow rate .lb/hr ' s-Y-. non-dimensional radius z axial ~ coordinate in channel, ft ' t
-z t tal length of channel, ft- \
- t. ,
. ~
6z length of'a calculation increment in'the chhnnel,,ft-
- u. ' dynamic viscosity. . ft-lb/hr ,
.q a void fraction or fraction of a channel'cros's section which is occupied by vaporf dimensionless ,
3 a surface tension, lb/ft ' ( vapor volume,-or volume of vapor-area of heated-surface, cubic t[. ,/ produced square in. per unit
's f
~
v kinematic viscosity, ft /hr - T shear stress, lb/ft 2
-p density, lb/ft 3
. c' / D, ' relative roughness , s iU s' '
t i ! up 1
; i 4
h' i s i I I 3-48 L f 'k $ _, . - , - , r-m-, * , - , . 4,- ,.-~m 4,,..._,.-,.-...4..-,-,,,,--,.,-v..-.r <w
~. SAR 9/84 Subscripts-e . conditions at channel: exit _
' i-conditions'at channel entrance or-inlet
" l- conditions in portion of channel. adjacent to lower reflector.
m -conditions averaged over the liquid and vapor phases
- o. bulk pool conditions u- conditions in portions of channel adj acent to upper reflector j' axial increment index k axial station index
- w. conditions at cladding outer surface
- v. vapor L. liquid 3-49
m SAR 9/84
\
23.'2.INUCLEAR DESIGN'AND EVALUATION The characteristics and operating parameters of t h'i s
' reactor- have been calculated' -and extrapolated- using
~
-experience-and' data obtained from. existing TRIGA reactors as bench marks in evaluating ..the calculated data. There are s everal- . TRIGA systems D with . essentially ' the same core and:
~ reflector relationship as this TRIGA so the values presented.
here - are ..f elt ' to ' be ' accurate to within 5% but, 'of course,
-are influenced by specific core configuration: details : as well as, operational details.
Table 3-3' summarizes the. typical Mark II-TRIGA reactor parameters for a core containing. high-hydride, stainless steel clad fuel elements. Table ~ 3-3' TYPICAL TRIGA CORE NUCLEAR PARAMETERS
-Fuel.. elements- SS-clad U-ZrH l.6 7
Col'd clean critical loading '64 elements o
~2.5 kg U-235
- 10perational loading ~90. elements 0
~
4 ~3.4 kg U-235-i ' . 1. Prompt neutron lifetime 41-usec 8 Effective. delayed neutron frction 0.0070
~4 a Prompt negative' temperature ~1.0 x 10 6 k h'. C
' coefficient Tg, Average' fuel temperature (-1MW) 260*C
~
T,-Average wa'ter temperature:-(1MW)' 64.*C 1 3.2.1. Reactivity Effects The reactivity associated . with the control rod is of
-interest both'in the shutdown margin and in calculations of possible abnormal' conditions' related to reactivity accidents. Table ~3-4 gives approximate reactivity values associated with a' total control rod travel o f - .15 in. (38.1 cm), the full travel in the core.
3-50
M _' _ 5 SAR 9/84 M Table 3-4
~ ESTIMATED CONTROL ROD NET WORTH diameter 6k/k in' . (cm) %
CLring - transient 1.25 (3.18). 2.1
, C ring'- shim 1.35 (3.43) 2.6
,D. ring' -safety 1.35 .(3.43) 2.0 D. ring ; regulating .1.35 (3.43)- 2.0 The. . maximum .-reactivity- -inse'rtion rate is that associated with the transientErod which can be fully removed
._, from the core -in- 0.1 see producing an average reactivity insertion rate of 21% 6k/k-sec.
The total' reactivity - worth of the control system is about 8.7%. With- a- core ~ excess reactivity of 4.9%, -the shutdown margin.with all rods down is about 3. 8% and 'with the' most. reactive rod-stuck out is about 1%. The reactivity worth of the. fuel elements is dependent on'their position within-t_he core. Tab'le 3-5 indicates ~the values-that are expected in this installation. >r Table 3-5 ESTIMATED FUEL ELEMENT REACTIVITY WORTH COMPARED WITH
-WATER AS-A FUNCTION OF' POSITION IN: CORE Worth (% 6k/k) Number of i' Core Position SS Clad U-ZrH Fuel Positions l.6 B ring 1.07 6
. C ring 0.85 12 D ring- 0.54 18 E ring 0.36 24 F ring 0.25 30 G ring 0.19 36 4 Y Because of the prompt negative temperature coefficient j.
.a significant amount of reactivity is needed to overcome temperature. and allow lthe reactor to operate at the higher power-levels in steady-state operation. Figure 3-23 shows the '-relationship of reactor power level and associated reactivity loss to achieve a given power level. Figure 3-24 relates fuel temperature to'-a given steady-state reactor power: level.
3-51
jr n, t - < SAR 9/84-4 a e 1
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-PowtR (Kw) 4 7
ESTIMATED REACTIVITY LOSS VERSUS POWER 4 Figure 3-23 3-52
SAR 9/84 400-F
~
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~$- Y l200-
=
100-
- 0 , , , , .
0 200 400 600 800 1000 POWER (KW) ESTIMATED MAXIMUM B RING AND AVERAGE CORE TEMPERATURE VERSUS POWER Figure 3-24 8 3-53
7 _
.. SAR 9/84 The reactivity effects associated with the insertion-and- removal- of experiments in or around the core _ are difficult to predict; however, Table 3-6 is supplied to provide ' a guide to the-; magnitude . of the reactivity effects associated with. the introduction .- o f experiments in the reactor core.
Table 3-6
-EXPECTED REACTIVITY. EFFECTS ASSOCIATED WITH EXPERIMENTAL FACILITIES Worth (% 6k/k)
Central thimble, fuel vs H O +0.90 2 Central thimble, void vs HO -0.15 2 Pneumatic transfer tube, 1 (G ring) void vs HO -0.10. 2 Rotary specimen rack, void vs H ,, 0 -0.20 3.2.2. Evaluation of Nuclear Design The TRIGA reactor system is well-known for its conservative design. The stability of this reactor type has been proven both through calculations as well as through tests performed with the many TRIGA reactors in operation throughout the world. The stability of the TRIGA type reactor stems from the prompt negative temperature coefficient associated with the U-ZrH fuel in conjunction with a suitable neutron thermalizing
- material. This TRIGA will have the stability-that has been demonstrated on other TRIGA systems over the years.
A review of the reactivity worths associated with the reactor core indicates that no single item listed can produce a step reactivity insertion greater than that offered by routine' pulse operation. In the pulsed mode of operation the results of a step insertion of $3.00 are far below those attributed to test pulses on the advanced TRIGA prototype reactor in which 3.5% 6k/k was inserted in a step as is shown in Table 3-7. Therefore before experimental facilities are used, the worth of the experiment should be carefully evaluated. ~ If the experiment is worth more than . $ 3. 00, a special safety analysis should be prepared. 3-54
7 i SAR 9/84 L Table 3-7 f COMPARISON OF REACTIVITY INSERTION EFFECTS f :i Pulse Resulting from
- Max Pulse Tested Insertion of Maximum on SS-Clad, High Excess Reactivity in Hydride. Fueled This TRIGA TRIGA types
' Reactivity insertion, S 3.00 5.00 Steady-State power before pulse, kW <1 <1 Peak power, MW '1400 '8400 Total energy release, MW-sec '18 ~54 Period, msec ~3.1 '1.4 Max fuel temperature.
*C ~540 '1050 Pulse width, msec ~11 ~5.5 The possibility of a reactivity accident which could produce reactor powers and fuel temperatures attributed to a
$4.00 step insertion has been considered and evaluated in the accident analysis section of this report. It is concluded'from this analysis that the peak and average fuel temperatures resulting from this accident are well below the temperatures indicated as safety limits described in the reactor design bases of this document. It is further concluded that the integrity of the fuel containment will
~ not be jeopardized and no adverse effects to the reactor system or personnel will arise from the advent of such an accident.
l 3.3 THERMAL AND HYDRAULIC DESIGN This TRIGA reactor will be operated with natural convective cooling by reactor pool water. This method of heat dissipation is more than adequate for the power icvel of the reactor; i.e. 1000 kW(t). That is,the thermal and hydraulic design of the reactor is well within the safety limits required to assure fuel integrity. 3-55 l
1 4 SAR 9/84;
- n
% y e gi-
- 3. 3. l . . ' Design Bases M The- thermal and -hydraulic design for ,this .TRIGA is
' ' ~
based; on assuring that fuel. integrity is maintained during
" ' steady-state and pulsed mode ' operation as. well as during
!! - those abnormal conditions which . might be postulated for reactor . operation. . During. steady-state operation fuel
. integrity is maintained by limiting reactor powers to values which assure-that-the fuel cladding can transfer heat from the! fuel to.the reactor ' coolant- without reaching fuel-clad temperatures that "could ' result in clad rupture. If these
- temperature conditions were exceeded, the maximum local heat iflux . in the core would be greater than the heat- flux at which there~is a. departure from the nucleate boiling regime-L and consequently film blanketing of the fuel. This heat flux safety limit is a function of the inlet . coolant- -
temperature.. Figure 3-22 summarizes the results--of the i'
' thermal and hydraulic analysis for steady-state operation of the TRIGA.. In the figure critical heat flux for departure
- from nucleate' boiling is plotted as a function of the
- coolant inlet temperature. The maximum power. density in-a TRIGA. core.is found by multiplying the average power density.
- . by a radial peak-to-average power generation ratio of 1.6
! and an axial value of 1.25. The correlation used to determine the heat flux at [ . which there . is. a departure f rom nucleate boiling 'is from-Bernath [33]. This correlation encompasses a wider range o f-experimental data than the usual correlations, e.g., the , correlation due to McAdams, and, generally gives a lower value for the DNB ratio than the other correlations. During pulsing operation the limiting thermal-hydraulic-condition is the fuel temperature and the - corresponding H 2 overpressure' beyond . which- clad rupture may occur. As indicated in Section 3.1, coolant temperature is not a limiting-condition in pulsing since heating- conditions are essentially adiabatic and significant transfer of heat L energy . to the coolant does not occur until after peak
. fuel-clad temperatures occur.
l , The safety . limit on. fuel temperature occuring in the pulse. mode of operation is 1150*C. This temperature will ' give an internal equilibrium hydrogen pressure (U-ZrH fuel, H/Zr; 1.6) less. than that which -would produce a stress equivalent to the . ultimate strength of the' clad at a
- temperature of-680*C.- This-clad temperature'is higher than
[ would'actually occur and therefore conservative even in'the i case of a pulse' producing a peak adiabatic fuel temperature-of 1150*C. i Table 3-8 lists the pertinent heat transfer and
- hydraulic parameters for the TRIGA operating at 1000 kW.
3-56
r SAR_9/84
- These- data were taken- from the results of calculations
~ described in Section 3.1.
Table 3-8 1000 kW(t) TRIGA HEAT TRANSFER AND HYDRAULIC PARAMETERS Number of fuel elements '00 Diameter 1.475 in. Length (heated) .15.0 in.
' Flow area 2 0.522 ft Wetted perimeter 34.75 ft Hydraulic diameter 0.0601 ft Heat transfer surface 43.44 ft 2 Inlet coolant. temperature 120*F (48.9'C)
Exit coolant temperature (average) , 174*F (78.'9'C) Coolant mass flow 63,700 lb/hr Average flow velocity 0.55 ft/sec Average fuel temperature 500*F (260*C) Maximum wall temperature 280*F-(138'C) Maximum fuel temperature 842*F (450*C) Average heat flux 78,600-Btu /hr-ft Maximum heat flux 157,100 Btu /hr-ft Minimum DNB ratio 2.0
-3.3.2. Thermal and Hydraulic Design Evaluation The validity and safety of the TRIGA thermal-hydraulic design is established in Section 3.1. In that'section it is shown that design-basis conditions evaluated for TRIGA reactors using stainicas steci clad U-Zrli (II/ Z r ; 1. 6) fuel elements provide a generous safety margin for this TRICA.
These general evaluations are supported by extensive experience in operation of TRIGA cores at equivalent fuel
' temperatures and reactor power levels. No adverse results are reported from operation of TRIGA cores at fuel temperatures and power levels greater than this design.
3-57
SAR 9/84
, 3.4. MECHANICAL DESIGN AND EVALUATION 3.4.1. General Description The TRIGA Mark II reactor core assembly is located near the bottom of a elongated cylindrical aluminum tank surrounded by a reinforced concrete structure. A typical installation is shown in. Figure 3-25. The standard reactor tank is a we3ded aluminum vessel with 1/4-in.-thick walls, a diameter of approximately 2 meters, and a depth of approximately 7.6 meters. The tank is all-welded for water tightness. The integrity of the weld joints is verified by radiographic testing, dye penetrant checking, and leak testing. The outside of the tank is coated for corrosion protection.
An aluminum angle used for mounting the ion chambers, fuel storage racks, underwater lights, and other equipment, is located around the top of the tank. Demineralized water in the tank is provided such that approximately 6.4 meters of shielding water above the core. The core is shielded radially by a minimum of 2 meters of ordinary concrete with a density of 2.4 g/cm 3 91 (or 2.43 meters of concrete with a density of 2.88 gm/cc), '45cm. (1.5 ft) of water, 25.9 cm. (10.2 in.) of graphite reflector and 5.1 cm. (2 in.) of lead (see Figure 3-26). 3.4.2. Reflector Assembly The reflector is a ring-shaped block of graphite that surrounds the core radially. The graphite is 10.2 inches (25.91 cm.) thick radially, with an inside diameter of 21-5/8 inches (54.93 cm.) and a height of 21-13/16 inches (54.40 cm.) The graphite is protected from water penetration by a leak-tight welded aluminum can. A "well" in the top of the graphite reflector is provided for the rotary specimen rack. This well is also aluminum-lined, the lining being an integral part of the aluminum reflector can. The rotary specimen rack is a self-contained unit and does not penetrate the sealed reflector at any point. The graphite, and outer surface of the aluminum can are pierced by an aluminum tube which forms the inner section of beam ports i that penetrate the reflector. The reflector penetrating beam tubes are connected by stainicss steci couplings to the corresponding beam tube section fabricated as part of the reactor tank stucture. The reflector assembly rests on an aluminum platform at the bottom of the tank, and provides the support for the two grid plates and the safety plate. Four lugs are provided for lifting the assembly. 3-58 i
SAR 9/84 CONTROL 800 bitte
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SAR 9/84 7:y UPPER GRID PLATE
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SAR 9/84 3.4.3. Grid' Plates The top grid' plate. is an aluminum plate S/8 inches (1.59 cm.) _ thick (3/8 inches. 0.95 cm., thick in the central
, region) that provides accurate lateral ' positioning for the core components. The plate is supported by a ring welded to the top . inside surface of the reflector container and is anodized to resist wear and corrosion.
One hundred twenty six (126) holes, 1.505 inches (3.823 cm.) are drilled into the top grid plate in six circular bands around a central hole to locate the fuel-moderator and graphite dummy elements, the control rods and guide tubes, and the pneumatic transfer tube. (See Figure 3-27.). An equivalent diameter center hole accomodates the central thimble. Small holes at various positions in the top grid plate permit insertion of foils into the core to obtain flux data. A hexagonal section can be removed from the center of the upper grid plate for the insertion of specimens up to 4.4 inches (11.18 cm.) in diameter into the region of highest flux; this requires prior relocation of the six fuel elements from the B ring to the outer portion of the core and removal of the central thimble. This removable section will not be used initially; a seperate license amendment will be obtained prior to its use. Two generally triangular-shaped sections are cut out of the upper grid plate. Each encompasses two F and one G ring holes. When fuel elements are placed in these locations, their lateral support is provided by a special fixture. When the fuel elements and support are removed, there is room for inserting specimens up to 2.4 inches (6.1 cm.) in diameter. The differential area between the fitting at the top of the fuel elements and the round holes in the top grid plate permits passage of cooling water through the plate. The bottom grid plate is an aluminum pinte 3/4 inch (1.91 cm.) thick which supports the entire weight of the core and provides accurate spacing between the fuel-moderator elements. Six pads welded to a ring which is, in turn, welded to the reflector container support the bottom grid plate. Holes in the bottom grid plate are aligned with fuel element holes in the top grid plate. They are countersunk to receive the adaptor end of the fuel-moderator elements and the adaptor end of the pneumatic transfer tube. 3-61
SAR 9/84 TH H E _. HE ONAL SECTION lRRA01ATl0N SPACE O ' " UR LOCATION IRRADIATION
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TRANSIENT ROD o -- CORE ARRANGEMENT Figure 3-27 3-62
C SAR 9/84 3.4.4. Safety Plate The safety plate is provided to preclude the possibility of control rods falling out of the core. It is a 1/2 inch (1.27 cm.) thick plate of aluminum welded to the extension of the inner reflector liner and placed about 16 inches below the bottom grid plate. A central hole of 1.505 inches (3.823 cm.) in diameter in the lower grid serves as a clearance hole for the central thimble. Eight additional 1.505-inch (3.823 cm.) diameter holes are aligned with upper grid plate holes to provide passage of fuel-follower control rods. Those holes in the bottom grid plate not occupied by control rod followers are plugged with removable fuel element adaptors that rest on the safety plate. These fuel element adaptors are solid aluminum cylinders 1.5 inches (3.81 cm.) in diameter by 17 inches (43.18 cm.) long. At the lower end is a fitting that is accomodated by a hole in the safety plate. The upper end of the cylinder is flush with the upper surface of the bottom grid plate when the adaptor is in place. This end of the adaptor has a hole similar to that in the bottom grid plate for accepting the fuel element lower end fitting. With the adaptor in place, a position formerly occupied by a control rod with a fuel follower will now accept a standard fuel element. The adaptor can be removed with a special handling tool. 3.4.5. Fuel-Moderator Elements The active part of each fuel-moderator element, shown in Figure 3-28, is approximately 1.43 in. (3.63 cm.) in diameter and 15 in. long (38.1 cm.). The fuel is a solid, homogeneous mixture of uranium-zirconium hydride alloy containing about 8.5% by weight of uranium enriched to 20% U-235. The hydrogen-to-zirconium atom ratio is about 1.6. To facilitate hydriding, a small hole is drilled through the center of the active fuel section and a zirconium rod is inserted in this hole after hydriding is complete. Each element is clad with a 0.020 in. thick (.508 mm.) stainless steel can, and all closures are made by heliare velding. Two sections of graphite are inserted in the can, one above and one below the fuel, to serve as top and bottom reflectors for the core. Stainless steel end fixtures are attached to both ends of the can, making the overall length of the fuci-moderator element 28.8 in. (73.2 cm.). The lower end fixture supports the fuel-moderator element on the bottom grid plate. The upper end fixture consists of a knob for attachment of the fuel-handling tool and a triangular spacer, which permits cooling water to flow through the upper grid plate. The total weight of a fully-loaded fuel element is about 3.18 kg. (7 lb). i i 3-63 l,
SAR 9/86 STAINLESS STEEL TOP END FITTING m s
< l GRAPHITE 3.45 IN.
O STAINLESS STEEL TUBE CLADDING THlCKNESS 0.02 IN. URANIUM ZlRCONIUM HYDRIDE-WITH AXIAL - ZlRCONIUM ROD 28.8 IN. 15 IN. TOTAL' l.43 IN. -
~
~ l.47 IN.
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STAINLESS STEEL v BOTTOM END FITTING TRIGA STAINLESS STEEL CLAD FUEL ELEMENT WITH END FITTINGS Figure 3-28 3-64
n b SAR 9/84 An instrumented fuel-moderator element will have three thermocouples embedded in the fuel. As shown in Figure 3-29, the sensing tips of the fuel element thermocouples are located about 0.3 in. (0.76 cm.) from the vertical centerline. The thermocouple leadout wires pass through a seal in the upper end fixture. A leadout tube provides a watertight conduit carrying the leadout wires above the water surface in the reactor pool. Thermocouple specifications are listed in Table 3-9. In other respects the instrumented fuel-moderator element is identical to the standard element. Most of the fuel for the initial core loading will consist of elements with burnups of a fraction of a MW-day to several MW-days. It is anticipated that an initial core loading of about 94 fuel elements, including instrumented elements, and fuel followed control rods, will produce a cold, clean excess reactivity of ~4.9% Sk/k. Table 3-10 gives a summary of the fuel element specifications. 3.4.5.1. Evaluation of Fuel Element Design. General Atomic has acquired extensive experience in the fabrication and operation of high hydride, stainless steel clad fuel elements. As shown in other sections of this report, the stresses associated with the fuel and cladding temperatures in all modes of operation, normal and abnormal, are within the safety limits described in the Reactor Design Bases. It is concluded that the chemical stability of U-Zril l 6 fuel-moderator material does not impose a safety limit an reactor operation (see Section 3.1.1). Dimensional stability of the overall fuel element has been excellent in the TRIGA reactors in operation. Analysis of the heat removal from elements that touch owing to transverse bending shows that the contact will not result in hot spots that damage the fuel. Tests have been conducted on TRIGA fuel elements to determine the strength of the fuel element clad when subjected to internal pressure. At room temperature the clads ruptured at about 2050 psig. This corresponds to a hoop stress at rupture of about 7 2,000 . ps i which compares favorably with the minimum expected value for 304 stainless steel. 3.4.6. Neutron Source and Holder A 2 or 3 curie americiam beryllium neutron source will be used for startup. The neutron source holder is made of aluminum, is cylindrical in shape, and has a cavity to hold the source. The source holder can be installed in any vacant fuel or graphite element location. A shoulder at the 3-65
i SAR 9/84 1 l i
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INSTRUMENTED FUEL ELEMENT Figure 3-29 3-66
.g
' SAR-9/84 Table 3-9 THERMOCOUPLE-SPECIFICATIONS ;. Tr 3
Type- Chromel-alumel , Wire! diameter 0.005 in.
< ' Resistance 24.08-ohms / double foot.at 68'F JJunction~ Grounded Sheath material- . Stainless steel' Sheath. diameter -0.040 in.
. Insulation Mg0 Lead-out wire
. Material Chromel-alumel Size- -20 AWG Color code Chromel -. yellow (positive).
Alumel - red (negative)- Resistance 0.59 ohms / double foot at 75'F Table 3-10 '
SUMMARY
OF FUEL-ELEMENT SPECIFICATIONS Nominal Value Fuel-Moderator Material H/Zr. ratio 1.6 Uranium content 8.5 wt % Enrichment (U-235) 19.7 10.2
- Diameter' 1.43 in.
, Length 15 in.
Graphite End Reflectors Upper Lower l Porosity 20% 20% Diameter 1.43 in. 1.43 in. Length 3.44 in. 3.47 in. 4 Cladding Material Type 304 SS 0.020 in, r Wall thickness Length 22.10 in. I' End Fixtures and Spacer Type 304 SS Overall Element i Outside diameter 1.47 in. Length 28.37 in. Weight 7 lb i 3-67
SAR 9/84 i upper end of the holder _ supports the assembly on'the uppet. grid -plate, the rod itself, which contains the source, extending down into the-core region. The neutron source is contained in a . cavity in the lower portion of the rod assembly at the horizontal centerline of the core. This cylindrical cavity is 0.981 in. - (2. 4 9 2 cm. ) in diameter and approximately 3 in. (7.62 cm.) deep. The upper and lower portions are screwed together. A soft aluminum ring provides sealing against water leakage into the cavity. Since the upper enc 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. 3.4.7. Graphite Dummy Elements Graphite dummy elements may be used to fill grid positions not filled by the fuel-moderator elements or other core compounds. They are of the same general dimensions and construction as the fuel-moderator elements, but are filled entirely with graphite and are clad with aluminum. 3.4.8. Control System Design The reactor uses four control rods:
- a. A safety rod
- b. A shim rod
- c. A transient rod
- d. A regulatory rod The regulating, shim and safety rods are sealed 304 stainless steel tubes approximately 109 cm (43 in.) long by 3.43 cm (1.35 in.) in diameter in which the uppermost 16.5 cm (6.5 in.) section is an air void and the next 38.1 cm (15 in.) is the neutron absorber (boron carbide in r,olid form).
Immediately below the neutron absorber is a fuel follower section consisting of 38.1 cm (15 in.) of U-ZrH fuel. The bottom section of the rod is 16.5 cm (6.5 in . ) I a'i 6r v o id . The regulating, safety, and shim rods pass through and are guided by 3.81 cm. (1.5 in.) diameter holes in the top and bottom grid plates. A typical control rod with fuel follower is shown in the withdrawn and inserted postions in Figure 3-30. The safety-transient rod is a scaled, 93.35 cm. (36.75 in.) long by 3.18 cm. (1.25 in.) diameter tube containing l solid boron carbide as a neutron absorber. Below the absorber is an air filled follower section. The absorber section is 38.1 cm. (15 in.) long and the follwer is 53.02 j- 3-68
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PLATE le.t S IN. _ SAFETY PLATE o i' _ ROD FULLY RODFULLY WITHORAWN INSERTED FUEL FOLLOWED CONTROL ROD Figure 3-30 3-69
SAR 9/84 cm. (20.88 in.) long. The transient rod passes through the core in a perforated aluminum guide tube. The tube recieves its support from the safety plate and its lateral positioning from both grid plates. It extends approximately 25.4 cm. (10 in.) 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. The control rods are connected to their individual drive units by screwing the upper end of the rod into a control rod drive assembly connecting rod. Vertical travel of each rod is approximately 38.1 cm. (15 in.). Reactivity worths and core positions for each rod are summarized in the section on nuclear design. A summary of other control rod design parameters is given in Table 3-11. Table 3-11
SUMMARY
OF CONTROL ROD DESIGN PARAMETERS Safety Shim and Transient Regulating Cladding Haterial Al Type 304 SS OD 3.18 cm (1.25 in.) 3.43 cm (1.35 in.) Length 93.35 cm (36.75 in.) 109.5 cm (43.13 in.) Wall thickness 0.071 cm (0.028 in.) 0.051 cm (1.35 in.) Absorber Material Boron Carbide (solid form) OD 3.02 cm (1.19 in.) 3.32 cm (1.31 in.) Length 38.1 cm (15 in.) 36.20 cm (14.25 in.) Follower Material Air U-Zrli OD 3.18 cm (1.25 in.) 3.34d46(1.31 in.) Length 53.02 cm (20.88 in.) 38.1 cm (15 in.) 3.4.8.1. Control Rod Drive Assemblies. The control rod drive assemblies for the safety, shim and regulating rods are mounted on a bridge assembly over the pool and consint of a motor and reduction gear driving a rack-and-pinion 3-70
SAR 9/84 as indicated in Figure 3-31. A helipot connected to the pinion generates the position indication. Each control rod drive has a tube that extends to a dashpot below the surface of the water. The control rod assembly is connected to the rack-through an electromagnet and armature. In the event of a power failure or scram signals, the control rod magnets are de-energized and the rods fall into the core. The time required for a rod to drop into the core from the full-out position is about I second. The rod drive motor is non-synchronous, single-phase, and instantly reversible, and will insert or. withdraw the control rod at a rate of approximately 11.5 in./ min. (0.5.cm/sec) for the safety and shim. A regulating rod drive withdraws the control rod at a rate of 24.0 in./ min (1.02 cm/sec.). A key-locked switch on the -control console power supply prevents unauthorized operation of all control rod drives. Electrical dynamic and static braking on the motor are used for fast stops. Liuit switches mounted on the drive assembly actuate
- circuits which indicate the following
- a. The "up" and "down" positions of the magnet
- b. The "down" position of the rod
- c. The magnet in contact with the rod 3.4.8.2. Transient Rod Drive Assembly. The safety transient control rod on pulsing TRIGA - Mark 11 reactors is operated with a pneumatic rod drive (see Figures 3-32 and 3-33). Operation of the transient rod drive is controlled from the reactor console.
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 be withdrawn from the core; administrative control is exercised to restrict its travel so as not to exceed the maximum licensed step insertion of reactivity ($3.00 or 2.1% ok/k). 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 I 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 and of the cylinder. At the end of its travel, the piston i strikes the anvil of an oil-filled hydraulic shock absorber, j which has a spring return, and which decelerates the piston
- 3-71
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RACK AND PINION CONTROL ROD DRIVE Figure 3-31 3-72
i I l SAR 9/84
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,j DAMPER I CONNECflNG #00 ADJUSTABLE TRANSIENT ROD Figure 3-32 3-73
r SAR 9/84
=-
SHOCK ABSORBER VENT H0LES --
/ SHOCK ABSORBER 7(1 ANVIL EXTERNALLY THREADED CYLINDER P1STON ANV1L SUPPORT
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CONNECTION TO 80TTOM LIMIT = PISTON R00 CONTROL R00 TRANSIENT ROD OPERATIONAL SCllEMATIC Figure 3-33 3-74
+ SAR 9/84 1 1 p 1st'a controlled-rate over its last 2 in . '(5 cm. ) of travel.. LWhen the solenoid
- is de-energized, the 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:its fully inserted. position-in.the reactor core.
The extent-of transient rod withdrawal f rom the core during . = a pulse is determined. by ' raising- or . lowering the cylinder, .thereby. controlling - the distance the piston travels. 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 on the direction of worm gear rotation. A mechanical indicator is driven by the worm shaft. The-distance the transient rod will be ejected from the reactor-core is determined by the position of.the. cylinder and the mechanical indicator.. The pneumatic cylinder position is controlled by a crank inserted at the rod drive. The crank is. inserted into the drive only when position changes are:to-be made.
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 connecting rod assembly that' moves .the transion.t 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. 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. Lir required to hold the rod in a withdrawn position and the rod drops into the core from the full out position in about.1 second. 3.4.8.3. Evaluation of 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 scram times for the rods are quite adequate since ' the TRIGA system does not rely on speed of control as being 3-75 (
J
, 1
.SAR 9/84
- 5 o ' paramount. to the safety. of the reactor. The inherent shutdown mechanism of. 'the TRIGA prevents unsafe excursions andi the control system is used only for the plann'ed shutdown-
^ , of. the . reactor .and -to control the power level- in
-steady-state.~ operation. ,
f N 3.4.9. Experimental and" Irradiation Facilities. ; f The. experimental and ' irradiation , facilities of ,the TRIGA Mark II reactor are extensive and versatile. Physical
' access and observation'of the3 core are possible at all. times through the vertical water shield.s Experimental tubes can Leasily.be installed.in the core region to provide facilities f or high-level irradiations or in-core experiments. Areas outside the core and reflector are available for . larger experiment ~ equipment or facilities.
3.4.9.1 Central Thimble. The reactor is equipped with a central thimble for access to the point of maximum flux in the. core.. The central thimble consists of an aluminum tube _.that.-fits through the center hole of the top and bottom grid plates. Dimensions of the tube are 1.5 in, o.d. (3.81 cm.)
-and'1.33 in.-1.d. (3.38 cm.).. Holes in the tube ~ assure that it; is normally filled with water.. Water is expelled from the . tube by compressed . air. Experiments with the central thimble include irradiations of - small samples and the
-exposure of aterials to a co111 mated beam of neutrons'or '
gamma-rays. L -3.4.9.2 Rotary Specimen Rack. A . rotary,- multiple-position (40) specimen rack located in a well in
.the: t o p o f. the graphite ref. lector . providos for the Large
. .s_cale production of radioisotopes and for the activation.and irradiation of mutilpe-samples. A11 positions in'this rack are exposed- to neutron. fluxes of comparable' intensity'. Specimen : positions are 1.25 in. (3.18 cm.) in -diameter ' by
.10.80.in. (27.4 cm.) in ' dep th. - . Samples are loaded into from! the top . 'of Et h'e reacto'r " through ' a w'ater-tight ttib e the
= rotary rack using a specimen lifting ^ Ndevice or pneumatic
. pressure- for insertion and removal o f- samples from the 9 '- sample rack positions. The rotary specimen rack can be turned from the top of the, reactor by-manual operation or,by a' motor drive. ,
1 ~ ' '
. 3.4.9.3 Pneumatic Speciben Tu ba .) A pneumatic transfer system permits applications with sbort-lived radioisotopes.
The in-core terminus of ,this system is normally located in ', the outer ring of fuel element, positions, a region of high neutron flux. The' sample cohtainer (rabbit) is conveyed;to a receiver-sender station- via 1.25 in. o.d. (3.18 cm.-) aluminum tubing. Effective space in3 the specimen transfer capsules is 0. 6 8. in . '(1.7 cm.{ diamecer3 by ' 4. 5 in. (11.4 ca. ) -height . An optional transfer box a s y , b'e employed to
- - g i '1
. 3-76 ,
~
~ '
l . , . - - . - _
.d
~
SAR 9/84-permit the sample to be-sent and.. received from up to three different receiver-sender-stations. 3.4.9.4 Beam' Tube Facilities. The beam ports provide
. tubular < penetrations through the concrete-shield and reactor
. tank water , .' making ' beams; of. neutrons - (or gamma radiation)
+
'available for experiments. The. beam ports also provide an irradiation facility for large sample specimens in a -. re gion close to the core. Beam port' diameters near the core are 6 in. .(15.2 cm.)..
There' are five beam ports divided into two . categories as follows:
.a.. Tangential beam ports. Two' beam ports are oriented tangential to the reactor core, penetrate the graphite reflector, the coolant . water, and the concrete shield. A hole is drilled'in the graphite tangential to '.th e outer edge of the core. .One beam port . terminates at the tangential
. point to the core. ~The other beam tubes exter.d both '
directions.from the reflector and out opposite sides of the reactor ~ shield.
- b. Radial beam ports. There are two radial beam ports, each which penetrate the concrete shield structure'and the coolant water. One radial port terminates at the outer edge.
-of the reflector. The second radial port also terminates at-the outer edge of the reflector.. However,'a' hole drilled in the graphite reflector -extends the ef f ective source of the
. radiations to the reactor core region.
~
A step.is incorporated into-each beam. port to. prevent 1 radiation. streaming through the gap between the beam tube and' shielding plug.- The inner section of each beam port-is .
- an ' aluminum pipe 6_ in. .(15._2' cm) in diame ter; the- outer section ~ is - a steel pipe 8 in. (20.3 cm) in diameter.- A
" shadow" - shiel'd . is - incorporated 'in the concrete strueture around ' _ the ~ step in each beam port.
~
This shield , .which is
= constructed of ,4-in. (10.2-cm)-thick steel plate, reduces
~
the . amount of. radiation passing through the concrete when the . beams port ' is in use. Special shielding reduces the ,
- radiation outside the concrete to a safe' level when.the beam ,
port is not in use. The shielding is provided- in four - sections as follows:
- a. An inner. shield plug.
- b. . An outer shield plug.
~
- c. A lead-filled shutter.
- d. A lead-filled door.
3-77 ' - - ,,P _ - ,-.- _ .--- _ m. ,.- _ ,,_ _ ,_- _ - - - ,-,.._ ,
h-'
~
4 SAR 9/H4
- The inner J shield plug consists of 40.in. (1.02 m) of graphite backed .with a O'.125-in. (0.32-cm) sheet of.boral
-and - . 5 :in. ( 12. 7 ~ cm) _. o f lead sandwiched between two 1.25-in.
l(3. 2-cm)-thick s teel- _ plat e s . . The graphite. .is 6 in. .(15.2
~
- cm). .in . diam'e te r . Three rollers are provided to facilitate c
the : insertion end -removal of the inner shield plugs. To
. help. guide the; innermost plug over the step in the, beam tube
- during insertion, the inner end of the-' plug is cone-shaped.
- A.threadedsholelis provided in the outer end of the plug for attachingJthe beam tube plug-handling tool.
The1.o' uter. shield plug is polyethylene and is - 8 _ in . E (20.3 cm)'in-diameter and 43 in. (1.09 m) long.- A handle _on the outer end of.this plug is.provided for manual handling'.. The~ plug weighsLabout 45 lb (20~kg). The plug is equipped-with . an . electrical circuit consisting of a position switch mounted in the front of the plug and an electrical connector
.at.the. rear of the plug. The switch can be actuated only by the inner. plug' when-the inner plug is installed in the beam tube.
The connector is inserted in an outlet box mounted in the beam - to complete the circuit to the control console. Lights-on the console for each beam tube indicate when the _ . plugs are inserted. The- lead-filled shutter and lead-lined door provide limited ' gamma shielding when the plugs are removed. The~
.g
- shutter is contained _in a rectangular steel housing recessed
;in the' outer' surface of the concrete shield. The shutter is-
~10 in. (25.4 cm) in diameter and 9.5 in. (24.1 cm) thick- .
It is': operated by a removable push rod on the face of the - shield' structure. and can - be moved even with the . shutter
. housing' door-closed. In the_open position, a section of.the shutte- consisting of an 8-in. (20.3-cm) pipe is aligned p
with ; the beam port and the outer shield center plug to-facilitate: insertion or removal of the beam plugs. _The shutter ~ h'o u's i n g is equipped with a steel cover plate'. lined
- . with '1. 2 5 , in . . (3.2 cm) of lead for additional shielding.
-Thel cover plate is- bolted in place and equipped with' a-rubber gasket. A~ removable-lead-lined center plug-provides easy- access co. the beam port. The plug'can be bolted shut
" Jso that the seal would prevent -loss of shielding water if
.the beam tube _should develop a serious leak.
W , 3.5. SAFETY SETTINGS IN' RELATION TO SAFETY LIMITS As has been indicated, fuel temperatures are the main safety considerations in the_ operation of the TRIGA system. .- :The . temperature ' of' the fuel may be controlled by The setting operating limits. on. other operating parameters. parameters of interest for TRIGA are: 3-78 y iT*T-g-dj --wr g p.+ +3y 9hIEy W - 9 -9. y+4*-h +m- m , gg v g-W gengpwqty s%ge 9 w--pg 5 4 p gma 9pp--gpeg- y& gy, ggepg 9tisyr-g @ w-mgp.i-ge,*rgMe g ,agg g -t ym-g,q g ,9 p gpq e & WM - m gr M
SAR 9/84
~
a.' Maximum. licensed steady-state power level
-b. Fuel temperature measured by thermocouple
- c. Maximum reactivity worth of transient rod
'd. Core inlet coolant water temperature.
'The safety settings as listed in . Table 3-12 are such that in all operation, normal and abnormal, the safety.
limits indicated in the reactor design bases will not be exceeded. Table 3-12 TRIGA SAFETY SETTINGS Parameter Limited Safety Setting Function Maximum power level 1100 kW(t) Reactor scram steady-state Maximum fuel temperature 500*C Reactor scram measured Administative limitations are imposed for the transient rod and coolant water temperature as follows: a '. Maximum worth of transient insertion of 2.28% 6k/k
- b. Core inlet water temperature of 48.9'C.
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 settings, it is reasonable that at some later date less restrictive safety system settings could be assigned in conjunction with the upgrading of the reactor to operate at higher steady-state power levels or in the pulsing mode, while still using the same fuel and core configuration. 3-79
SAR 9/84 Chapter 3 References
- 1. Merten. U., et al., T h e rm a l Migration of Hydrogen in Uranium-Zirconium Alloys", General Dynamico. (:e n e r a l Atomic Division Report GA-3618, 1962.
- 2. Coffer, -C. -O., et al., "Research in Improved TRIGA Reactor Performance, Final Report", General Dynamics, General ~ Atomic Division Report GA-5786, 1964.
- 3. Johnson, H. A., et al., " Temperature Variation, Heat Transfer, and Void Volume Development in the Transient Atmosphere Boiling of Water", SAN-1001,. University of California, Berkeley, 1961.
- 4. McAdams, W. H., Heat Transmission, 3rd ed, McGraw-Hill Book Co., New York, 1954.
- 5. Sparrow, E. M. and R. D. Cess, "The Effect of Subcooled Liquid on Film Boiling", Heat Transfer, 84, 149-156 (1962).
- 6. Speigler, P., et al., " Onset of Sts. ole Film Boiling and the Foam Limit", Int. J. Heat and Mass Transfer, 6_ ,
987-989 (1963).
- 7. Zuber, W., " Hydrodynamic Aspects of Boiling Heat Transfer", AEC Report AECV-4439, TIS, ORNL, 1959.
- 8. Rohsenow, W., and H. Choi, Heat, Mass and Momentum Transfer, Prentice-Hall,-1961, pp. 231-232.
- 9. Ellion, M. E., "A Study of the Mechanism of Boiling Heat- Transfer". Jet Propulsion Laboratory Memo. No.
20-88, 1954.
- 10. Coffer, C. O., et al., " Characteristics of Large Reactivity insertions in a High Performance TRIGA U-ZrH Core", General Dynamics, General Atomic Division Report GA-6216, 1965.
- 11. Fenech, H., and W. Rosenow, " Thermal Conductance of Metallic Surfaces in Contact", USAEC NYo-2130, 1959.
- 12. Graff, W. J.", Thermal Conductance Across Metal Joints", Machine Design, Sept. 15, 1960, pp. 166-172.
- 13. Fenech, H., and J. J. Henry, "An ar.a ly s i s of a Thermal Contact Resistance", Trans. Am. Nucl. Soc. 5, 476 (1962).
3 , - - . , . -. . - . . . - . .,
SAR 9/84 14 . . Bernath, L..;"A Theory'of Local Boiling Burnout and Its Application t o .Exis ting Da t a'.' , He a t Transfer - Chemical Engineering-Progress Symposium Series, Storrs, 18.- Lenihan, S. R., " GAZE-2: A'One-Dimensional, Multigroup, Neutron Diffusion Theory Code for the IBM-7090", General . Dynamics,' General -Atomic Division Report GA-3152, 1962. u
- 19. _Dorsey,;J. P., and-R. Froehlich, " GAMBLE-5 -
A Program for the Solution : o f- the Multigroup Neutron-Diffusion Equations in. Two Dimensions, with Arbitrary Group Scattering, for the.UNIVAC-1108 Computer", Gulf' General Atomic Report GA-8188, 1967. 20.- Lathrop, D. K., ."DTF-IV, A FORTRAN-IV Program for Solving the Multigroup . Transport Equation with Anisotropic Scatterings", USAEC Report LA-3373,. Los Alamos Scientific: Laboratory, New Mexico, 1965.
- 21. Adler, F. T., G. W. Hinman,.and L. W. Nordheim, "The Quantitative Evaluation of . Resonance Integrals", in Proc. 2nd Intern. -Conf. Peaceful -Uses At. Energy
. (A/ CONF. 15/P/1983), Geneva, International Atomic Energy Agency, 1958.
- 22. Brown, H..D., Jr., Gulf General Atomic Inc., "THERMIDOR A FORTRAN. II Code for Calculating the Nelkin Scattering Kernel for Bound Hydrogen (A Modification of Robespierre)", unpublished data.
- 23. Nelkin, M. S., " Scattering of Slow-Neutrons by Water",
Phys. Rev. 119, 741-746 (1960), g 24. .McReynolds, A. W., et al., " Neutron Thermalization by-Chemically-Bound Hydrogen- and Carbon", _in Proc. 2nd Intern. Conf. Peaceful .Uses At. Energy (A/ CONF. 215/P/1540), Geneva, International Atomic EnergyLAgency, 1958. 25.. Whittemore, W. L., " Neutron Interactions in Zirconium Hydride", USAEC Report- GA-4490 (Rev.-), General Dynamics, General Atomic Division, 1964. 2 6.- Bell,. J., '.' S U MMIT : An- IBM-7090 Program for the Computation of Crystalline Scattering Kernels". USAEC Report, General Dynamics, General Atomic Division Report GA-2492,.1962. 4
- 27. Beyster, J. R., .et al., " Neutron Thermalization in Zirconium Hydride", USAEC Report, General Dynamics, General Atomic Division Report GA-4581, 1963.
3-81 i
SAR 9/86
- 28. Woods', A..D. B '. , et al.,- " Energy Distribution of Neutrons Scattered from Graphite, Light and Heavy Water, Ice, Zirconium Hydride, Lithium Hydride, Sodium
" Hydride, and ' Ammonium Chloride, by the Beryllium
-Detector Method", in : Proc. Symp. . Inelastic Scattering of Neutrons in Solids and Liquids, Vienna, Austria, Oct. 11-14, International Atomic Energy Agency, 1960.
'2 9. Jordan, D. P. and G.-Leppert, " Pressure Drop and Vapor Volume with Subcooled Nucleate Boiling", Int. J. Heat
' Mass Trans. 5,-751-761 (1962).
- 30. McAdams, op. c i t'. , pp. 390-392.
- 31. - Levy, S., " Forced Convection Subcocled Boiling-Prediction of Vapor _ Volumetric-Fraction", Int.
J. Heat Mass Trans. 10, 961-965 (1967).
- 32. McAdams, op. cit., p. 397
- 33. - Bernath, op. cit., pp. 95-116 3-82
SAR 9/84 Chapter 4 INSTRUMENTATION AND CONTROL SYSTEM Design of the instrumentation and control system is
-intended for new TRIGA reactor facilities and replacement of old- reactor consoles. Completion and verification of the design will occur in a GA Technologies reactor facility before installation at The- University of Texas at Austin.
An active evaluation by the University is intended of the GA instrument ~ and control-console for the TRIGA as part of the original installation of the console by the vendor.
-4.1 DESIGN BASES A new Instrumentation and Control System (ICS) [1] for the TRIGA reactor is a computer-based design incorporating the.use.of a GA-developed, multi-function, NM-1000 digital neutron monitor channels. Two complete NM-1000 systems provide redundant safety channels (percent power. with scram), wide-range lor power (below source level to full power), period, and.muu 4.-range linear power (source level to full power). The control system logic is contained in a
~
separate control system computer (CSC) with color graphics display which 1s the interface between the operator and the reactor. While information from the NM-1000 channels is processed and displayed by the CSC, each NM-1000 safety channel is independent, has its own output displays and connects directly to the safety system scram circuit. The NM-1000 digital neutron monitor channels were developed for the nuclear power industry and are fully qualified for use in the demanding and restrictire conditions of a nuclear power generating plant. Their design is based on a special GA-designed, high sensitivity fission chamber, and. low noise ultra-fast pulse amplifier. The control system computer manages all control rod movements, accounting for such things as interlocks and choice of particular operating modes. The CSC also processes and displays information on control rod positions, power level, fuel and water temperature and can display pulse characteristics and forced flow system parameters. Many other functions can also be performed by the CSC, such as calibrating control rods, and monitoring reactor usage. A computer-based control system has many advantages over an analog system in terms of speed, accuracy and reliability, and the ability for self-calibration, improved diagnostics, graphic displays and logging of vital information. 4-1
SAR 9/84
- 4.1.1 NM-1000 Safety and Neutron Monitor Channel The NM-1000 nuclear . channels have multi-function capability _ to provide safety (scram) action as well as . neutron' monitoring over a wide power range from a single detector. The selectable functions are:
- a. Percent power with scram.
- b. Wide-range log power.
- c. Power rate of change,
- d. Multi-range linear power.
For the TRIGA ICS, one NM-1000 system is designated to provide the wide-range log power function and the percen't power safety channel with scram (linear power level from 1 to 125%). The wide-range log power function is a digital version of the patented GA 10-decade log power system to cover the reactor _ power range from below surface level to-150% power'and provide a period signal. For the log power function, the chamber signal from startu'p (pulse counting) range through the Campbelling (root mean square [RMS] signal processing) range covers in excess ~ of 10-decades of power level. The self-contained microprocessor combines these signals and derives the power rate of change (period) through -the full range of power. The microprocessor automatically tests the system to ensure that the upper decades are operable while the reactor is operating in the lower decades and vice versa when the reactor is at - high power. The second NM-1000 system provides the multi-range linear power range data as well as the percent power safety-channel with scram. For the multi-range function, the NM-1000 utilizes the same signal source as for the log function. However, instead of the microprocessor converting the signal into- a log function, it converts it into 10 linear power ranges. This feature provides for a more precise reading of linear power level over the entire range of reactor power. The same self-checking features are included for the log function. The multi-range ranging function is 'either auto-ranged or slaved to a position switch on the operator's console via the control system computer. Each NM-1000 system is contained in two NEMA (National Electrical Manufacturers Assoc.) enclosures, one for the
- amplifier and one for the processor assemblies. The amplifier assembly contains modular plug-in subassemblies for pulse preamplifier electronics, bandpass filter and RMS electronics, signal conditioning circuits, low voltage power supplies, detector high voltage power supply and digital 4-2 L
7 - SAR 9/84-1
-diagnostics _ and; communication electronics. The processor assembly is made up ' of. ' modular plug ~-in subassemblies for
. communic'ation electronics.(between amplifier,and processor _),
E the L microprocesso_r , la control / display module, low . voltage powerisupplies,. isolated 4 to-20'mA outputs, and _ isolated alarm .. ou tpu t s . . Ou tpu t s ,are : Clas s _.1 E -- as specified - by:! IEEE 323-1974. Communication between the. amplifier and processor- - assemblies- is ;via 2 twisted pair shielded . cables. . The amplifier / microprocessor'; circuit design . employs the . latest
. concepts- in- automatic. on-line self di' agnostics. and calibration verification. Detection'of' unacceptable circuit (performances is automatically alarmed.
The neutron detector- uses .the standard 0.2 counts /sec
. 'p e r - nv fission chamber that has prov'ided reliable serviceLin'. -
the past.- .It .has, however, been . i m p r o v e.d_ by' additional shielding-to provide a greater signal to noise ratio. 'The low noise construction of the chamber- assembly allows the _ system to:. respond'to a' low reactor shutdown-level which is subject to being -masked by' noise . An illus tration of 'the neutron channel operating ranges is shown in Figure 4-1. 4.I'.-2 CSC and Control Console A-conceptual layout of the control console is shown in Figure 4-2. The control- console contains the several components needed by the operator for_ -reactor control'.
-Included are:
- a. Reactor control' panel.
b.-Control system computer (CSC).
- c. Color-CRT monitor..
- d. Power and temperature meter panel.-
- e. Two disc drives, printer and terminal.
. Functions of the control panel are represented in Figure.4-3-and are presented as:
~
.a. Key switch for rod magnet power.
- b. Digital rod position indicators.
c -Rod control switches and' annunciators. d '. Reactor - operating mode -switches,
- e. External switch annunciators (such'as beam port open-close reactor bay access etc.).
As previously mentioned, the power and period information from the-NM-1000 channels is processed and 4-3
i SAR 9/84 j i 1 7; f. 2000 MW - -- 200 MW - PULSE _ MODE 20 MW - A B C 2 MW -.- -- -- -- --
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~ - - - - - - - - - - - - - - - - - 10-7%
SOURCE INTERLOCK' TRIP 0.0002 W 10'8% A = Wide Range Log Channel B = Wide Range Linear Channel C = Manual, Automatic, and Squarewave Modes EL-0350B t i' NEUTRON CHANNEL OPERATING RANGES Figure 4-1 4-4 ! 4
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gn-i l SAR 9/84 1 displayed by-the-CSC. . However, each NM-1000 safety system is independent, has its own output displays and connects directly to the control system scram circuit. Thus, both percent power ' channels , vide-range log power, period and multi-range linear, power have their output displayed on meters as well as on the color CRT. This is also true of fuel temperature and control rod position indication. The~ control-system computer (CSC) provides all of the logic functions needed to control the reactor and augments the saf ety system by monitoring for undesirable operating characteristics. It displays reactor operational information in a color format on a CRT monitor for ease of compreh'ension and 'o n demand will display secondary information concerning the operation of the reactor auxiliary systems. Essentially all of the control system logic contained in previous TRIGA reactor control systems is incorporated into the CSC. However, instead of utilizing electronic circuits and electrical relay circuits, the logic is programmed into the computer. The availability of the computer allows great versatility and flexibility in operationally related activities aside from the direct control of rod movements. Many other functions can also be performed by the CSC, such as calibrating control rods, monitoring reactor usage and locating individual fuel element positions and burnup, isotope irradiation data, etc. The control system logic regulates control rod movements are based upon the operating mode selected and'the reactor operating characteristics that are constantly monitored, such as power and fuel temperature. 4.'1.3 Reactor Operating Modes There are four operating modes: the basic modes - manual, automatic, square-wave and. pulse - that operate with natural convection cooling. Four additional modes are available with operation logic for forced flow cooling. The manual and automatic modes are steady-state reactor conditions; the square-wave and pulse modes are the conditions implied by their names. The natural convection mode provides cooling without operation of a primary cooling pump. This mode has a lower maximum power level than the forced-flow mode, which uses primary coolant pumps. The manual and automatic reactor control modes are used for reactor operation from source level to 100% power. l These two modes are used for manual reactor startup, change in power level, and steady-state operation. The square-wave operation allows the power level to be raised quickly to a desired power level. The pulse mode generates high-power levels for very short periods of time. High power and low power pulse mode options are available. 4-7 2
s b- s , 4,, L. SAR 9/84 T a rod' control.is accomplished by the lighted; push Manu'1 m'~
. buttons' on the_-r'od control panel. . The top row of annunciators, when illuminated, indicates magnet contact with the armature and-magnet current. Depressing any one of.
,, the, CONT /0N push. buttons will interrupt the cu rren ti . t o that magnet and extinguish the magnet' current ON. indicator. If the rod is above the down limit, the rod will fall back into the ; core : and the CONT light will remain extinguished'until the magnet is driven to -the down limit where it again-contacts the armature.
.W hen- illuminated, the annunciators in the middle row indicate the upper (UP) limit position .and the bottom row annunciators indicate: the lower (DOWN) limit position ~of the-rods. Depressing, the . indicators _ causes the control rod to move in the. direction indicated. Sever.al interlocks prevent:
the" movement of the' rods in the UP direction such-as:
-a. Scrams not reset.
b.sMagnet not coupled to armature.
- c. Source level below minimum count.
-d. Two-UP switches depressed at the same time.-
i
- e. Mode switch-in one of the pulse positions.
~
- f. Mode switch in AUTOMATIC position (regulation rod.
only). There is no interlock inhibiting the DOWN direction of-the control rods except in - the case of the regulating rod
- -' while in the AUTOMATIC mode.-
Automatic power control can be obtained by - switching from. manual operation to automatic operation. All the instrumentation, safety, and interlock circuitry _ described above applies and is in operation in this mode. However, the regulating ' rod is now controlled automatically .in response to a power level and period' signal. Reactor power
' level.' is compared with the demand level set by the operator and'is used to bring the reactor power to the demand. level on a fixed preset period. The purpose of this feature is to . Laaintain automatically the preset power level during long-term power . runs.
Options. are available to maintain power by.' . movement .o f a single rod or by bank operation of the rods. -Under bank operation, rods can be controlled by equal position (all rod positions equal) or by set position (rods move as a bank from pre-established positions).
'In square-wave operation, the. reactor is first brought to ' criticality below 1 kW, leaving the transient rod partially in the' core. All of the steady-state 4-8
SAR 9/84 instrumentation 'is iri operation. The transient rod is ejected from the core by means of the transient rod FIRE push button. When the power level reaches the demand level, < it is maintained much the - same as in the automatic mode except that two rods are used to maintain power after'the pulse rod is ejected. Reactor control in the pulsing mode consists of establishing criticality at a flux level below 1 kW in the steady-state mode. This is accomplished by the use of the motor-driven control rods, _ leaving the transient rod either fully or partially : inserted. One of the two pulse mode selector switches is then depressed (selected to give an on-scale reading for the peak power level of the pulse to be produced). The MODE SELECTOR switch automatically connects the - gamma pulsing chamber to monitor and record peak flux (nv) and e n e r g y . r e'l e a s e (nyt). Pulsing can be initiated from either the critical or sub-critical reactor state. 4.1.4 Reactor Scram Syste'm A reactor protect!,ve- action interrupts the uagnet current and results in the immediate insertion of all rods-under any of the following:
- a. One out of two.high neutron fluxes on safety channels,
- b. High fuel temperature.
- c. High-voltage failure on either or both safety channels.
- d. Manual scram.
- e. Peak neutron flux (pulse mode).
- f. Minimum period (available for use as desired).
- g. External safety switches (as required).
- h. Water level (forced flow systems).
- i. Coolant flow (forced flow systems).
- j. Coolant temperature (forced-flow systems).
For systems desigr.ed for long-term, high power steady-state operation, three safety channels are used with a 2. out of 3 logic, allowing one channel to be out of service without requiring reactor shutdown. 4-9
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' SAR~9/84
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'All. scram, conditions are, automatically' indicated on the-TCRT monitor and byithe annunciators. -_ A ' manual . scram .will alsoTinsert the control rods and may,be used for~a'. normal fast shutdown of :the reactor.- On the external. switch panel, s a< bank of annunciators is available for additional auxiliary
~
scrams'or alarms.- Upper limit. power scrams are set and actuated . w'ithin the r NM-1000 safety system. Upper limit . temperature scrams
! are 3 set ist . the ~ temperature - transmitters.. However, lower
*l e v e l ,' operational scrams for power and temperature can be reset within,theLCSC prior to reactor startup. These lower l'evel, operational:scrans are actuated within the CSC logic.
- - 4 .' 1 . 5 L o n i c - F u n c t i o n s A control system logic-diagram isLshown in Figure 4-4.
Each- NM-1000 system receives its input. from a low noise 1 ~ fission ' chamber mounted adjacent to the< reactor core. For the log _ powe'r - f unction. .the chamber signal is processed to p, indicate-reactor. power from/below source level to 150% power (in excess. of 10-decades) :and. period. The system also-
' derives the cpercent power safety signal of 1% to 125% of
- fpower. :The microprocessor compares the percent power output against-a pre-determined' scram. level. .as set by an on-board
~
~ microterminal,'which when exceeded, causes an opto-isolated relay to ~ change state. For the functions u s e d - in s e c o nd ---
NM- 10 0 0, . - th e microprocessor-converts the-signal source into for- the' multi-range -linear power and 10.. linear ranges derives .a redundant percent power safety signal of 1% to 125%'of. power. The'same level comparator and self checking features are inclu'ded.
.The fuel temperature transmitters are accurate. highly stable units ' which conveir the 0-1000*C fuel temperature signal.- A .1 comparator is into a 4-10. mA output
> included which.provides scram capability .4 rough an isolated: contact state' change when the-preset level is exceeded. jThe. water temperature transmitters are standard Resistance Temperature Detector (RTD) transmitters which convert the 0 to 100'C temperature into a 4-20 mA signal. n The' transmitters have a self-contained power supply.
.A gamma' chamber provid'es the signal for peak The power (ny) nv/nyt and energy release (nyt) in the . pulse mode.
provides the high impedance . interface, high amplifier voltage and calibration circuits for the pulsing detector.
-The control rod interface accepts the digital commands from the data acquisition and control system '(DAC) to control rod motors.- It contains the operate the opto-isolation circuits which send the up-down limits and An loss.of contact signals to the control rod logic system.
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SAR 9/84 excitation power supply provides a stable reference voltage for the rod position indicator system. The magnet supply furnishes the required 200 mA needed for the rod magnets to hold control rods in contact with the armature. An opto-isolator detects the absence of magnet current to each drive magnet. The external switches are provided with terminal strips to terminate and connect various switches to the DAC chassis (beam port open-close, etc.) All of the analog signals are routed to the DAC chassis. However, the prime reactor operating signals are also sent directly to the control room. These signals include log power, period, percent power (2), fuel temperature (2), and all of the control rod position indicators. The DAC system converts the analog signals to a digital equivalent for transmitting along with the digital signals to the CSC in the control room. The DAC chassis receives control instructions from the CSC, via the communication link, which in turn moves the control rods as requested by the operator and causes the individual subsystems to go to the calibrate mode when commanded by the system or operator.- 4.1.6 Mechanical Hardware The typical reactor installation will be contained in seven NEMA enclosure junction boxes, normally installed in the reactor hall, and the reactor operator console components installed in the reactor control room. Figure 4-5 indicates the placement of the hardware items. The control console consists of the components needed by the operator for reactor control. These components include rod control switches and annunciators, the digital rod position indicators, on-line reactor status meters (power and temperature), the control system computer (CSC), reactor operating mode switch panel, color CRT monitor, printer, disc drives (2) and external switch annunciators (beam port open-close, reactor access, etc.). Enclosures 1 and 3 each contain NM-1000 high and low voltage power supplies, a pulse pre-amp with discriminator, an RMS Campbell convertor and a communications module. Enclosures 2 contains the NM-1000 microprocessor selected to provide the 10-decade log signal from the information provided by the circuits in enclosure 1. The information processed by the microprocessor is 10-decades of log power, rate of power change (period), linear percent 4-12
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. power from 1.!to 125%, . level trip s f roa c the' log and linear- I percent power,7 calibrate and' failure signals...
Enclosure 4 contains the N M-- 10 0 0 microprocessor
' selected'to provide the multi-range linear function from the.
information provided lby Mthe- circuits in enclosure 3.: In
-a dditionlto the multi-range function,;a; linear percent power signal ~ o f.. > l to 125% power is developed -along with a trip.
~
level and assoc ~iated calibrate and failure circuits. Enclosure 5 hopses the terminal' strips which interface the : digita11- signals .o f the DAC with the rod ' drives and
. external. switches (such as. beam'. port-' open-close). The magnet power - supply is also mounted in _ enclosure 5' to facilitate. easy . access to the rod drive magnets'.
.Opto-isolators provide the sensor inputs for a loss of
, magnet contact, magnet current and the up-down limit.
s wit c h e s'. .
. Enclosure 6 provides the analog interface where modular electronic. sub-sys tems convert the various primary inputs into' a 4-20 mA signal loop. The sub-systems contain their own. power supplies and comparators for providing the
-required level trips. Calibrate circuits in each sub-system respond,to.the CSC requests for a specific input value as.a check on. system performance.-
Enclosure 7 contains the DAC sub-assemblies and the communications loop to the CSC' located in the control room.- The: digital a'nd analog sub-assemblies contain calibration
. circuits which respond.to periodic or programmed calibration
~
requests from the CSC to verify proper operation and validity of.the information being handled. 4.2. Design Evaluation The TRIGA reactor console has developed through the successful' operation'of many installed facilities throughout the world. Design of the new ICS unit incorporates similar basic logic functions proven effective in prior designs. Incorporation of digital electronic techniques in the design to replace analogue circuits is justified by improved performance. Functional self-checks, circiut calibrations, and automated data logging are implemented effectively and afficiently. Installation and verification of the original design is planned ' for a reactor operated by the ICS manufacturer, GA Technologies. Subsequent installation of the ICS unit at The University of Texas facility is planned with appropriate design changes, inclusion of facility specific parameters, and completion of an acceptance evaluation. 4-14
SAR 9/84 References Chapter 4
- 1. GA Technologies, private communications.
I 4-15
SAR 9/84
' Chapter 5 REACTOR COOLANT SYSTEM
'TRIGA is designed for operation with cooling provided by natural convective . flow of demineralized water in the reactor pool. .The suitability of this type of cooling at the power levels for this TRIGA has been demonstrated by numerous TRIGA installations throughout the world.
The primary functions of the coolant system are:
- a. to dissipate heat generated in the reactor,
- b. to provide vertical shielding above the core for radiation shielding.
Heat dissipation is satisfied by natural convective flow of pool water through the reactor core and forced circulation of the pool water through an external heat exchanger. The ' pool coolant volume is composed of approximately 38.6 cubic meters in a 2x3 meter oval pool with a vertical depth of 7.4 meters. A vertical shield is provided by about 6.4 meters of water above the reactor Core. Other functions provided by the coolant system are:
- a. minimize corrosion of all reactor components, particularly the fuel elements,
- b. maintain a minimal level of radioactivity in the reactor pool water and
- c. maintain optical clarity of pool water.
These three functions are accomplished by a purification system that is included as a part of the coolant system. 5.1 DESIGN BASES The design basis for the reactor coolant system is predicated on its primary function, reactor cooling. Other coolant system functions establish the design bases for the purification circuit. u S-1 f E [
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SAR 9/84
- 5 .1 '.1 Reactor-Core' Heat-Removal To assure adequate reactor cooling, .the . e~f f ec tiv ene s s of: . natural- convective cooling has been evaluated . with
, respect to'the peak hedt flux which may be achieved in the reactor.. This evaluation then. establishes.the maximum heat flux beyond which fuel element cladding integrity cannot be assured.
Based on these evaluations. which .were discussed in
; Chap ter. 3 i' it is concluded that for steady-state operation
.the coolant inlet temperature and maximum heat flux at which f ue1~ - clad integrity' is no longer assured is determined .by the curve relating heat flux and; coolant temperature for the hottest-coolant channel. The maximum design temperature of the coolant system, coolant inlet temperature, is '120*F (48.9'C). The maximum allowable peak heat flux at this
- temperature is 325 kBtu/hr-ft2 (103 watts /cm2) corresponding to a power level of 1900 kW for an 85 element core (see Chapter'3). Since the maximum licensed power level is 110%
of design or 1100 kW, the resulting maximum heat flux will be 188 kBru/hr-ft2 (59.4 watts /cm 2
) which is well below the value at which clad integrity may be questioned.
5.1.2 Reactor Pool Heat Removal Supplemental cooling of the . reactor -pool is required for continuous operation at the rated power level. A heat
-rate of 22.3*C/ hour is expected with the reactor operated at 1000 kW. Heat removal from the pool is provided by heat
- exchange with a chilled water -supply. The chilled water.
supply is operated by the University for cooling of Research Center buildings and- equipment. Chilling capacity is provided by multiple 1200 ton (4220 kW) units. At reactor rated power the heat removal capacity required is represented by about 25% of the chilling system capacity of one unit. A tube and shell heat' exchanger is installed for heat removal from the reactor pool to the available chilled water system.
-5.1.3 Heat Exchanger Design Basis Heat exchanger capacity is designed for a stable operating temperature of the reactor pool at or below the coolant design temperature, 120*F (48.9'C) for convective reactor core cooling. The stable temperature is maintained by 'a heat exchanger capacity equivalent to the reactor core thermal. output capacity. Other heat losses such as evaporation, or heat gains such as the pump, are considered negligible.
Heat transfer is defined by q = U A 6T, (1) 5-2
m, h SAR 9/84 where U = overall heat transfer coefficient (watt /m _.c) 2 A'= surface area for heat transfer (m2) 6T, = true mean temperature difference ( *C ) For .a tube and shell heat.. exchanger the overall heat transfer. coefficient is . composed of three terms, the convective heat. transfer from the fluid in the tubes to the tube walls, the conductive heat transfer thru the tube wall, and the convective heat transfer from the outside tube wall to~the fluid in the shell of the heat exchanger. Based on the outside tube area for heat transfer, the overall heat transfer coefficient is defined as [1] U = A, . A, in (rg/r g) . 1 (2) Ahg g 2wk1 h, where A = total outside tube area (m2) A g= total inside tube area (m2) r g = tube inside radius (m) r, = tube outside radius (m) h = convective heat transfer coefficient between 1 fluid in tubes and tube wall (w/m2..C) h = convective heat transfer coefficient between fluid in shell and tube wall (w/m2 *C) k = conductive heat transfer coefficient in the tube wall (w/m2 ..C) 1 = total tube length in heat exchanger (m) A correction is applied for fouling of heat exchanger caused by buildup of various deposits. The overall heat transfer coefficient for a fouled heat exchanger is determined by U g= 1 (3) Rg + 1/U c where Rg is the fouling factor, (non-dimensional) The convective heat transfer coefficient is defined as h = Nu k (4) d where Nu = Nusselt Number 5-3
SAR 9/84 k = thermal conductivity of the fluid evaluated at the appropriate average' temperature (w/m *C) d' = tube diameter or applicable hydraulic diameter (m) The complicated nature of turbulent. flow heat transfer is described by a Nusselt. number determined by experimental correlation with the Reynolds and Frandt1 Numbers. Dittus andlBoelter [2] recommend the following- relation for fully developed turbulent' flow in tubes: 0.8 Nu t
= 0.023 Re Pr" (5) where paran.eters are measured.inside the tubes Re = Reynolds Number based on tube diameter Pr =-Prandtl Number at average fluid temperature n= 0.4 for heating n = 0.3 for cooling The relation for the'shell side of a baffled cross flow heat exchanger is suggested by Colburn [3] as follows:
Nu, = 0.33 Re Pr 0.33 (6) where parameters are measured outside the tubes and
'Re = Reynolds number based on tube outside diameter and velocity at minimum shell cross sectional area Pr = Prandt1 Number at average fluid tempersaure.
The product terms, A 6Tm, are defined consistent with
-the definition of U and heat exchanger design. .The total cross sectional area.of the tubes is represented by the heat transfer area, A, as specified by the heat transfer coefficient, U. The true mean temperature difference, 6Tm, is related to the heat exchanger ' type by a correction factor, F, and a log mean temperature difference, LMTD. The correlation relates a simple single pass heat exchanger with more complex multiple pass baffled units. A relation 'is defined by
'6T, = F
- LMTD (7) where F = correction factor [4,5]
LMTD = (T, - Tb)/in(T,/Tb) (8) 5-4
SAR 9/84 T, = (T inlet - outlet); tube side, T b
= (T outlet ~
inlet); shell side. 5.1.4 Water Purification Bases The functions of corrosion control, radioactivity control, and optical clarity of _the coolant water are provided by. filtration and ion exchange. Control of the water' purity is performed by ~ analysis of the water conductivity. Measurements of water conductivity as~ low as 2.0_micromho per centimeter ( or resistance of 1 megaohm per centimeter ) are maintained by filtration and ion exchange. The conductivity is reduced further by control of materials exposed to the reactor coolant , dust settling to the pool surface, and occasional cleaning of pool surfaces. Experience has shown that conductivities of 5.0 pmho/cm are sufficient to _ maintain acceptable limits on corrosion plus good . quality for water optical quality and removal of activation products in the water. 5.2 SYSTEM DESIGN Principle components of the coolant system are the aluminum reactor pool tank, the external cooling loop. consisting of heat exchanger and pump, and purification loop consisting of filter, resin bed and pump. Most of the total coolant ' volume is represented by the approximately_ 38.6 cubic meters of water in the reactor pool. A typical flow diagram for the system is shown in Figure ~4-1. 5.2.1 Coolant System Suction of water from the pool is provided by inlets that extend no more than 2 meters below the top of the reactor tank. The suction intake is composed of a bulk flow inlet for most of the intake volume and a limited volume flow inlet for water surface skimming. The coolant water is drawn through the coolant pump and forced through the heat exchanger. Return of cooled water to the reactor pool is provided by a discharge outlet above the reactor core or an outlet near the tank bottom. A diffused water jet is created at the outlet by a nozzle above the reactor ' core. Delay and diffusion of the reactor core convective coolant column is enhanced by the action of the coolant discharge nozzle. Accidental siphoning of reactor pool water is prevented by the presence of suction breaks on both suction and discharge lines of the coolant system. Siphon breaks are created by holes located in the lines approximately half a meter below the normal water level. 5-5
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e , :the cooling -l system , Jare located' at - about' the-same vertical
-7 . level? ' as - the : reactor core.. . Valves 'are- provided; in -the 9 M . coolant loop'for contro1~ aid isolationfofjthe co o lin g ' s y s t e'm ! ~
~ ; function.= Specificat'ionsS o f cooling ' system' components are; r 5.
~
siisted in Table ;"5--1. -A-positive iressure difference ofilpsi: '
, : ; .. .? ( 7 kil'op as c als') ~ f b~etween .Lthe- shell = sids outlet and' tube sidez
- inlet ' of' the'.. heat; exchanger is designed' toi prevent leakage
, of1 primary pool, coolant: into the secondary chilled water system.;
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.O MTable-5.1. '
REACTOR COOLANT SYSTEM DESIGN -
SUMMARY
a
'[
~ '
Reactor: Tank
' Material .
jf. ' - ' *
'Al, plate (6061)
. Thickness pj 1/4in v. (0.635 cm) 3 i
Volume. (maximum) 10525: gal. (39.85 m-)- .. J r F-Al minum (6061)
- Coolant: Lines j Valves. Aluminum
~
- Diameter-31n.'(7.62 cm) -
Coolant-Pump' ~
, LType Centrifugal: 9 -
Material ~ Stainless steel S Capacity- 375 gpm:-(23.6 ~ 11t,er/sec)' z
- m. Heat' Exchanger ~
., s
-Type. .
# Shell and-tube * '
, u: Materials: shell Carbon steel
- tubes '304-stainless steel
, , Heat Duty 1000 kW '~ Flowrate:-tubes 225 gpm.(23.6' liters /sec) shell 375 gpm ( 14'. 2 ' 11t e rs / s e c ) Typical Parameters:- Tube' inlet 100*F '40~: psia Tube-outlet' 68'F 30 psia-She11' inlet 45'F 65 psia _Shell outlet 67'F 45 psia
- 1r l
?
5.2'.2bPurification System i A . nurification loop is U sur . ated into the cooling Tsys' tem. The loop' bypasses the heat e m.c h a n ge r and.is located Ta t ' ' ab ou t the ^same ' vertical location-as the heat exchanger. H A-portion ofithe cooling system flow, less.than 10 gal / min p q V , 5-7 N 1 - s..-..s....a_-_.....-._.,_..___.__.___.__..-...- - . , . _ _ . _ . ,
.- .~ ~- . .. . - . _ - .. - . . - _~
~ 7 SAR 9/84
- (0. 6 . liters / sec) , is diverted through the purification loop
' during? foperation of 'the cooling system. A small t
. j purification' loop . pump is operated when the cooling system
- is'not in' operation-to allow continuous' removal of suspended-particulates and' soluble ions from the water coolant. Water suction:and~' discharge is. accomplished'by the same. lines that are'used.by -the cooling-system. The purification system is
, operated either independently or in. conjunction with the cooling-system. l Purification-functions.of the loop are generated.by two-
. components , - a filter.for removal of, suspended materials and for removal of- soluble . elements. . Typical' a resin - bed 1 filtration .is provided with 25 micron filters. Typical ion exchange is - provided by .085 cubic ' meters of mixed cation and ' anion - re s.in .- . Water purity is-measured by conductivity at the inlet and outlet of the resin beds.
cells Purification loop flow-rate is indicated by a flow meter so that: flow rate through the resin bed is. controlled. , 5.2.3 Water System Instrumentation Several~ monitoring ~ sensors are installed to allow
.' ~ remote . readout of ' water system parameters in the reactor control room. Other system , parameters are indicated' by4 local monitoring . devices. Parameter monitoring points are 4-1. The parameters. that are
, illustrated in. Figure-considered - part: of th e .wate r system instrumentation system are presented in-Figure 4-2.
- Indication o f - t h'e reactor pool status is determined,by two sensors located . in the - pool. Pool level and bulk-pool
- . temperature sensors'in.'the pool'are monitored in.the contro1~
rcom. An. annunciator and alarm indication'are generated'by high.or low pool levels and by high' pool temperatures.- f Measurements of cooling system function - are indicated 4 by two temperature probes, one on the pool suction-line.and are one on the pool discharge line. Both temperatures ' observed on the< bulk pool temperature meter by actuation of a momentary contact' switch. Typical temperature probes used are resistance temperature detectors (RTD's). Water quality.is. determined from two conductivity cells " in the purification loop. The cells are located on inlet ' and -outlet line s-'o f ' the' demineralizer with arereadouts' designedlocated-in' the control room.- Conductivity cells A with Wheatstone bridge platinum electrodes' shielded by glass. circuit 'in the control room is connected to the cells byThe a
. switch for selecting - inlet or outlet conductivity.
circuit is compbsed of a rectified power supply, one stage amplifier, temperature compensation and electric tuning eye. p 5-8 p
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(Heat exchangerioperation~and coolant flow indication in-the: coolant; room. is. generated from two pressure taeasurements. ,0ne pressure . sensor ~ in the coolant. line is
' ~
' installed for indication of coolant flow.- . A pressure emonitor isLprovided.'with.a trip to stop-the coolant pump on
;1oss 'of flow. pressure. . An alarm in'ication d from the:
pressure l o s s -'c a u s e d by- flow : loss is transmitted' to .the
~ ,
reactor control room.- A_second pressure-measurement of the 1 -. p re s s u r e difference be tween -- the high pressure point lon the heat; exchanger tube side and low pressure-point on'the heat
; exchanger _ shelli side 'is ~made with a pump - trip provided on Jloss- of positive- . dif f erential - pressure.
The differential pressure is; designed for a difference substantially greater than :7 kilopascals-(1 lb/sq.in.). _ Alarm indication'of loss iof. pressure. differential'is provided in the reactor control room.
.Several water system parameters are measured by- local pressure or temperature . sensors in the _ system olines. Both; temperature and pressure probe are located on the inlet and outleti lines .of the. pool water-side and' chilled water side of the heat. exchanger. A_ local indication of flow in the~
coolant loop is provided .b y ' t h e pressure drop across an orifice .in_ the flow path. Purification loop flow 'is measured by an in -line . flow meter. Water pressure before
'and af ter the . filter in the purification loop is measured for indication of filter condition.-
-5.3. WATER SYSTEM DESIGN EVALUATION The _ water system including the reactor pool and the x -external' cooling'and purification;1 oops.'have similar design features as used in many other operating'_TRIGA facilities.
-- The: demonstrated capability . and integrity of this system m 1provides ' assurance - that- the coolant--system will perform its function properly and safely.
' Availability of pool water for cooling and vertical shielding -is assured ' by designing _ the system: with siphon-breaks'on suction lines and-discharge' lines within 2 meters of the normal pool level. _ Greater: losses of pool water are extremely , improbable, although they could conceivably be ;
initiated by. rupture of'the reactor tank. As shown in the ' l loss of. Lpool ~ water accident analysis, even with complete ~
'{'
loss.of' pool water fuel clad-integrity.is not threatened. Adequacy ( of reactor cooling- is assured by the- large amount ~ of cooling capacity inherent in the reactor pool volume as - well as the capacity of the external cooling circuit which .can dissipate- ' heat at a rate equivalent to. 1000, kW- steady-state operation. If the available heat ; exchanger capacity is ~ diminished to 900 kW and the initial j pool temperature is 100*F, the reactor can be operated for j 5-10
y y SAR-9/84'
~ , ,
2 more thant5 hoursfbefore'the-bulk-pool temperature reaches
~120*F. The' actualitime would be considerably longer 'since
.as the' bulk pool . temperature increases t h'e . h e a t exchanger cheat removal: ' c apacity increases. Without external cooling or. L other heat- loss the bulk. pool: ' temperature will rise approximately 22.3*CT .after one. hour -of~ operation at ni 4 steady-state power level-of"1000 kW(t).. _
h ~
~ ' Heat; removal capacity -and thus
' pool ~ heat rate is specified i by Eanalysis of a-six' baffle tube and shell-= heat exchanger. At a flow rate of 250 gal / min (15. 75 : liters /sec)
~
of. chilled ' water a t. 45'F - (7. 22* C) a heat removal rate of
.10 5 0. ; kW is expected.- The - presence ~ of fouling in the heat
-exchanger is. considered
.two . heat ' exchanger fluids.
minimal based on the purity of the Capacity is reduced to 890 kW
. f o r- a fouling factor -of .0004. The heat transfer and hydraulic parameters used are shown in Table 4-2.
~
Experience ' with this purification equipment in other TRIGA = systems has shown. that . coolant ' conductivity can be easily maintained at levels of less than five micromhos per centimeter using system design.' Furthermore',the materials. contained in the coolant-this experience has shown that no apparent corrosion of. fuel clad or other components will occur.ifuthe conductivity of--the water does not exceed' five micromhos per centimeter when averaged over a 30-day' period. Control of radioactivity. in the coolant is provided by the > purification system. 'Should radioactivity be . released ' from a clad leak or rupture of an experiment, detection of. the - release would be signaled-by the continuous air monitor or by. the reactor - room area monitors, Based on coolant
- transport
- these time calculations in the' safety analysis section, monitors should register an increase in coolant radioactivity within approximately.60 seconds of the, time of-ra'dioactivity release. The transport time is estimated from the ' time f or the coolant exposed in the core to reach the.
surface of'the water.where'the continuous air monitor will detectDa release of radio ac tiirit y from the pool water. An alternate indication of radioactive release is provided if a water activity monitor is installed or by a GM detector area monitor.- 5-11
m - 1 SAR-9/841 o Table 5-2 HEATLEXCHANGER HEAT TRANSFER AND HYDRAULIC PARAMETERS L Tubes:
- Outside-Diameter- '0.750 inch (1.91 cm.)
Wall. Thickness '0.049 inch-(0.124 cm.-)
~
--Thermal Conductivity - 8 .' 21 Btu /hr, ft. 'F Flow Areat Tube Side 8.0 in2 (51.'6 cm2) s Shell Side 133.7 in2 (217 cm2)-
Heat Transfer Surface '341 ft 2 (31,7 m2)
. Average Prandt1 Number-Tube ~
5.69
-Shell 9.07
'Reynolds Number-Tube 6.33 x 10 4
-Shell' 4 1.80 x 10 Corrective Heat Transfer Coefficients Inside Tubes 1755 Btu /hr'ft 2.y (9966 w/m2.C)-
~0utside Tubes 1316 Btu /hr ft 2.p LMTD 22.7 oF (12.6*C)
Corrective' Factor F. 0.83 Average Kinematic Viscosity
. Tube -0.9 x 10 -5 c 2 7,,c (8.36 x m 2/sec)
Shell - 1.'3 5 x 1010~5 g 2 (1.25 x 10-6 ,e2/sec- fgec) I 5-12. E i E_ . I
6 -- SAR 9/84-A Chapter 5 References
'1. Holman, J.P., " Heat Transfer", McGraw-Hill,
' Edition, 1976, pp 386-321. Fourth
.2. D i t t u's , F.W. and Boelter, L.M.K.,
_(Berkley) Pub. Eng.", vol. 2, pp 443, "1930. Univ. California
- 3. Colburn,. A.P. "A Method Correlating Forced of Convection Heat Friction", Transfer Data and Comparison with Fluid <
Trans. AIChE vol. 29, pp 174-210, 1933.
- 4. Bowman, R. A. . Mueller ,
A.C., and Nagle, W.M.,
"Mean Temperature (1940), pp 283-294 Difference in Design", Trans. ASME, vol. 62 5..
Tubular TEMA/3rdExchangerEd.", NewManufact'rers u York, 1952. Association, " Standards
- 6. Kreith, F.,
1976, pp 557-560. " Principles of Heat Transfer", Third ed.,
,i i
s
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i t 4 5-13
SAR 9/84 Chapter 6 FACILITY DESIGN The. TRIGA Mark II reactor is located in a special
. laborai;o ry bay of an engineering laboratory building. Most of ', the building. design is determined by criteria that are not_necessarily directly related to the reactor. However, several design features are incorporated. to assure . safe facility operation and effective utilization of facility
. equipment.
Structural , engineering design of the building is
.specified by standard University procedures developed from the Uniform Building Code and State of Texas Building Code.
All elements are designed for seismic activities specified for Zone O conditions. The provisions of the Lif e 'Saf ety Code and National Fire Protection Code are . included in' building features. The - building site is located above a rock- subsurface composed of limestone that will accommodate . substantial loads (1690 kg/m2). Building foundation is " composed of poured concrete piers with concrete slab on compacted fill. Building superstructure is constructed of reinforced concrete for columns, beams, floors and roof. Exterior structure walls se fabricated of precast concrete slabs. Building orientation, floor plans, and a section of the reactor . bay area are shown in Figures 6-1 thru 6-5. Total floor . space of the facility is approximately 1675 square meters. Areas of the building include office space, general laboratory areas, specialized laboratory areas, support areas and the' reactor facility. Some operations of the reactor are supported by specific' f eatures or functions of areas not within the reactor bay and control area. Shop areas for mechanical or electrical work and laboratories for radiological measurements are operated within the building for activities of both the reactor operation and education, ' or. applications in nuclear engineering. 6.1 DESIGN BASES 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 6-1 ,
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SAR 9/84 by physical containment and surveillance. The physical containment will also control the release of radioactive accident materials during routine operation or potential conditions. Release of airborne radioactivity consists mostly of air activation products from routine operation or fission product materials from a non-routine fuel element failure. Liquid and solid radioactive material are also controlled to assure compliance with appropriate release criteria standards. Other potential releases may be associated with specific types of experiments that require special equipment to provide sufficient control designedof material to releases. The reactor room containment is control the exposure of operation personnel and the public from radioactive material or its release caused by reactor operation. Release criterion are based on Title 10 Chapter 20 of the U.S. . Code of Federal Regulations [1]. 6.2 REACTOR BAY AND CONTROL AREA Reactor tank, shield and primary experiment facilities are located in a reactor bay area that is roughly 18.3 meters on each side. A total of 4575 cubic meters of volume is enclosed in the reactor bay above the 335 square meters of floor space. Support of reactor operation and of reactor experiment activities is provided by a control area located adjacent to the reactor bay. Space in the control area is divided into a conference room, office, control room and entry way. Total control area (7.3 by 18.3 m) is 134 square meters of floor space and roughly 489 cubic meters of air space. The floor level of the reactor bay is one level below the office and laboratory areas located on levels two and three of the building. 6.2.1 Physical Design Design of the TRIGA reactor facility area is specified by constraints on the function of the architecture design, access control for security, radiological control for safety > and applicable building code standards. All access points to the reactor bay are located inside the engineering building and enter from the north side. The remaining three sides of the reactor bay area are enclosed by exterior walls. Both emergency exits and equipment bay doors on the first level open into the adjacent area within building from which building exits are accessible. the Adjacent areas on the north side of the reactor bay provide some laboratory support functions in conjunction with other building support functions. On the third level from the reactor floor the adjacent area to the reactor bay is supplemented by the control room area, conference area operation office, and routine entry point. The third level entry way is provided for access to the control area from 6-7
SAR 9/84 the laboratory building and access to the reactor bay from the control area, Access at tha third level is to the top of the reactor shield structu... A stair structure is attached to the reacor shield with a supplementary access point to the reactor bay on the second level. Design of access points and interior walls are specified for security control, fire control, and ventilation control. Penetrations, besides the doors, into the reactor bay and control areas are limited in size and are sealed to limit air leakage. Details of the reactor bay area are presented in Figure 6-6. A 5-ton bridge crane is installed in the reactor bay for movement of shield structures, heavy equipment and fuel transport loading. 6.2.2 Ventilation Design Ventilation' design is specified to isolate the reactor bay area in the event a radioactive release is detected in the reactor area. The ventilation system is designed to maintain a negative pressure within the reactor bay with respect to the building exterior and other building areas when the high volume system is operating. During normal operation ventilation is exhausted through a roof stack. Isolation of the room is initiated by a signal from the detection of high radiation levels by an air particulate monitor. Isolation is achieved by air control dampers and
. leakage prevention material at doors and other room penetration points.
Schematics of the ventilation system for he reactor bay area and a diagram of the ventilation control system sensors and controls are provided in Figure 6-7. An auxiliary low volume room air purge is configured for supplemental movement of air in the reactor bay. The purge system is designed for the potential installation of High Efficiency Air Particulate. Operation of the purge system is controlled by an air activity sampler with the air exhausted t- the roof stack. Routine operation of the purge system is d signed for ventilation of the radioactive gases from the laboratory. A schematic of the auxiliary ventilation system is shown in Figure 6-8. 6.2.3 Reactor Shield Structure Reactor tank and shield design are specified to provide a reliable containment for the reactor coolant and an effective shield for reactor radiations. Shield thickness is determined by radiological exposure constraints with a goal of 1 mrem /hr specified for most accessible areas of the shield. Radiation doses above the shield and at selected locations are expected to exceed the design goal. I i 6-8
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SAR 9/84 Composition and design of the shield are also analyzed for structural properties such as seismic response. Special features of the shield such as beam tubes, are designed with 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. Other shield penetrations are designed for appropriate radiation shielding. Features of the shield structure are illustrated in Figure 6-9. 6.3 SUPPORT FACILITIES 6.3.1 Radioactive Waste Control Gaseous radioactive effluents from the operation of the rotary specimen rack and pneumatic irradiation tube are filtered with HEPA filters and vented through the auxiliary ventilation system. Argon-41 from reactor beam tubes is vented via a collection manifold that also exhausts to the auxiliary ventilation system. A system for retention of liquid radioactive waste is installed on the first level of the building. An area separated by a chain link enclosure from the building loading area is provided for liquid waste storage tanks and solid waste storage barrels. The area is designated for temporary storage of materials of low radioactivity and is I not expected to require shielding. Temporary shicid block will be installed as required. Liquid waste from the sink and shower of the Sample Handling Laboratory and drains from the reactor shield, reactor bay and liquid waste storage area are connected to a 1000 gallon retention tank in a pit exterior to the building. A sample line, dilution line and discharge line are connected to the retention tank. Discharge from the tank is accomplished by a pump with gate valves to route the discharge to supplemental storage or sanitary sewer. Schematics of the gaseous and liquid radioactive effluent handling systems are presented in Figure 6-10. Solid radioactive wastes are segregated for disposal and transport to an approved radioactive waste burial facility. 6.3.1 Sample Handling Laboratory A sample handling laboratory is situated adjacent to the reactor facility on the third level for the processing of radioactive samples and materials. Access ports via air or gas transfer tubes are installed to move samples between the reactor area and the Sample Handling Laboratory. Two separate sample transfer systems are provided, one for the 6-12
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REACTOR SHIELD STRUCTURE Figure 6-9 6-13
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= I enesecass m so nner
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DOUID WASTE stAseDLING SY3 TEM SCHEMATIC DeAGRAM . >, . t W l t
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f~~ SAR 9/84 e pneumatic tube irradiation facility and one for loading the rotary specimen rack facility. , A hood for handling radioactive materials, a sink for disposal of radioactive liquids and a safety shower for decontamination are installed in the Sample Handling Laboratory. 6.3.3 Health Physics Laboratory A Health Physics Laboratory is situated adjacent to the reactor facility on 'the second level. Radiation counting systems for evaluation of radiation exposure or contamination, are maintained in the Health Physics Laboratory. Equipment such as a thermoluminoscent reader and an alpha-beta proportional counter are maintained in the laboratory. Other equipment and supplies operated or stored in the Laboratory are as portable radiation monitors, coveralls, gloves and related items, d 6.3.4 Other Laboratories Neutron activation of samples is supported by a Chemistry Laboratory for special sample preparations and a Gamma Ray Spectroscopy Laboratory for evaluation of sample irradiations. Student instruction and research areas with radiation counting systems are located in the Radiation Detection and Measurement Laboratory, and a Undergraduate Equipment Laboratory. Supplemental activities are planned for the Radiochemistry Experiment Laboratory. A major experimental facility of the engineering laboratory is included in the shielded Neutron Measurement Laboratory. The area is a cubical area roughly 9 meters on a side with 1.2 meter thick concrete walls. A control room is located near the measurement area for the control of accelerator generated neutrons and acquisition of experimental data. ; 6.3.5 Support Areas Several areas adjacent to the reactor bay on the first level and the second levels are intended for support of some ; reactor functions such as air ventilation and repair or ' assembly areas for mechanical and electronic equipment. A staging area on the first level is designed for heavy j equipment access to both the reactor bay and the building. 6.4 SPECIAL EXPERIMENTAL FACILITIES 6.4.1 Reactor Core Facilities ! l 6-15 , t
SAR 9/84 4 Reactor core experiment facilities are designed for replacement of either single fuel element positions or a special multielement position. Access to the core peak flux is provided by a central thimble. The wet central thimble is designed to allow insertion of an encapsulated sample into the core center or extraction of a vertical neutron beam from the core center. Two experiment facilities are located in single element positions in the reactor core for insertion of samples into the reactor neutron flux. A pneumatic terminal is provided for short irradiations of small samples and a dry tube is provided for long period irradiations of small samples. Experiments with reactor characteristics are demonstrated by a reactivity oscillator apparatus. The , oscillator is fabricated with rotating absorbers that are inserted in a single element position. Multielement experiment locations in the reactor are positioned in the cor center or near the core reflector interface. Fabrication of experiments for the multielement positions are projected for future facility development. 6.4.2 Beam Tube Facilities Access to horizontal neutron beams is created by five beam tubes penetrating the reactor shield structure. All beam tubes are 6 inch diameter tubes originating at or in the reactor reflector. One tangential beam tube is composed of a penetration in the reactor reflector assembly with extensions through both sides of the reactor shield. A second tangential beam tube penetrates and terminates in the. reactor reflector. The two remaining tubes are oriented radial to the reactor core. 6.4.3 Cobalt-60 Irradiation Facility A cobalt-60 irradiator consisting of 156 pencil size sources is installed in the reactor coolant pool. Each source is contained in a space 0.5 inches diameter and 11.25 inches length. Sources are doubly encapsulated with inner aluminum clad and outer stainless steel clad. The source pencils are arranged in a double staggered, circumferential array with an inside radius of 10.5 inches and outside radius of 12.5 inches (Figure 6-11a). The total source strength when installed was 9260 Curies. A shelf suspended in the pool at the end opposite to the reactor location holds the cobalt-60 irradiation facility below 10 feet of water. Access to the facility is provided by water tight canisters or dry irradiation tubes. Two canisters with 0.625 inch wall thickness, 5.0 inch radius and 11.5 inch height and several 2 inch diameter vertical beam tuben provide exposure volumes for routine experimentation. 6-16
SAR 9/84
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l
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. C. 04AMeft AtM4 Cit 0 ASMuttV SPECIAL EXPERIMENT FACILITIES Figure 6-11 6-17
SAR 9/84 6.4.4 Suberitical Reactor and Moderators Cylindrical assemblies of graphite and polyethylene are utilized for student laboratory experiments with neutron sources and a suberitical uranium-235 reactor assembly. The plutonium beryllium neutron sources and uranium dioxide used in the polyethylene suberitical assembly may be stored and used in the room containing the reactor but are licensed separately from the reactor. The suberitical core and moderator assemblies are products of Lockheed Nuclear Products (Figure 6-11b). The suberitical polyethylene core is a cylinder 10 inches in diameter and 14 inches long. neflector assemblies can be assembled with or without the fueled core. Dimensions of the cylindrical reflector assemblies are 30 inch diameter by 34 inch length for the graphite moderator and 22 inch diameter by 25 inch length for the polyethylene moderator. An additional graphite moderator cylinder 30.5 inches high by 24 inch diameter is available for neutron source moderation. 6.4.5 14 MeV Neutron Generator A small accelerator is designed for the production of D-T reaction neutrons for research measurements and activation experiments. Application areas of the source of neutrons are proposed in neutron dosimetry, neutron activation, and neutron interactions for analysis of related research problems. 6.5 CONTAINMENT DESIGN EVALUATION Containment evaluation depends on the quantity of airborne radioactivity release possible from the air and water that are in the region of the reactor during operation. Calculation, measurement, and experience of similar research reactors support the evaluation. Evaluation is limited to routino effluents and should be supplemented for experiment conditions that present specific release problems. Analysis of fission product releases are treated in another chapter. The most significant radiological effluents of the reactor are argon-41 and nitrogen-16. Measurement and experience of other facilfties have shown that for a facility of this type the most significant routine radiological contributions are caused by argon-41 generated by the exposure of air in experimental facilities and by nitrogen-16 transported in the coolant from the reactor core region. Argon-41, a noble gas, is contained for decay or eventually released to the atmosphere for 6-18
SAR 9/84 decay. Nitrogen-16, a dissolved gas, is contained in the coolant and generally is dissipated by radioactive decay in the coolant. 6.5.1 Release of Argen-41 and Nitrogen-16 from Pool Water Argon-41 is produced in the reactor pool as a result of the (n,y) reaction with argon-40 dissolved in the pool water. Most of this argon-41 remains in solution but some of it .is transferred to the reactor room air at the pool surface. Calculations of the argon released from the pogi aurface estiate a concentration in the room of 1.6 x 10 uCi/cc with the reactor operating at 1000 kW. The nitrogen-16 produced through the (n.p) reaction with the oxygen in the water molecule has a very short half-life (7 see) so only a very small fraction of that produced in the core will find its way to the pool surface. The principal radiological effect of the nitrogen-16 is as a contributor to the radiation level at the pool surface. Calculations of the nitrogen-16 transported to the pool surface estimate radiation dose rates of between 16 to 400 mrem / hour with the reactor operating at 1000 kW. 6.5.1.1 Argon-41 Activity in Reactor Room. The argon-41 activity in the reactor pool water results from irradiation of the air dissolved in the water. The following calculations were performed to evaluate the rate at which argon-41 escapes from the reactor pool water into the reactor room. The calculations show that the argon-41 decays while in the water. The changes in argon-41 concentration in the core region, in the pool water external to the reactor, and in the air of the reactor room, are calculated using the variables as defined below: N atomic density (atoms /cm3) A decay constant (sec~I), 1.06 x 10~' o absorption cross section (cm2), 0.61 b q volume flow rate, reactor room exhaust (cm 3/sec) V volume of region (cm 3) p density (gm/cm3) A f channel free flow area (cm 2) i channel length (cm) w mass flow rate (gm/sec) 6-19
6 SAR 9/84 3 y volume' flow rate through the core (cm /sec) - I average thermal neutron-flux in the. core (n/cm -sec) The volume flow rate through the core is g/sec 3 g., , ,8000 = 8.0 x 10 cm3 /sec (1) D 3 1 g/cm
'From the f1w channel volume, A i the exposure time in'the core is f c' t = V/v =Aic/ v =
g 485.0 x 38.1 = 2.3 sec- (2) 3 8.0 x 10 It remains to find the atom density N for dissolved argon in the reactor pool water. According to Henry's law for gases in contact with liquids the equilibrium concentration in the liquid is proportional to the partial pressure of the gas. _ The saturated concentration of argon in watet at one atmosphere , of standard air is given in Table 6-1. Table 6-1 SATURATED ARGON CONCENTF.ATION IN VATER l' ) Temperature 3 ('C) S(atoms A-40/cm H2 0) 16 10 1.14 x 10 16 20 0.94 x 10 16 30 0.79 x 10 16 40 0.69 x 10 16 50 0.62 x 10 16 60 0.5( x 10 ! 70 0.52 x 10 16 16 80 0.48 x 10 l The argon-40 concentration (N) in the water at the core inlet temperr.ture; (38'C) is , t 6-20
- o. '
n' " - 1
- u .
-SAR'9/84:
N^=;.7.'l x 10_5 1
,g,,,fc,3' .
-(3)
" ~
'he~
T concentration of argon is.5.3 x 10 15: atoms /cm3 at
.6 7 .~8
- C , .the.: core exit. water temperature.
The - argon-4 '1 d e n s i t y. . . in water-(at equilibrium) at.the. exit lfrom the core-is given by
.A. =.A,e + NJoe ('l - e- t). ,
.[4).
andJat:the entrance 4 LA , =-A ,e7 , (5) where t:ils.the exposure-time - in the core =(2.-3 sec), and T is the cycle time in the pool. The average.out-of-core cycle time T is given by T = b == 3.86'x-10 7 cm =-4.83 x 10 3 sec , (6) v' 8.0 xL 10 3- c,3/s~ec
- where V is the pool volume and v is, again, the volume flow rate tRrough the core. The solution to this set' of equations is A, = N 04 1-e . (7)
.g , , -A(t + T)-
4 Substituting the values from above one obtains
'Ak = (7.'1 x 10 5) (0.61.x 10-24) (1.2 x 10 13) 1' exp'{-(1.06 x 10-4) (2.3)]
1 - exp [-(1.06 x 10) 4.83 x 103 )] A =~31.6 dps/cm 3 , and'N 43
=
31.6 / 1. 0 6 - x 10~4 = 2.98 x 105 ,go,,fe,3 ,
.One source of argon-41 in the room results from the reduced solubility of argon in water as the temperature increases. Considering the expected temperature rise of the water passing through the core, an immediate release of about 25% of the argon-41 made could be expected during passage. Some of this argon-41 might be redissolved as it is transported into cooler water, but since the cooler water is in' equilibrium with the air above, it is nearly saturated with argon and uill not absorb all of the argon released.
Measurements of argon-41 in the water as a function of height above the core indicate that approximately 60% of the release argon-41 is reabsorbed. 6-21
.(
SAR 9/84 A-combination of the two. sources of argon-41 mentioned above will -give the upper limit' of the fraction of total j argon atoms that can leave the water per second. 1 Assuming that the' 10% (i.e., 40% of 25%) of the argon-41 comes out of solution, remains undissolved after leaving the core, and escapes to the air, this source would be S g = 0.10 N47 v
= 0.10 x 2.98 x 10 5 atoms /cm x 8.0 x 10 3 cm 3 7 ,,c
= 2.38 x 10 0 atoms /sec The tendency. of the balance of- the argon activity in the pool to escape to the air owing to its proximity to the water-air boundary will constitute the additional source of argon-41 at the water surface.
Estimates of the surface. exchange coefficient B (i.e., the' gas in a unit volume that is exchanged at the surface per.-unit time per unit surface area) for argon vary considerably. One method of arriving at a value f or this parameter is through the diffusion coefficient of the gas in
. water. The mean square distance traversed by a molecule is
<6X>2 = 2Dt .3 (8) where D = diffusion coefficient (cm /sec).
t = time (sec). , The exchange coefficient is assumed to be evaluated for.I sec. as t' B= (<6X>2)1/2j t, = (2D/t)
! . (9)
The diffusion coefficient at 40*C is about 1.1' x 10 cm 2/sec, and, if one assumes that..only one-half of the argon. atoms within one diffusion length of the surface. escape, 8 ='I/2 (2 x 1.1 x 10-5)1/2 = 2.35 x 10 -3 cm/sec 6 Measurements have been made of the argon-41 activity in a'TRIGA Mark III reactor pool and from the data acquired from these measurements it was possible to construct a value for the surface exchange coefficient. This value at 40*C is about 2.9 x 10-4 cm/sec. Values for the surface exchange coefficient have been reported by Dorsey [3] for air, 02, and N2'e The values for these three gases are all about equal. ' Assuming argon 6-22 st
E' , . - SAR 9/84
-behaves'as do these gases, a value is obtained'of 5'.7 x 10-3
~
t Tem /cecifor 8.
~ Inc this analysis the largest of the three. values wag Tused for; the . surf ace ' exchange coefficient, i.e. 5.7 x ' 10 j ..- 'em/sec.,
. . The ; rate" at which argon atoms are transf erred ' across the water-air interface'is determined by-2L = 0.93 BN 4g g (10)
S A
= 0.93 x 5.7 x 10 -3 x.2.98 x 10 5 x 5.05 x'10 0
= 7.98 x.10 7 atoms /sec- ,
. where A g is the-surface area of the pool (5.05 x 10 4 .cm 2).
The -total. transfer rate of- argon-41 nuclei t o' the 4 reactor room is 41 S = S g.+.S2 =.2.38.x'108 + 7.98 x 10 7 (11)
= 3.18-x 10 8 nuclei /see .
_ Air is; removed at the rate of qfem /sec from air-filled volumes from which the- accumulation of argon-41' as a
. function of' operating time is given by.
IdN 41- 41 dt
=S /V:- (A 41 + q/V)N-~41 ,
(12)
. where-S: is the source of radioactive atoms released:to the
~
air.
' Integrating from N41 = 0 at t = 0,'one obtains, 4l N41(g) ,
S /V (1 _ (,xp_(g41 + (13) 41 /V)t]) , A +=q/V i Por prolonged reactor operation at a steady power level in-which
-t >> 1/(A 41 + q/V). ,
1
- the source: term and removal rates of- argon-41 are in
.. equilibrium, in which case 41 N 41 8 /V 3
=
41
< atoms /cm . (14)
A , qfy
.The argon-41 concentration in- the reactor. room and building._ exhaust-air 6-23
SAR 9/84 N R'= = (15) A-+ q/V R YR*9 0 3.18 x 10
-4 5
=
578'at./cm 3' 1.06 x 10 x 4.12 x.109+ 1.14 x 10 3 where the effective room volume is 4.12 x 10 9 cm and a room air exhaust rate -assumes that the air exchange' rate is two per hour'and that the effective. room volume is 90% of.4575
- cubic meters. This corresponds to an activity concentration in the room of
-4 10 x 578 -6 A=fE=1.06x = 1.65 x 10 Ci/cm 3 . (16) 3.7 x 10 The actual effect of argon-41 releases from the reactor . pool would be substantially less than those estimated as a -result of the various conservative estimates in the calculation. Among the maj or conservative assumptions are the transfer amounts of argon from the pool. surface, period 'of full power operation, release rates and volumes.
6.5.1.2 Nitrogen-16 Activity in Reactor Room. The cross-section threshold for the oxygen-16 (n.p) nitrogen-16 reactions is 9.4 MeV; however, the minimum energy- of the incident neutrons must be about 10.2 MeV because~ of center of mass corrections. This.high threshold limits the production of nitrogen-16 since only about 0.1% of all fission neutrons-have an energy in excess of 10 MeV. Moreover, a single hydrogen scattering event will reduce the energy of these high-energy neutrons to below the threshold. The. e f f ective cross-section of oxygen-16 (n.p) nitrogen-16 reaction averaged- over the TRIGA spectrum- is 01.021 millibarns. This value agrees well with.the value obtained from integrating the effective cross section over the fission spectrum. The concentration of nitrogen-16 atoms per em 3 of water.
- as it leaves the reactor core is given by I * -A 2= 1 (1 -e 2' ) , (17) 2 where N2 = nitr gen-16 atoms per em of water, 13 n/cm 2-sec,
= neutron flux = 1.0 x 10 (0.6 - 15 MeV),
N g = oxygen atoms per em of water = 3.3 x 10 22 #/cm 3 ,
-o g = (n.p) cross section of oxygen = 2.1 x 10 -29 c,2 6-24
SAR 9/84
-(averaged over 0.6 - 15 MeV)'
-2 -1 A 2.= nitt gen-16 decay constant 9.35 x 10
= sec ,
t = average time of exposure in reactor.
.The average exposure time in the. core (2.3 sec), was derived in the discussion on argon activity. Solving for N in the equation above, one obtains 2
-2 2= 7.41 x 10 7 (1 -9.35 x 10 x (gg)
- e 2.3) 7 3
= 1.43 x 10 atoms /cm as the density of nitrogen-16 in the water leaving the core.
If it is assumed that the water continues to flow at the same velocity to the surface, a distance of '640 cm, the transit' time from core to surface is t rise
= 640 = 38.1 sec , (19) 16.8 where the flow velocity, 16.8 cm/sec (Table 3-7), was given in the discussion on heat transfer.
This assumption is quite conservative as energy losses from-the fluid stream resulting from turbulent mixing will reduce the velocity significantly. Furthermore, delays in transit time resulting from operation of the diffuser pump are sizeable. Measurements made of the dose rates at the pool surface of several TRIGA reactors show that the operation of the diffuser pump reduces the nitrogen-16 contribution to .the surface dose rate by an order of-magnitude of more depending on the size of the pool. In 38 seconds the nitrogen-16 decays to 2.86 x. 10 -2 times the value of the activity leaving the core. Thus the concentration of nitrogen-16 atoms that reach the region near the3 surface of the pool is estimated at about 410,000 atoms /cm per cubic centimeter. Only a small proportion of the nitrogen-16 atoms present near the pool surf ace are transferred into the air of the reactor room. When a nitrogen-16 atom is formed, it i appears as a recoil atom with various degrees of ionization. { For high purity water (approximately 2 umho) practically all } of :the nitrogen-16 combines with. oxygen and hydrogen atoms l of the water.- Most of it combines in an anion form, which has a tendency to remain in the water (4]. It is assumed that at least one-half of all ions formed are anions. . Because of its 7.1-see half-life, the nitrogen-16 decays before reaching a uniform concentration in the tank water. l The activity will be dispersed over the surface area of the pool and much of it will decay during the lateral movement. 5 6-25
- r ll
SAR 9/84 For the purpose of the analysis it is postulated that the water-bearing nitrogen-16 rises from the core to the surface and then spreads across a disk source with an equivalent radius of 125 cm. For a constant veloc1Ly of 16.8 cm/see the cycle time for distributing the nitrogen-16 over the pool surface would be t,
= 125 cm/16.8 cm/sec = 7.4 sec . (20)
The average concentration during this time is 8 - t N = 1/t s Ne o dt , (21)
= N, (1 , ,-Ats) = 4.10 x 10 5
(1 - e -0.69)f7,4
-2 s 9.35 x 10 5 3
= 3.0 x 10 atoms /cm The interest from the point of safety is then the number of nitrogen-16 atoms escaping into the air from the diffusing source above the core. The number escaping to the air would be about is estimated from the escape velocity, 0.009 cm/sec, from Dorsey [5] as (3.0 x 10 5
atoms /ce) (0.9 x 10 -2 cm/sec)
=
2700 atoms /cm2-sec . In the room, the activity is affected by dilution, ventilation, and decay. Thus the rate of accumulation of nitrogen-16 in the room as a whole is given by
/V) VN 16 N +
d(VN 16) , 3 _ (z , (22) dt where S = number of nitrogen-16 atoms entering the room from the pool per secogd, 8 atoms /sec, (2700) (5.05 x 10 ) = 1.36 x 10 9 V = volume of the reactor room, 4.12 x 10 cm 3, (effective), q = volume flow rate, 2.29 x 10 6 3 cm 73,,, (reactor room exhaust). For saturation conditions 8 VN 16 , S , 1.36 x 10 (23) (A N
+ q/V) 9.35 x 10-2 + 2.5 x 10-5 9
= 1.4x10 nuclei 6-26
4 SAR 9/84 e
.This corresponds to'an' activity concentration of 8.8 x 10 -7 pCi/cc. '
'The. -gammac dose' rate- from nitrogen-16 of this-concentration in'the air is
- 3. 7 ac 10' pho tons x 8.8 x 10 -7 pCi 3 x 1000 cm
'D =. see-uCi cm 5
2x 1. 6 'x 10 (photons /sec-cm2/ rad-hr)
~
-4 rad /hr = 100 prad/hr-
= ~ 1.0 x 10 ,
when. the effective radius of. the- -room, taken to be a hemisphere with a volume of 4120 cubic meters is 10 m. LThe thickness of the layer of nitrogen-16 bearing water is 3 x 7,4
* . x 0 h= = 1.17 cm , .(24) s 5.05 x 10
.where the volume flow rate 8.0 x 10 3 cm3 /see was given in
^
the. discussion-on heat transfer. The dose rate at the pool surface arising from- the nitrogen-16 near the surface is D= [1 .E2(ph)] , (25) 2pK where p = attenuation coefficient for 6 MeV photons in1 water (0.0275 cm g). 5 K= 1.6 x 10 . photons /cm2 -sec per rad /hr flux to dose rate conversion factor, E2 = second exponential integral. This yields, approximately, D = 400 mr/hr . This value is. larger than those extrapolated from' measurements made on other TRIGA reactors. Transport times from the reactor core to the pool surface in excess of those estimated will lower the calculated dose substantially.- A
- delay time twice as. long as 38 sec. will generate a
- calculated dose rate twenty-five times less.
6.5.2 Activation of Air in the Experimental Facilities-In the TRIGA reactor installation, the following experimental facilities.contain air: rotary specimen rack, pneumatic transfer tube, and neutron beam tubes. Of the 6-27
SAR 9/84
- radioisotopes produced in these air cavities, Jargon-41 (with half-life of 110 min.).is the most significant with respect to ' airborne ~ radioactivity hazards. Nitrogen 16 (7.11 sec.
half-life) and oxygen-19_ (26.9 sec. _ half-life) are considerably less significant. Estimated. releases of argon-41 fr9A reactor operations indicate an upper limit for the release exposure as 190 urad/hr. . -Actual values are expected to be less than 1/50 of the estimated value. The; saturated activity of. argon-41 in an experimental cavity is_ calculated from 41 g A=N A41 , A C1/cm 3 , (26) OI C(A +q/V) 0 where~C = '3.7 x 10 disintegrations /sec per pCi, S= $I, n/cm -sec. I, = 1.59 x 10 cm -I
~
The effective air volumes of several experimental facilities are listed in Table 6-2. -Also given are conservatively high estimatesof the average thermal neutron fluxes-for 1000 kW operation. Table 6-2 VOLUMES AND THERMAL FLUXES OF FACILITIES Effective Average Thermal Air Volume Flux at 1000 kW 1 Region (cm ) (n/cm2 -sec) x 10 Central thimble 7
- 5. 3' x 10 3 21.6 Rotary specimen rack 3.3 x 10 0 6.5 C Pneumatic tube 1.6 x 10 3 6.5 -
1 Tangential beam ports 3.0 x 10 5 0.1 Radial beam ports 1.8 x 10 5 0.1 For volume exhaust rates where the decay constant is negligible, such that A41 << q/V , (27)
. the activity released from each volume is given by 6-28 m
P SAR 9/84 41 A q ,
)41 ( E,)1 V 1/C pCi/sec . (28)
With a. flow rate of 4. 7 5. x 10 3 3 cm /see this condition- is achieved for a volume of approximately 0.45 cubic meters. Total volume of; air cavities . without any experiments in place is about! 0.5 cubic- . meters. The total . activity calculated.for the air leaving the experimental facilities is therefore, 176 pCi/sec. - It. should -be emphasized that .the air activation and
- subse.quent- release activity are predicted for vacant beam.
ports and conservative neutron fluxes. Actual release rates depend on the particular . configuration of experiments and the air exchange rate in.each facility. The release of argon-41 from the facility is diluted by the ventilation exhaust rate, assumed to' represent two air changes per hour, and averaged for a 5 day, 8 hour operation
. schedule at full power. The release concentration from the pool averaged for one year is,
.24 (1.65 x 10-6) = 3.96 x 10 -7 uC1/cm 3 ,
Only 20% of the experiment facility argon-41 is assumed
- to exhause since experiments will replace some or most of-the exposed air. '
176 (.20) (.24)'/2.29 x 106 , 3,7 x 6 Ci/cm 3 . it Total estimated release is 4.1 x 10 -0 pCi/cm .
.The whole' body gamma ray dose rate to a person immersed in a semi-infinite cloud of radioactive gases can be approximated by D= 900 EA D (29) where E = the photon energy, 1.3 Mev A = effective exposure concentration, Ci/m .
D The concentration downwind from the point at which .the
- activity is discharged from the building is A = Aq $(x), (30)
D where $ = the. dilution factor at the distance x, ( s e c / m.3 ) , A. = activity concentration in the discharge (C1/m3), q = 'the building exhaust rates (m 3 /sec). 6-29
c,- , e '
.g.
SAR-9/84' . if it 'is . assumed -that . the discharge is . a' t ' the roof line .the-dilution factor iniche lee.of the building -(x=0), . is given [6] by:
$(0) =
1/csu , (31) a- where c =.a constant-(0,-5), s = building cross-sectional area normal to the. wind direction- (m2), u,= wind veldeity (m/sec). A minimum cross-sectional area is aasumed of 234 m2 (60 x 42 ft) and, for a-wind velocity of'1 m/sec,
-3
$(0) =
1/(0.5 x.1. x 234) = 8.5 x 10 .sec/m 3 . .(32) The averaged dose rate.at'the exhaust stack is
-3 rads /hr, D = 900 x -1.3 (4.1 x 10-0) = 4.8 x 10 an' average of 4.8 mrad /hr in the stack or D = 4.8 x 10-2 (8.5 x 10-3) = 4.1,.x 10
-5 rads /hr, an average of 41 prad/hr at ground level.
;, , . Actual dose values for argon-41 release will- be substantially lower. It is - expected that release values will be less than 4.8 prad/hr which- is equivalent .to 10 mrad /yr.' Lower neutron fluxes, smaller air volumes, shorter '
operation times and-larger dilution factors will assure that releases are below the estimated values. P k 6-30 y t
- - - - - _ _ _ - _ , _ _ . - - _ _ - - _ - - - - . - - _ . - . - - - . - - - . - _ _ _ _ _ - . - - - _ ~ - . - . - _ _ - - _ _ _ _ _ . . _ _ - _ - . - _ . _ _ . _ - - - _ _ _ - _ . _ _ . _ . _
SAR 9/84
. . Chapter 6_ References
- 1. Code of Federal Regulations, Chapter 10 part 2 0,- U.S.
Government Printing Office, 1982.
- 2. .Dorsey, N.E., Properties of Ordinary Water-Substance, Reinhold Publishing Corp., New York p. 537.
- 3. Ibid., p. 554..
- 4. Mitti, R.L., and M.H. Theys, "N-16 Concentration in y- EBWR," Nucleonics p. 81 (1961).
- 5. Dorsey, N.E., op cit., p. 554.
.6. Slade, D.H., (ed.), " Meteorology and Atomic Energy,"
USAEC-Reactor Develop, and Tech. Div. Report TID-24190, DFSTI, Springfield, Virginia, 1968. O 6-31
jr-SAR 9/84 e Chapter 7 SAFETY ANALYSIS In this section an analysis of abnormal operating conditions will be made with conclusfans concerning the effects on safety to the reactor, the public, and the operations personnel, as a consequence of any abnormal operations. The abnormal conditions that will be analyzed.are.
- a. Reactivity accident
- b. Loss of' reactor coolant
- c. Fission. product release from clad rupture 7.1 REACTIVITY ACCIDENT c7.1.1 Summary Rapid insertion of reactivity into a TRIGA reactor is a designed f eature of the fuel performance [1]. Thus, most plausible reactivity accidents do not subject the fuel to conditions more severe than normal operating situations.
~ Postulated accidents for other undetermined scenarios also are predicted not to exceed fuel element safety conditions.
The standard TRIGA fuel element of U-ZrH (H/Zr; 1.6) is . composed of a stable gamma phase.ZrH that does not undergo a phase transition-at temperatures less than about 1250*C [2]. Pulsing limits for fuel elements clad in stainless steel are set by the hydrogen equilibrium pressure within the fuel element. This pressure is a function of temperature and must not . exceed the rupture stress of the fuel element cladding. For the stainless steel cladding (0.02 inch thick), the rupture pressure has been measured' to be 1800 psi ati100*C. The fuel temperature at which the equilibrium hydrogen pressure will be 1800 psi is about 1150*C. The average and peak - f uel .cemperatures at 1.5 Mw steady-state operation are about 220*C and 400*C, occurring well below the limit. The average and peak fuel temperatures occurring after a 2.8% reactivity insertion at respective power icvels of 1 kW and 880 kW are also less than 1150*C. Values of 376*C and 843*C are predicted for a low power insertion and values of 460*C and 795'C are calculated for a high power insertion. 7-1
- .. , . ~ . . - -- . . - . . . - - .. . -
SAR 9/84 1 I Twof reactivity- accident scenarios' are presented. 'The first-is.the-insertion ofJ2.8%freactivity at .zero . power by-theDaudden removal of the. maximum worth ~- control rod. The
. second' eis thei sudden removal ' of' the same . :2'.8% reactivity.
~with the . reactor operating ' at- a _ power. Icvel equivalent to-the ibalance L o f--'the . core excess reactivity. ~ It ' is t unlikely
~
1 that . ;novenwnt o f reacto r .. f uel or experiments'would. lead to
; thel po s tu),a t ed; ac c id ent s . - Movements of control rods for the ifirst case are controlled administrative 1y while ' movements '
Joficontrol.--rods for:the-second case ~are prevented by control'
. c i r c u i t ._ design. .Provided ~ the' total ' worth of- reactor-experiments ~ are limited to $3.00 no experiment movement- ;
'could. generate the postulated = accident.
. .. Pulse ' powers jpredicted from k'inectics formulations.
based on the Fuchs-Nordheim-Scalletar model are displayed.in s Figure ~7-1. Pulse shape, energy and temperature for $3 and
-$4fpul'se insertions are shown.
.7.1.2 Analysis of 2.8% insertion at' 1 kW.
~A rapid insertion of excess reactivity in the reactor
-system, is postulated. The method of inserting- this reactivityf is through the rapid removal of. a control rod. ,
- This . reactivity insertion -is the most serious that -could
. occur. ' It .is also the normal pulsing - condition and .the analysis is presented here as a. point of'information since it is'not actually'an accident' condition.
The sequence of events leading to- the postulated .i
~
= reactivity; accident is:
~
-a. The reactor is:just critical at some low power level (less than~1 kW).
.b. Upward force is. applied to a high worth contro'l-rod
-causing-it to be~ ej ected from the core and to introduce an excess reactivity of $4.00. !
~
The consequences ~ of--the above sequence of events are: a.. Reactor power is increased to a maximum power of approximately 4220 MW.
- b. A maximum energy release of approximately 36 MW-sec is reached when the maximum fuel temperature of 4 843*C is reached.
- c. Stresses are predicted in the stainless steel cladding of.approximately 2940 psi. These pressures are caused by expansion of the air and fission product gases and the hydrogen release from the fuel material. Neither of the preceding stress values will cause cladding rupture.
7-2 7 lhei . ___.._._,_b___o.___________..m__u. m _______m_.m._[_____. _____..m______U.___ _ _ _ . . - _ _ _ _ _ _ . _ . _ _ . . _ . _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ . _ _
I SAR 9/84 J 10^4 ..
'10^3._
[- i A Teraperature
. *C 10^2 _
} Energy N-sec 10^1-..
- Power
- - m-
~
10^0 . . . . . 1J .......i....'.... 0 .04 .00 .12 .16 .2
$4 Pulse-Initial Power 1h 10^4__
10^3__ _ A Tmperature
. *C
- 10^2__
- Energy
~
Re sec 10^1 _ Power
=
w
~
10^0 . . .t. .................. 0 .04 .08 .12 16 .2
$4 Pulse Initial Power 880 h CALCULATED PULSE' SHAPE, ENERGY AND TEMPERATURE Figure 7-1 7-3
g-i^ 1e x SAR 9/84-
~
- 9. ; . 1The; analysis; .of tihis accident- is . conservative in, a numbe r' '._o f (way s ', some of. which have ~been indicated in the reactor . design- : bases) (Chapter. 3). . -For example, the Le qu ilib rium' _ L p r e s s u re of hydrogen- .over the fuel is not c , .achievedEduringya' pulse or. step insertion of reactivity.
,I t K w a s assumed.that.the reactor.is just critical at'a
~~ -
-.lowfpower level'with a fuel and. coolant temperature of 25*C.
Additional input parameters are summarized in' Table 7.1.
. T ab l e . 7 - 1.
REACTIVITY. TRANSIENT. INPUT PARAMETERS ; ReactivityLinsertion, $ 4.0 Temperature coefficient, -4 prompt (6k/k)/*C -1.1 x 10 Delayed; neutron fraction 8, % 0.70 Neutron lifetime 1, usec. 4'1 Heaticspacity'Cp, watt-sec/ element 817 + 1.6 T g ,1 The' computations -leading to these conclusions- are determined by the following lumped parameter analysis. The '
~
Fuchs-Nordheim model for reactor dynamics yields the coupled setrof differential equations: 1 6 P / 6 t. = (6k - a T)-P (1) C 6T/6t = P-P O- ' ( with. C=CO+CT g (3) where i = Prompt neutron lifetime, sec.
. P = Power-. level, - (P initial power), watts.
0 6k = Reactivity above. prompt critical, 6k/k - 8. a = Magnitude:of the ~g negative temperature coefficient, *C T = Temperature (avg. over fuel) above the equilibrium temperature at P O, C. C = Heat capacity of the fuel-in the core, w-sec/*C. 7-4 : e - -
v l , SAR 9/841 C 0 =7 Heat capacity at th'e equilibrium temperature
' corresponding'to PO, w-sec/?C.
.C g
= Rate of.changefof= heat capacity with temperature, w-sec/*C2, 1The (above lumped parameter syatem neglects hent'
. t rans f e'r : and Ldelayed neutron effects and' averages space and ~
neutron' energy _ variations so _that all coefficients are
.assumedEconstant.f . Combining equations
-dP = (6k-aT) (C0+ 1}- (4) dTI =1.
(P-P0}' Integrating, using the condition that T = 0 when P =P O'
-yields-in(P/P0}
~~
- 1 ((P-PO) 0
~T (6kC 0_- (aC 0 -
1
) ! ~"
1 T2/3)} (5) Maximum (or -minimum) temperatures occur when- the pulse initiates and after culmination of the pulse such that
'P'-
PO=0 ,
. P-= P '~
O and then, TE(6kCO --(aC0-C g 6k)T/2'- aC gT2/3)} = 0_ . (6) The roots of this equation are
.T,~= -3/8-(o - 1) . -( l ' + [1 + 16/3 o/(o -1)2)1/2} Tg(7)-
where1a = aC0 /6 C g ,- Tg = 2=6k/a ,- n and the positive sign is taken.for the square root if a < 1. . -From the-fuel heat capacity the core heat capacity with , 90~ elements is C0 = 817-+ 1.6 T, w-sec/*C-element (8)
= 857 w-sec/*C-element x 90 elements / core 7.71 x 10 4 y_,,ef.C
~
= .
For the insertion of $4.00 of reactivity a value of 6k equal to 0.021 = ($4.00 - $1.00) (0.7%/$).
.7-5
, - - s SAR 9/84
~
o= .1 x 10
, _ .(857/1.6)- =.;2.81:
9) 2.1 x 10
!and:Tf=2 (2.'1 x-10-2)/1.1 x=10-4 = 382*C . (10)
Thus; T, - -3/8 (2.81: - 1)
.{1-- [1 + _16/3 2.8'1/(2.81 -1)2j l/2) T f (11).
= 0.92 T =-351*C, .
f Therefore, at the conclusion of the pulse the average fuel-temperature.will be T;, = 352.+ 25 = 376*C . (12) To determine the maximum temperature in the' hottest-
~'
fuel. element, the; average _ energy release is determined and~ then 1 multiplied by- the peak power ratio to 'obtain the maximum energy release in the center element. The ; peak power ratio _ includes- radial, . axial and element peaking factors. Then one returns to the energy versus temperature
^ equation to determine the maximum' temperature.
Let-E equal the energy necessary - to raise- the av'erage core temperature from the temperature at the initial . power: ~
" level'to the temperature at the final' power. level. Then
, T, , T,
.E= C dT =
[ CO # 1 T ]dT -( 13) - 0- - O s E=C O T, + Cy T, /2 (14 E = 3.57 x 10 7 wat't-sec For_a peak-to-average power ratio of 2.2 and.an element
- p'eaking factor of 1.4 the energy release of the element producing the peak power is 3.1E or 110 Mw-sec.
A peak
. temperature is calculated by substituting this energy into the. previous-equation and solving for the temperature.
T p
= -C 0 / C'1 + ((C O/C y) + 2(3.1E/C g)]1/2 (15)
. T p
= 818'C 1 with
-T,, = 818 + 25 = 843*C .
~
_During the time of peak fuel temperature, the stress on the. clad from the pressure produced by the expansion of air 7-6
* ~ -TL'
P
'SAR 9/84
.andL. fission product. gases'and the hydroger released from-the
~
' fuel is ~ less than .the strength of - the clad material and therefore there is no loss of clad-integrity.
.The. partial pressure exerted by gases _is Pg ='N RT -(16)
V.
~
where initiallyf the volume. V, is taken as a 1/8-in. space between the-fuel and . reflector end piece. This result is
. conservative since.the, porosity of the graphite reflector'of 20%.isineglected.
'The volume then is 2 3 V = wr h = x(1.80)2 0.317 = 3.23 cm ,
(17) The. partial pressur'e of the air in the element is P = x 10 = 4.46 x 10 -5 RT . (18) Calculation of the fission product -- gases .in a fuel Lelement,is determined by burnup. For an element operated at three times the ' 4.5 : MW-days' discussed in Section 7. 2, a total of 0.016 - moles of stable and radioactive gases are produced. If the release fraction is taken as .0015% as
~
discussed in Section 7.3, then fp = 3.22.x 10 -7 (1.5 x 10-5) = 2.4 x 10 mole s . (19'). 6.02. x 10 From'this, one obtains,
-8 P
gp = 7.45 x 10 RT . (20) The. total pressure exerted by' the air and fission products is P g = (4.46 + 0.007).x 10-5.RT.= 4.47 x 10 -5 RT (21) P g = P,gj .. Also-we have P dr = 14.7 1 psi . (22) 237
'As an upper' limit, assuming an air temperature equal to <
'the peak fuel temperature of 843*C or 1116*K , one obtains
-P = (14.7) 1116 = 60.1 psi . (23) 1 273 7-7
SAR 9/84 The equilibrium hydrogen pressure over ZrH (H/Zr; 1.6) at 843*C.is120; psi. The total internal pressure then is P l= P h
+ P g = 80' psi . (24)
Assuming'no expansion of the clad, the stress produced in the' clad by this pressure is 3, rP 0.735 P g = 36.75 P g (25) t = t 0.020
=
(36.75)-(80) = 2940 psi. For a reactiv'ity insertion of $4.00, the~ clad surface temperature would be approximately equal to the saturation temperature of the water which is 113*C at a- pressure of 23.4 psia. At - this temperature, the ultimate tensile
- strength .for type 304 stainless steel is greater than-60,000 ipsi with a yield stress of approximately 36,000 psi.
Cemparing this strength with the stress applied to the cladding during the reactivity insertion, it is seen that the strength. of the material far exceeds the stress which . would-be. produced. Therefore there would be no loss of clad integrity or damage to the fuel as a result of the reactivity accident. 7.1.3 Analysis of 2.8% insertion at 880 kW. The reactivity accident considered here would take place in.the following manner. Initially, the reactor is cold clean with all control rods inserted. The reactor is loaded'with 4.9% 6k/k excess reactivity and the pool coolant is at a temperature of 42*C. This accident requires someone deliberately violating the operating license and several interlocks and scrams. The sequence of events leading to the postulated reactivity accident is:
- a. The operator slowly withdraws all the control rods except the maximum worth rod, until all the rods are completely out and the reactor is operating at a high steady state power.
- b. Upward force is applied to the maximum worth rad ejecting it (by some means impossible to conceive) from the hot operating reactor.
The consequences of the above sequence of events are:
- a. Reactor power and fuel temperatures are , increased by the_ compensated reactivity of $3.00 (that is 4.9% -
2.8% = 2.1% = $3.00) to levels of 880 kW with fuel temperatures of 380*C, peak, and 207'C, average. 7-8
SAR-9/84 b'. A prompt ' insertion'of-2.8% 6k/k results in an average temperature _in the core of 460*C and a. peak' temperature of--795'C. ut . Stresses'are' predicted in the clad of about 2100 psi. Even'if the clad were at the maximum fuel temperature this stress is a factor of ten below the ultimate strength of the clad. The analysis of this accident is conservative as described in the previous accident case. Equilibrium element conditions and pressure are not expected although calculations include finite reactivity insertion time.
~ delayed neutrons and heat transfer.
It 'i.s ' assumed that the reactor power level is 880 kW ($3.00 of power coefficient); the average fuel temperature is ~(165 +-42)*C; the peak fuel temperature is (338 + 42)*C; ~ with arpool coolant temperature of 42*C assumed. Values of the input-parameters are summarized in Table 7-2. Table 7-2 REACTIVITY TRANSIENT INPUT PARAMETERS.
-Reactivity insertion, $ 4.0
' Temperature coefficient,
~
-4 prompt (6k/k)/*C -1.1 x 10 delayed neutron fraction 8, % 0.70 I ' neutron lifetime 1, usec 41 Heat capacity, watt-sec/ element fuel C p, at 0*C 817 fuel C p, temperature dependence 1.6 T fuel water.C p, at 25'C 879 Thermal resistance, 'C/MW:
0 fuel to cooling channel 5.29 x 10 cooling to pool 1.42 x 10 3 7-9
s-SAR 9/84 A' computer program .was used to calculate the energy release'in the transient.. The program is a'one-dimensional combined reactor kinetics heat transfer program that . works extremely well for reactor transients in which detailed heat transfer analysis is - not required. Delayed neutrons and finite reactivity insertion time are included in the program. Using the parameters given above it was found that the addition of $4.00 reactivity (2.8% 6k/k) from an average fuel temperature of 207'C the . average fuel temperature at 880kw) produced an energy release of 32 Mw-sec in the 90 element core. The energy density at the axial midplane of the maximum power density element, E,, is: E = 1.1 P P, E/n where P = relative gover in element, P, = axial peaking factor, S-E = energy release in the-transient, n = number of elements,. 1 and the' factor of 1.i13acounts for uncertainties. In chapter 3 the radial power distribution within a fuel element is shown. The energy depcsited at radius r per unit volume is: E . = f(r) E, where f(r) is taken from data in chapter 3. The after pulse temperature at the radial distance r is given by: T p(r) =
-CO/C g + ((C O/C g) + 2Er/C y}
where -4 C 0
= 8.17x10 Mw-sec/'C and C y
= 1'6x10 -6 Mw-sec/(*C)2 ,
Calculation of the final element fuel temperature is accomplished by adding the -temperature of the pulse deposited energy to the fuel element temperature prior to the pulse. The radial temperature distribution in the fuel prior
. to the initiation of the transient is given by:
T(r)-T, = (T,-T,) [1 - r'- r'in(1/r')]/r,' 7-10
7 S Alt 9/84 I, where.T(r) = temperature at radial distance from center r,
.T, = temperature at fuel surface at axial center, *C T, ' = maximum temperature ^ in element. *C r,r, = radial position and radius of fuel, em, ^#
r' = (r/r,)2 r,' = [1-r"-r"In(1/r")] r" = (r y /r,) and r g
= inside fuel radius, cm.
In Figure 7-2 there is shown the before pulse and after pulse temperature in the axial misplane of the maximum power density element. As can be seen the maximum temperature occurs at the periphery of the fuel. The adiabatic value is 795'C. In Chapter 3 a plot - is shown of pulse temperature
' distribution as a function of time. This figure shows the typical dependence of temperature as heat flows quickly toward the fuel center and toward the clad.
For a fuel temperature 'of 800*C the equilibrium hydrogen pressure .over the fuel would be less than 15 psi and the pressure exerted by air within the element would be less than 60 psi even if it were at the maximum fuel temperature. The stress imposed on the clad by 75 psi would be about 2800 psi. The ultimate strength of the clad. is over'20,000 psi at 800*C. Therefore one can conclude that the clad integrity would not be compromised as a consequence-of either of these events. A similar analysis was made.in which the reactor was assumed to be operating at 1.5 Mw. In this case only $2.79 (1.95% 6k/k) was available to be inserted. The peak before pulse temperature was 475*C and the reactivity insertion
. resulted in an energy release of almost 30 Mw-sec. The peak after pulse adiabatic temperature was 823*C which occured at the inner fuel radius because of the' high initial temperature at that point.
7.2 LOSS OF REACTOR COOLANT 7.2.~1 Summary The reactor will operate at a calculated maximum power density of 18 kW/ element when the reactor power is 1000 kW and there is 90 elements in the core, all of which are standard TRIGA fuel. If the coolant is lost immediately after reactor shutdown, the fuel temperature, indicated in 7-11
. i
44 4 -, ,. SAR 9/84 i , 800 - 700 . h 600 . v500- - O 1 Q 5400 E z La "300 . 200 - 100 - E A 1.0 2.0 RADIAL DISTANCE - CM FUEL TEMPERATURE DISTRIBUTION BEFORE AND AFTER PULSE FIGURE 7-2 7-12
r".; . ,
~ ,,
SAR 9/84 4 _
-c Figure 7-3, will' rise to a maximum . value of 750*C. The stress imposed;on the fuel element: clad by the internal gas i pressure,7 presented.in; Figure 7-4 is about 2300~ psi when.the fuelfand. clad' temperature is 750*C and the yield stress for-the~ clad: :is about 19,500 . psi.- Therefore, it .can- bc ,
concluded thatfthe postulated loss-of-coolant accident ..will
~
- not . result in any' damage to the ~ f uel, . will . not result :. in release of fissionLproducts to the environment, and will not-
. require emergency cooling.
If the reactor tank is - drained Mof ' water,- the fission productL decay heat- will be removed through the natural-convective flow ,of air up through the reactor core. If the-idecay-heat production is sufficiently low because of a low fission' product inventory or a long. interval between reactor
-shutdown and coolant loss, the flow of air will be enough toi maintain the -fuel at a temperature at- which. the fuel
. elements are undamaged. The following analysis shows that: ;
- a. The maximum temperature to which the fuel can increase is 900*C without substantial yielding of the clad orLaubsequent release-of fission products, b.: This temperature will never bevexceede'd under any.
conditions-of coolant' loss if:the maximum operating-power density is 22 kW/ element or lesa,
~'
- c. For maximum operating power densities greater than' 22 kW/ element, emergency cooling can befprovided:to ensure that the fuel element temperature does not.
exceed 900*C. The required emergency cooling time as a function'of maximum operating power: density is ,
-shown in' Figure 7-5.
7.2.2 Fuel Temperature and Clad Integrity a
. The strength of the fuel element clad is a function of its temperature. The stress- imposed on the clad is' -a function of the fuel temperature- as well as the
-hydrogen-to-zirconium ratio, the ' fuel burnup, and the free 1 gas volume within the element. In the analysis of' the Tstresses imposed on the clad and strength of the clad the following assumptions will be made:
- a. The fuel and clad are at the same temperature. I
_b. The hydrogen-to-zirconium ratio is 1.65.
- c. The free volume within the element is represented by a space 1/8-in. high within the clad.
d.-The reactor contains fuel that has experienced burnup equivalent to'only about 4.5 MW-days.
- i 7-13
~ - -
FT SAR 9/84 2000 - 1800 - COOLING TIME (SEC 1600 - lemmamens rest temperature versue p.wer desetty 1400 -
. ter t .c e t t r., vert e . tt times between react.e caustd we and coetaat tese 3 1200 -
t a 1000 - U 800 - g 1 600 - 10 400 -
- 200 -
. . e i 35 40 45 O $0 5 30 o S to 15 OPERATING POWER DENSITY-KW/ ELEMENT EL-1872 FUEL TEMPERATURE AND POWER DENSITY FOR ELEMENT COOLING TIMES Figure 7-3 7-14 ww--v3-ee---'---e+Wmerto
r . SAR 9/84 5 10 . ULTIMATE STRENGTH _,*%g N N YlELD STRENGTH
\
0
-10 -
\
N e N
- t. \
N N N
= \
\
N STRESS IMPOSED ON CLAD l 3 - 10 - Strength and applied stress as a funetton of temperature, 0.Zrst .4 5 fuel, fuel and clad at same temperature 2 , , 10 , , , , , , , 400 600 500 1000 I200 TEMPERATURE ('C) EL-1873 U-ZRH(1.6) STRENGTH AND STRESS VERSUS TEMPERATURE Figure 7-4 7-15
SAR 9/84 1 105 . T,= 50 60 70 00 90 100 S T S 5 '2 2 E E 1ok . 5 E e N O 5 ac g Cooling thnee after reactor shutdown w , g necessary to limit maainum fuel tempera. W ture voraus power density J E 30 3 E ' e-E 3 8 u T"= MAX. FUEL TEMPERATURE AFTER WATER LOSS (*C) 10 , . > ' ' ' 0- 10 20 30 40 50 60 70 POWER DENSITY-KW/ ELEMENT EL-0706A COOLING TIMES AFTER REACTOR SHUTDOWN TO LIMIT MAXIMUM FUEL TEMPERATURE VERSUS POWER DENSITY Figure 7-5 7-16
7.- . .
^
c,.
- SAR.9/84-v
.W ( 'I '
Th'e fuel element' internal pressure p is1given by-P=Ph+Pgp,P air
- where' Ph- - hyd rogen .; p re s sure ,
Pg g = pressure exerted by-volatile fission products, P,g = pressure exerted.by trapped air. For hydrogen-to-zirconium ratios greater.than about 1.58 the equilibrium hydrogen prescure can be, approximated by P h' exP [1.767 + 10.3014x - 19740.37/(TK)] .(26) (atmospheres), where:x1= ratio of-hydrogen. atoms to zirconium-atoms, and T "
. fuel t em P e'r a t u r e . (
- Kl '.'
K: This expression was d'c rlve d from least-square fits to.the data of Dee and. Sianad. [3]. For ZrH (H/Zr; 1.65) the hydrogen pressure.becomes 8-P h= 1.410 x 10 exp [-19740.37/(TK)] (atm spheres). The pressure exerted by the fission product gases is
;given by Pg = fn RT K ',
EV whereif = fission product release fraction, n/E = number of moles of-gas evolved per unit of: energy produced, moles /MW-day, R = gas constant, 8.206 x'10-2 ' liters-atmospheres / mole *K, V~= free volume occupied by the gases, liters, and E= total energy produced in the element, MW-day. The. fission product release- fraction-[4] is given by f= .f n dn - (28)
'n f, = {1.5 x 10 -5 + 3.6-x 10 .
3 exp [- 1.34 x 104 /T }} where.-T = fuel temperature in the differential volume of the element during normal operation, *K, 7-17
- SAR,9/84
} ,
7 - ff =/ differential' release fraction and n = fuel volume-normalized to 1. '
.e 1Th e . ..f is s ion _ product ~ gas _ production r a t e) . n / E is not independent- .o f 4 power'. density ~ _( neutron flux) but varies slightly ~ with -the - power . density... . Th e . v a l'u e n'/ E = 0.00119 moles /MW-day' is accurate to :within. a lf ew . percent ' over the
~
_.y -
-range' -from,1 a f ewi kilowatts per element to well, over 40 kW/ element. :The free volume. : occupied by- the gases is
- assumed 6to be a space:1/8-in._(0.3175-cm)-high at the top of-lthe_ fuel'so that 2
'V.=-.0.~3175'wrj
.. , ( .2 9 )
where'rjc= inside radius'of the clad'(1.822 cm). For standard.TRIGA-fuel the maximum burnup is about 4.5-MW-days / element. Pressure exerted by fission product gases
~
11s_ ino t ~ significant . .
. . The l air trapped within the fuel element clad would exert a pressure P
g = RTg/22.4 , (30)
- where-it is' assumed that the initial specific volume of the air (22.4 -liters / moles) . is present at the time.of the loss of ' coolant. - Actually, the air - f orms - oxides ~ and nitrides twith the' zirconium'so that'after relatively short. operation the-air is no . longer present in-the= free volume inside the fuel element clad.
For .ZrH (H/Zr; 1.6) fuel' burned up to 4.5 MW-days / element, .with a' maximum ; operating temperature _ of
'600* C , . the internal: pressure as a function of maximum fuel-temperature T is K
P= 8 ~3 -T 1 . 4 1'O x 11' 0 exp (-19740.37/TK) + 3.66 x 10 g (31)
-( a tmo s ph e re )-
sor P = 2.073~x 10 9 exp' (19 7'40. 3 7 /TK).+ 5.38 x 10
~
T (P81)- K The stress imposed on the clad by the gases within the free volume inside the.. clad is
.S= (rc/t)'p. . , (32) where.r c.= clad outside radius (1.873 cm),
t = clad thickness (0.051-em). 7-18
3
' ~
f h( o V SAR 9/84
'j .c Y-o
't Q / 9 y s
s _ a jf 4 - v - : y s.g ,
- lf - -;thef previous :and, initial' ' equation 'are stress etnJbe rewritten as-combined ;t.he
.N ,
4 , .. L
'Y "
'Sc= 36;75 p_
/ ' / (33). _
~
= 7.'614x'10.10'exp.(-19740.37/TK ) + .1.97 T K (Psi).
IntFigure 7-4 this ~ imposed' Etress 'is _ plotted l as a function of maximum fuel temperatures. Also plotted are the yield and ultimate strength of the t,y p e ; 3 04 stainless. steel clad.
=The clad ul timate" s t rength 'is not exceeded :if the maximum.
a if uelf temperature is' - maintained - below- about. 950
- C - -and 'thec
, yield : strength is not- exc'eeded f or any f uel ' temperatures { ',.
below f ab'out 920*C, _ s. lightly below ;the yield point.and.well c
, .below'the rupture point.-
f, %
, 4
.) , , *
, j'
.c 7.2.3 After-Heat Removal Following Coolant Losi sY . .
It isl assumed"that 2 the3 reactor operates continuosly.at " _' a 'constanti, power-. density ' level: P e s o" that5-she--maximum 0 ? inventory of fission products is s'/En il'a b le . t o produce heat-4 after .the reactor .is ..s hu t d own '. . ' The - ~ powe r( dins ity - a f te r . ' reactor. shutdown'P islgiven by , _ > t, ; s v Pf= 0.1=Po-[(t:+ 10)
- 0.87 (t + 2'x-107)-0.2) u. ,
s t'
*-(1.3'cos('2[45 (0.*26x - 0.5)]),,
,7 .
, 'where -P
=.operat ng'pbder density, W/cm ,
--,. t'
?t = ' time-after~ reactor shutdown. sec , .
. .i
. x: .= d'istance:from.the~ bottom'oit -the f u e l- . ~r e g io n ,1 c m .
L Atitheitime that-the coola'nt is lost'.'from the core.the
~ fuel and:its surroundings are assumed to.be'at a temperature t of 27'C. This: is'not.necessarily. true, for an accident; cani be-postulated in' which the: coolant 3 loss is the mechanismiby
~
- - 'which the- ' reactor; is~ ' shut- down. ' (For the standard non'-g ap p e d ;f u e l . e,lem'e n t , under nordal conditions, the time f
t to cool down-fr'ca ing temperatures is a matter o f ~ o r.e . i Pto two1 minutes.):goperat
- Although such ' anLe accident doch not appear 9
4 to ' b'e ; conceivable ,- calculations -indicate that: -if ,i t ' is
. assumed: that. lthe~-average fuel temperature a t, t h e - ' t icae o f-coolant los s - is c equ'ival4nt to the ope rating . . ave rage ~ fuel-r , ;;. temperature, _ thef maximum .;tempe rature - af te r th.e. coolant loss
~
?is inot : appreciably - dif f erent (2% - 4 % ' ' h'i gh e r ); f rom that gg J calculated ' assuming 27'C fuel initially.
t , iThe after-heat removal will be accomplished.by the flow of Eair ~ through ',the ' core. - To determine the flow through the s core c the buoyant force's were equated to the friction, end, and . acceleration losses in the channel as ,shown in the expression ~ ;# 7-19 c q y . t'{ e , , i_ 3
SAR 9/H4 < n:
; 6pb:= 6p g + 6p, 4 '6pg + 6 p ,- . (35)
Theibuoyant forces are given by L- , 6pb " 00 b~ Pdx = 9 b ~ ~
# .(36) 0b ~ P 'O o f 1 t where.p ' P = the entrance, mean, and exit fluid 0' 1 densities, respectively, L = the'-' effective length of the channel
, (~L = L, +L g + L }' t L,, Lg,Lg= the length of the channel adjacent to the bottom end reflector.. fuel, and top end reflector plus ten channel hydraulic diameters, respectively.
'The friction. losses in the flow channel are given by.
" ( }
6pg.= f Fi 2 D, 2gp gAc where.the summation - is over the lower unheated length, the heated length, and the upper unheated length, and f =~the' friction factor (23.46/R ) [5]', pg D = the hydraulic diameter (0.0601 ft), e= A = the flow area through the core per element
- f. (O.0058 fe 2 ),
8 g = 4.17 x 10 ft/hr2 , The sum of the exic and inlet losses, using appropriate expansion coefficients, is given by 6p, + 6p = (EK)'w , (38) 2goo^$ with EK = [k (A /A ) ] = 1.57 . where k is appropriate expansion or contraction coefficient
'from rekions of area A .
The acceleration losses are given by 6p, = (1/p g 1/p0} (" !8A c ) * ( By substituting the appropriate expression in Equation 35, using the definition of the Reynolds number, and L = 7-20
t f. SAR 9/84 2.40 ft,'L, ='0'.29'ft, L g = 1.25 ft, and L g
= 0.87_ft,-one obtains
-4 y 2 (0.700/p g - 0.149/p0) _x 10 + (40) n
-2 (0.153pg/p0 + 0.665p/p + 0.153p2 /0 1x 10 y .
- (1.25p +,0.889p 1.- 2.139p0) = 0 with ' the flow w in units of Ib/hr and u the viscosity in.
. units of Ib/hr-ft.
The- properties of air for use in Equation 40 are expressed as p g = 40/T 1 (lb/ft3h (41)
-8 2 and u i
= 5.739 x 10-3 + 7.601 x 10-5.T I - 1.278 x 10 T I
(1b/hr-ft),
-where T is the appropriate temperature in 'R.
P The heat transfer coefficient was calculated through the relationship
~
N = 6.3 R, 5 1000 (42)
= 0.806 R a 0.2976 Ra > 1000 ,
where N = the Nusselt number = h9,/k, 4 ' R, = the Rayleigh number D e p'gB6Tcp/ukL I h- = the heat transfer coefficient, Btu /hr-ft2_op, k = the thermal conductivity of the laminar film, Btu /hr-ft 2_ep,
~
the volumetric expansion coefficient, *F ~
=
S , 6T = the temperature rise ov'r e the channel length, L(*F), c p
= the specific heat of air (Btu /lb *F).
The expression for the Nusselt number was derived from l_ ~the work of Sparrow, Loeffler, and Hubbard (6) for laminar '- flow between triangular arrays of_ heated cylinders. The thermal conductivity and specific heat are given by k = 2.377 x 10-4 + 2.995 x.10-5 T- 4.738 x 10 -9 T 2 l (Btu /hr-ft *F) (43) l 7-21
SAR 9/84 and c ' = 2.413 x 10 1.780 x 10 -0 T + 1.018 x 10 -8 T 2 P (Btu /lb *F), (44) where T is the appropriate temperature-in *R. These two expressions, as well as that given for the dynamic viscosity of air in Equation 41, are least-square fits.to the data presented by Etherington [7]. TAC 2D [8], a two-dimensional t ratk s ie n t -h e a t transport computer code developed. by GA Technologies, was used for calculating the system temperatures after'the loss of tank water. :The parameters derived above were programmed into the calculations. The maximum temperatures reached by the fuel are plotted as a function of operating power density in Figure 7-3 for several cooling or delay times between reactor shutdown and loss of coolant from the core. For reactor operation with maximum power density of 18 kW/ element, or less, loss of coolant water immediately upon reactor shutdown would not cause the maximum fuel temperature to exceed 750*C. Operation at maximum power densities greater than 18 kW/ element will not result in fuel temperatures above 750*C, if the coolant loss occurs sometime after shutdown, or if emergency cooling is provided. (The time required between shutdown and the beginning of air cooling depends on power density.)
.In Figure 7-5, the data presented in Figure 7-3 were replotted to show the time required for natural convective water cooling or emergency cooling, after reactor shutdown, to produce temperatures no greater than a given value.
Thus, for example, for a reactor in which the maximum operating power density is 27 kW/ element and to limit the temperature to 950*C, or less, there must be an interval of at least 3730 sec (or 1.04 hr) between reactor shutdown and either the loss of tank water from the core or the cessation of emergency cooling. The 65 minute delay time applies to the power density of a 90 element core operated at 1.5 MW but shrinks to a negligible value for the power density in a 100 element core. 7.2.4 Radiation Levels Even though the possibility of the loss of shielding water is believed to be exceedingly remote, a calculation has been performed to evaluate the radiological hazard associated with this type of accident (see Table 7-3). Assuming that the reactor has been operating for 10 hours and 1000 hours at 1.5 MW prior to losing all of the shielding water, the radiation dose rates at two different locations are listed below. The first location (direct 7-22
m 4 SAR 9/84 E. radiation) is 6.4 meters above the unshicided ' reac tor. core , near'theJtoplof the reactor tank. The second is at the top-
- of theJ reactor..shiel'd; ~ this location- is shielded. from
. direct radiation but is 'subj ect to scattered radiation from a thickfconcrete ceiling 4.6 m above the.. top of the reactor X : shield._) 'The: assumption that .there- is a thick' concrete c'eiling-maximizes the reflected radiation dose. Normal-roof atruetures- .would':give ,c o n s id e rab ly. less backscattering.
1 Time.is measured from-the conclusion of operation at-1.5 MW.
-Dose rates assume _no' water in.the. tank..
. =The tabulated data show that if an individual does not-expose .himself-. directly to the core he could work for approximately 2 hours ~(3 hours for 1 MW) at the top of the-shield tank i day after shutdown without receiving'a dose in excess- of_ that permitted by regulations for a . calendar quarter.-
Table 7-3 Calculated Radiation Dose Rates For Loss of Shield Water Direct Radiation Scattered Radiation
' Decay R/hr R/hr
' Time 10 hour- 1000 hour 10 hour 1000 hour-1 minute 3980 4920 3.7 4.6 1 hour 929 1820 .87 1.7 1 day 87 681 .08 .64 1 week 10 281 .009 .26 1 month 2 104 .002 .10 For persons outside the building, the radiation from the unshielded core would be co111 mated upward by'the shield structure and, therefore, would not give rise to a-public hazard. The method of calculation follows.
The-core, shut down and drained of water, was treated as' a - bare cylindrical source of 1-MeV photons of uniform strength. Its dimensions were taken to be equal to those of-the' active core lattice. The source strength as a function of time was determined from Way and Wigner's [9] (Equation 45)-data on fission product decay. No accounting was made of sources other than fission product decay gammas (i.e., activation gammas from the steel cladding and the aluminum
. grid ' plates) or of' attenuation through the fuel element end ,
. pieces and the upper grid plate. The first of these assumptions is optimistic,'the second conservative; the net 7-23
W
, .,. S AR .- 9 / 8 4 effect is e conservative. The conservative - assumption of a .--
- unif ormly distributed source of 1-MeV- photons was-balanced
-byfnot' assuming anyfbuildup in the' core.
-An approximation.of the fission product energy release.
term is taken as:
-I*
- r ( t)-: = 1. 2 6 t , (45)
.where.
~
T(t) = energy release'in MeV/sec-fission, t =-is the time after fission in seconds. By integration the total-core source term is j S(t,T) = 3.1 x 10 10 p t+T P(T) dT (46) 1.95 x '10 I1
~
= P o (1 - [1 + T/t] ~ * }-t (47) wh e r e :
.S(t,T)'= energy' release.in MeV/sec-watt, P, =--reactor power,-watts, T'= period of time at power.
The volumetric source of 1 MeV photons is S
= S(t,T) (48) wr c *c The direct dose rate at a point outside and on the.nxis of a cylindrical' source is given by:
'x 'T D v-
- p*z 2wrdedx (49) d = li 'O <0 4wR' where 3
S y = source strength in photons, 1 MeV)/cm -sec,
'K =
flux-to-dgsephotons conversion factor,
/cm2 -sec per rad /hr, 5.77 x 10 2nrdrdx =' cylindrical volume element, dV r = core radius, 26 cm c
x = core height, 38 cm c 7-24
N+
~
SAR 9/84
-I
-u c = core attenuation coefficient, 0.207 cm R = distance from volume'elementJto receiver, em
.= slant penetration in core = xR/(a+x), em
~
z-a = distance from top to core to receiver, 640 cm t . . For: distances far from the core (i.e.-for a >> r and Yc) the above expression. reduces to v c (50) D d
=
4p aZK- ( '" * ""c*c} e The scattered dose rate was calculated from D =-6'03 x 10 23 Z I,C Qa (51) s p A K(E)x 2 , where-p = Density gf scattering material, concrete. 2.3 g/cm - Z = Ratio of average atomic number to atomic mass-A of the_ scatterer, 0.5 and I C = Incident current times cross section of beams, photons /sec K. = Photon cugrent to dose2 rate conversion, 2.75 x 10 photons /cm -sec per rad /hr E = Energy of scattered photon, Mev x = Distance-from scattering point to detector, J 400 cm I 0# Qa" - 90 *"1 (cos00 /cose ) 60g UO'"1
= Attenuation coefficient in scatter for incident g
and scattered photons, cm , .146, .292 00,0 1 = Incident and scattered angle (measured from the normal to the scatterer), 0, 25 degrees 60/60 = Differential Klein-Nishina scattering cross section, cm 2/clectron-stcradian It was assumed that all of the source photons that exit the top of the reactor pool were incident normally to the concrete roof ' (i. e . , 0 0
= 0) at a point directly over the core, thus IO C=S W0 (52) 7-25
p - M+ . ~ (
~s
- G .13 s' Tid ..
, s SAR 9/84
. .e-f TwherefS O v c2!"c (53)'
-l ~YO (# 0 ^*0 )*# 0 (# 0 *0 )
_ r-
>W .
'= { sin - .[ ] - w /2 } /2n-(r 0 #
'*0) (# 0 *Y O
)
. and. ~ 0r- =- Distace from'the core to the top of the
.; pool,6.4 m
. e T x = - Half width of-the-pool, 'l m-0 y0; = Half length of the pool, '1.5 m S y,r'c' "c have already beentdefined.
The energy'of the scattered photons is given by E = 0' (54)' 1 +;EO(1 -c so)/0.51
. whe re E i the incidentLphoton energy (1 Mev) and 0 :is the
- s c a t t e r kng s_ angle'= x-(00'
- l}*
N The? differential. scattering cross section is given by .
* (55)
=
2
.[ -
( sin 0)2 ( )3). 2.818 x 10
~'
~here w r,.is the classical electron radius
=
- cm.
7.3 FISSION PRODUCT RELEASE In- the Lanalysis ..o f fission product- releases under accident . conditions ,- it is assumed that a fuci element -in the. region of highest power. density fails. The failure is assumed to' occur in_. air after operation at full power for an s extended period.
~7.3.1 Fission Product Inventorv
,. Table 7-4 gives the inventory of radioactive noble gases and halogens in the TRIGA Mark II after continuous
_ operation at 1.5 MW for four yeara (i.e. 6MW-yr). The estimated inventory is conservative since actual operatio'n
-after 4 years-is' expected to be less than 5% of 4 MW-yrs.
F 7.3.2 Fission Product' Release Fractions
~
The release of fission products from U-ZrH fuel has been studied at some -length. A summary report of these 7-26 e a
q __ .- 4 SAR 9/84
* 'h studies [4] indicates that.the release from the U-ZrH (H/Zr;
- 1.6)' fuel meat at.the - steady-state operating temperatures is principally.through recoil into the fuel-clad gap.
~
At high a . temperatures (above 4 00
- C o r' 500
- C)', - the release mechanism
. is'.through a diffusion' process and is temperature dependent, J/. ], .unlike recoil. -
For the accident considered here, it is assumed that- a fuel' element-in"the, region'of highestipower density fails _in water and that ~t h,e peak fuel temperature ~ in the_ element is
' less than 400*C. At this temperature, the long-term release f raction ' would be less than 0.0015%. . For ' the purpose of this - analysis it . is also assumed that. 100% - o f the noble
- gases and 100% of the' halogens are. released from the highest
.powe r' density fuel element in.which 2.22% of the total _ power
- is-generated. .
, Table 7-4 NOBLE CAS AND HALOGENS IN THE REACTOR l
. Isotope. Quantity (C1)
Br-83 6,120. Br-84m 6,120 Br-84 12,360 Br-85 12,900 Kr-85m 12,900
~
Kr-85 678 Kr-87 32,400 Kr-88 46,200 Kr-89 ~58,500 Kr-90 65,100 Kr-91 44,100 1-131 35,700 Xe-131m 288 '
-I-132 53,100 1-133 86,100 Xe-133m 2,100 Xe-133 86,100
-I-134- 9 6 ,' 6 0 0 1-135 80,400 Xe-135m 24,300 Xe-135 83,100 1-136 77,700 4
Xe-137 75,300 Xe-138 70,200 Xe-139 70,800 Xe-140 48,600 h 7-27
{ 1. SAR 9/84 f n' It ,i s important to note that the . release f raction in. characteristic o f. -the- normal [ accident- ' conditions is operating . temperature ' and not. the temperature .during the accident 1 cond itions . This is because' the fission products
-releasedtas aoresult of a^fueloclad failure.are those that have,. collected in-the fuel-clad gap during normal operation.
Other. assumptions- -concerning estimated accident scenario doses are: 1
'a. Assume an element fails in air such that' all (1007) noble gases and. halogens in.the gap are effectively released.
- b. There is no plate-out of;any released fission 1 products.
- c. After the failure a ventilation rate of 10 air changes per hour occurs with no air filtration.
- d. Doses are calculated assuming exposure to a semi-infinite cloud.
- c. Doses'are calculated for release from the total core.
and.a single fuel. element (90 element core with peak to' average flux of 2.0).
- f. Doses external to the building are calculated.by assuming a minimum building-dilution factor for releases (1.0 m/sec wind velocity with building
. cross section of.234 m2),
- g. Doses were also calculated for personnel in the reactor room ~by dumping rapidly a small fraction'of the total inventory into the' room such that the continuous release is equivalent to a constant concentration.
The net effect of these assumptions is that for the single element accident condition, the fraction of the noble gases
~
released from the building is: f NG
= .0 x 10
~
x 1.0 x 2.22 x 10-2
= 4.44 x 10~7 , (56) ^
and of the halogens is: fH = 2.0 x 10
-5 x 1.0.x 2.22 x 10-2
~7 (57)
= 4.44 x 10 ,
where a conservative release fraction of 0.002% is applied. 7-28
-.g
9"
, - _ 'SAR 9/84 7.3'.3 Downwind' Dose Calculation:
The minimum roof level. dilution factor was' calculated, P assuming- a building cross . sectional- area of- 234' square meters. The. factor is based #on mixing . in the lee of the building when the wind velocity is .1 m/sec. A dilution
. factor of10.00854 seconds per-cubic eter is applied.
The ' calculation of whole body gamma doses and thyroid doses downwind .f rom the point ~of release was accomplished
- through.the'use of the computer code CADOSE [10]. In this.
code the set of differential-equations describing the rate of production of an isotope through the' decay 'of .its
- precursors and the rate of removal through radioactive decay and1 removal by the ventilation system is integrated for each member of the, chain =. . The-release rate q, to the envfronment for the ich isotope at time t g in hours 18 : =
qg(t) =gg Q(t) (1/V)/3600 , (58)
~
'where Q (t)-= the release of~the ich isotope in Ci, 1/V = the building leakage rate in (m /hr)/m ,
c f
= the. filter efficiency for the isotope, gg = 1 - ' c y'.
The-quantity Q g (t) is the amount of the ich isotope.In t .t e discharged air at the time, t. This quantity is given by
.Q g (t)
= f t Qg(0) e
-( i
* !Y)'
(59) where Q (0) = the inventory of the ich isotope as found in Table 7-4, Ai = the decay constant for the ith isotope, and f t= the release fraction to the reactor hall. The concentration downwind at a distance x for the ith isotope is calculated from Q g (t,x) =qg(t - t) $(x) e di' , (60) where t = the transit time from the release point to the dose point, hr,
$(x) = the dilution factor at the distance x, sec/m3, The whole body gamma ray dose rate for the ith isotope.
D yg, at'the distance x and time t is calculated, assuming a semi-infinite cloud, through the expression: 7-29
- s A R '# / n4 -
Dy f(t,xg) =_900 Eg Q (t,x)_ , -(61)-
- n. i.
- where E =-the average. gamma ray energy perTdisintegration, MeV,.ank?the' constant includes the attenuation coefficient for' air as1well as the conversion f actors required. Dose rate is in units of-rad /hr.
V Internal: dose rates, in this case'the dose rate to the thyroid, are calculated by:
.Dg '('t,x)~= 3600-B.Kg Qg(t,x) ,
(62)', i where B = the breathing rate, m 3 /sec, and l' Kg =
'the internal dose effectivity of the ith isotope, rem /Ci.
The' values'for the breathing rate are given in Table 7-5 and are taken from a published regulatory guide [11]. The average gamma ray energy.per disintegration and the internal dose effectivity for each, isotope considered are given in Table 7-6. The-decay products of these isotopes are also included in the ca).culation; however, their' contribution to the dose rates are .small' and' therefore the data. f or these isotopes were not included in the table. 7.3.4. Downwind Doses The whole' body gamma dose and thyroid dose in the lee of the building are shown in Table 7-7. These doses are acceptable relative to the conservative -nature of the calculations and likelihood that an accident scenario would actually lead tosthese results. i 7-30
5 kJ ' [
'SAR19/84 Table 7-5 AS$tfMED H Ris ATil LNG H ATE:; l Time'(hr)- . Breathing-Rate-(m /sec)-
0 to 8 -4 3.47-x-10
^8 to 24 -4 1.75 x-10 Over.24 2.32 x 10 -4 t.
Table.7-6 AVERAGE GAMMA RAY ENERGY AND INTERNAL DOSE EFFECTIVITY FOR EACH FISSION PRODUCT ISOTOPE Isotope- Eg(MeV) Kg(rem /C1) ,-
-2 Br-83 0.92 x 10 Br-84 1.87 6 I-131 0.40 1.486 x.10 0-
. 1-132 l'.96 5.288 x 10 I-133 0.56 3.951 x 10 5 4_
I-134 3.02 2.538 x LO I-135 1.77 1.231 x 10 5 I-136 2.91 Kr-83m 0.8 x 10 -3 Kr-85m 0.16 Kr-85 0.4 x 10-2 Kr-87 1.07 Kr-88 2.05 Kr-89 2.40 Xe-131m 0.82 x 10-2 Xe-133m' O.37 x 10-I Xe-133 0.29 x 10-1 Xe-135m 0.46 Xe-135- 0.25 Xe'-137 1.22 Xe-138 1.57 7-31
s I, I 1 I DOSES FROM FISSION PRODUCT RELEASE I Dose (Rad 1 Time After Release Balogens Noble Geses Accident conditten (Hr1 Thole Body Internal Whole Body Internal I Core Inventory totesse 0.1 0.3 5.9 s 10 0.19 1.0 s 10[ 1.5NW 3.0 2.1 5.8 s 10 1.14 5.0 s 10 -2 8.0 8.3 4.2 s 10*3 3.57 5.7 s 10
-1 24.0 13.8 1.1 s 10 4 4.51 1.4 s 10 O 720.0 31.4 8.5 s 10 6.37 4.3 s to g
-a o
.g U" i
La Core Inventory Release 3 0.1 0.019 3.8 0.013 5.6 s to ed m bJ 1.5MW 3.0 0.028 6.0 0.017 1.7 s 10[ 3.0 0.028 6.0 0.017 1.7 s 10 4 I w
~I Stagne Element Release 0.1 0.004 8.7 s 10 0.002 1.5 s 10[3 1.0 MW
- 2.22% of energy 1.0 0.031 8.6 s 10,g 0.016 7.4 s 10,4 S.0 0.123 6.2 s 10> 0.053 4.4 s 10
-2 ~3 24.0 0.204 1.6 s 10 0.067 2.7 s 10 -2
~3 720.0 0.465 1.3 s 10 0.090 6.0 s 10
-4 ~4 ~8 Stagle Element Release 0.1 2.8 s 10 0.056 1.9 10 8.3 s 10 ~
1.0NW
- 2.22% of Energy 1.0 4.1 s 10[ 0.089 2.5 10[4 2.5 s 10,7 8.0 4.1 a 10 0.089 2.5 s 10 2.5 a 10 1.) Doses celestated assastag 4 years falt power operation and release frastion of 0.002 %.
2.) Release calculated for semi-taitatte cloud la room votame of 4290 m with so en ventitettom. -> 3.) Release calculateJ st sero distsace with an air chsage rate of tea per hour and M battding dilation factor of 8540 ses/co, g3
%s .
3
G^ ( SAR 9/84 {g Chapter 7 References f
- 1. West. G.B.,-W.L. Whittemore, J.R. Shoptaugh Jr., J.B.
Dee,. C.O. Coffer, "Kinectic Behaviour of TRIGA Reactors", General Atomic Report GA-7882, 1967. 2.: 'Simnad, M.T.,- "The U-ZrH Alloy- : Its Properties and Use in TRIGA Fuels" . - Gene ral Atomic Report E-117-833, c 1980.
- 3. ~-Simnad, M.T., and J.B. Dee, " Equilibrium Dissociation Pressures and Performance of Pulsed U-ZrH Fuels at
. Elevated Temperatures," Gulf General Atomic Report GA-8129, 1967.
- 4. Foushee, F.C., and R.H. Peters, " Summary of TR7GA Fuel Fission Product Release . Experiments," Gulf Energy &
Environmental Systems Report Gulf-EES-A10801, 1971 p. 3.
- 5. Sparrow, E.M., and A.L. Loeffler, " Longitudinal Laminar Flow Between Cylinders Arranged in a Regular Array, AicleJ. 5, No. 3, 325 (1959).
- 6. . Sparrow, E.M.,A.L. Loeffler, Jr., and H.A. Hubbard,
" Heat . Transfer' to Longitudinal Laminar Flow Between Cylinders," T r a n s'- ASME J. of Heat Transfer, Nov. 1961,
- p. 415.
- 7. Etherington, H. (ed.), Nuclear Engineering Handbook, 1st ed., McGraw-Hill Book Co., New York 195t, p. 9-1.
- 8. Peterson, J.F., " TAC 2D, A General Purpose, Two-Dimensional Heat-Transfer Computer Code -
User's Manual," Gulf General Atomic Report GA-8869, 1969. / 9. Way, K. and E.P. Wigner, " Radiation from Fission Products", Physics Review, 70 p. 115, 1946.
- 10. Lee, E., R.J. Mack,- and D.B. Sedgeley, "GADOSE and
.DOSET Programs to Calculate Environmental Consequences of Radioactivity Release," Gulf General Atomic Report GA-6511 (Rev.), 1969.
- 11. " Programs for the Monitoring Radioactivity in the Environs of Nuclear Power Plants". U.S. NRC Regulatory Guide 4.1.
7-33
E-SAR 9/84 Chapter 8 FACILITY ADMINISTRATION A TRIGA type reactor facility will be owned and operated by The University af Texas at Austin. The facility located at the Balcones P.esearch Center is to be operated as part of the Nuclear Engineering Teaching Program, a division of the Mechanical Engineering Department in the College of Engineering. Licenses for the facility operation will
' include a facility specific license for operation issued by the U.S. Nuclear Regulatory Commission and a university broad license for radioactive materials issued by the State of Texas Department of Health. Additional licenses may be obtained as required for facility activities. Operation as a utilization facility will be for education, training, and the conduct of research and development activities.
8.1 ORGANIZATION 8.1.1 Structure. Figure 8-1 illustrates the organizational structure that is applied to the management and operation of the reactor facility. Responsibility for the safe operation of the reactor facility is a function of the management structure of Figure 8-1 [1]. The responsibilities include safeguarding the public and staff , from undue radiation exposures and adherence to license or other operation constraints. Facility staff is typically organized as three full time persons consisting of a supervisor, operator or technician, and researchist, and two half-time persons consisting of an operator and secretary. Faculty, researchers and students supplement the organization. Descriptions of key components of the organization follow. 8.1.2 Vice President for Research and Academic Affairs. Research and educational programs are administered through the office of the Vice President for Research and Academic Affairs with functions delegated to the Dean of the College of Engineering and Chairman of the Mechanical Engineering Department. 8.1.3 Director of Nuclear Engineering Teaching Laboratory. Nuclear Engineering Teaching Laboratory programs are directed by a faculty member of the Mechanical Engineering Department that teaches courses in nuclear engineering and performs research related to nuclear applications. 8-1
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Office of the President University of Texas J at Austin
> Vice President for University Safety Radiation Safety C Business Affairs Manager Officer
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.,3 $ Radiation Safety rM Committee oo >
cH cm et M l 3 m< w ps Vice President for oo Research and Academic 8
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, c i l O Dean of the College l c of Engineering i' W PS Chainnan of the Departsrent Director of The Reactor Supenisor of Mechanical Engineering Nuclear Engineering Teaching taboratory Reactor Connittee us W e N CD
l SAR 9/84 8.1.4 Radiation Safety Committee. The Radiation Safety Committee is established through the Office of the President of The_ University of Texas at Austin. Responsibilities of the committee are broad and include all policies and practices regarding the license, purchase, shipment, use, monitoring, disposal, and transfer of radioisotopes or sources of ionizing radiation at The University of Texas at Austin. The President shall appoint at least three members to the Committee and appoint one as Chairperson. The Committee will meet at least once each year on a called basis or as required to approve formally applications to use radioactive materials. The Radiation Safety Committee shall be consulted by the University Safety Office concerning any unusual or exceptional action that affects the administration of the Radiation Safety Program. 8.1.5 Radiation Safety Officer. A Radiation Safety Officer acts as the delegated authority of the Radiation Safety Committee in the daily implementation of policies and practices regarding the safe use of radioisotopes and sources of radiation as determined by the Radiation Safety Committee. The Radiation Safety Program is administered through the University Safety Office and University Safety Engineer. The responsibilities of the Radiation Safety Officer are outlined in The University of Texas at Austin Manual of Radiation Safety. ' 8.1.6 Reactor Committee. The Reactor Committee is established through the Office of the Dean of the College of Engineering of The University of Texas at Austin. Broad responsibilities of the committee include the evaluation, review, and approval of facili'ty standards for safe operation. The Dean shall appoint at least three members to the Committee that represent a broad spectrum of expertise appropriate to reactor technology. The committee will meet at least twice each calendar year or more frequently as circumstances warrant. The Reactor Committee shall be consulted by the Nuclear Engineering Teaching Laboratory concerning unusual or exceptional actions that affect administration of the reactor program. 8.1.7 Laboratory Supervisor. The aparation of the Nuclear Engineering Teaching Laboratory is governed by a Laboratory or Reactor Supervisor, who shall be qualified as ' a USNRC licensed Senior Operator for The University of Texas at Austin TRIGA Reactor facility. Responsibilities of the supervisor include reactor operation, equipment maintenance, experiment operation, instruction of persons with access to laboratory areas, and development of research activities. A 8-3 O
SAR 9/84 UT TRIGA Operations Manual will be maintained by the Reactor Supervisor. 8.2 QUALIFICATIONS 8.2.1 General. Personnel associated with the research reactor facility shall have a combination of academic training, experience, skills, and health commensurate with the responsibility to provide reasonable assurance that decisions and actions during all normal and abnormal conditions will be such that the reactor is operated in a safe manner. 8.2.2 Academic Administration and Radiological Safety. Administrative positions not principally responsible for facility operation and staff positions of the radiological safety program are subject to qualifications standards determined by the University. Administrative qualificationa depend on academic credentials appropriate to the nature of the University organization. Staff qualifications for radiological safety are subject to personnel descriptions developed for various University employment positions. 8.2.3 Facility Director. A combination of academic training and nuclear experience will fulfill the qualifications for the individual identified as the facility director. A total of six years experience will be required. Academic training in engineering or science with completion of a baccalaureate degree may account for up to four of the six years experience. 8.2.4 Reactor Supervisor. A person with special training to supervise reactor operation and related functions will be designated as the facility supervisor. The reactor supervisor will be qualified by certification as a senior operator as determined by the licensing agency. Additional academic or nuclear experience will be required as necessary for the supervisor to perform adequately the duties associated with facility activities. 8.2.5 Operators, Technicians, and others. Qualifications for operators will be determined by the certification of the licensing agency for either senior operator or operator permits. Qualifications for technicians will be determined by training and experience appropriate to the required duties to be performed. A consolidation of the duties of operator and technician may occur 'to better utilize staff resources. Other persons with access to the laboratory will be qualified by academic experience or by special training and instruction of persons with operator certifications. 8-4
e SAR 9/84 8.3 REACTOR OPERATIONS Operation of the reactor and activities associated with the reactor, control system, instrument system, radiation monitoring-system and engineered safety features will be the function of staff personnel with the appropriate license certifications [2]. Operation will include the implementation of required procedures, execution of appropriate experiments, actions related to safety and the preparation of required reports and records. 8.3.1 Staffing. All activities that require the presence of license certified operators will also require the presence in the facility complex of a second person capable of performing prescribed written instructions. Unexpected absence of a second person for greater than two hours will be acceptable if immediate action is taken to obtain a replacement. A designated license certified senior operator will be readily available on call during all periods in which activities requiring a certified operator are being performed. The person on call will be considered available if less than 1 hour is required to initiate a call request and respond on site. Movement of fuel or control rods and relocation of experiments with greater than one dollar reactivity worth will require the presence of a person license certified as a senior operator. Other activities such as initial startup, recovery from unscheduled shutdowns and modifications to control and instrument systems, radiation measurement equipment or engineered safety featuces will require concurrence and documentation of a license :ertified senior operator. Operation of reactor controls, movement of reactor experiments, maintenance of control, instrument and radiation measurement systems will require the presence of a license certified operator. A license certified operator will be present in the control room whenever the reactor is not shut down by more than one dollar of reactivity or the control and instrument panel is not secured. The staff required for performing experiments with the reactor will be determined by a classification systep specified for the experiments. Requirements will range from the presence of a certified operator for some r o u t in'e experiments to the presence of a senior opetator and the experimenter for other less routine experiments. Other activities that occur in the area of the reactor will require knowledge of a license certified operator but not necessarily the presence of the operator. Such activities will include maintenance, handling of radioactive materials and experiment preparation. 8-5
m r> SAR 9/84 L 5 t s 8.3.'2 Procedures. Written procedures shall' govern many of' .the. activities associated . with- reactor operation. Preparation of_the procedures and minor modifications of the procedures will be- by certified operators. Substantive changes to procedures or major modifications, and prepared procedures will=be submitted . to the' reactor committee for .c reviev. and: approval. ' Temporary- deviations from the procedurer, may' be made by the reactor supervisor or designated-senior operator.'provided changes of ~ substance are reported for review and approval.
- a. Activities subject to written procedures will include routine startup, shutdown-and operation of the reactor; fuel
_ loading, unloading and movement within the reactor; . routine maintenance of maj or components of systems that could have. { an' affect 'on reactor safety; surveillance tests and
, calibiations'that may effect reactor safety; administrative
, controls for operation and maintenance that could effect core reactivity or reactor safety; ' personnel radiation protection and implementation of the emergency plan.
8.3.3 Experiments. Proposed -experiments will be submitted to the reactor committee.for review and approval of the experiment and its safety analysis [3]. Substantiv.e changes -to approved experiments will require reapproval while insignificant changes that do not alter experiment safety may be approved by the reactor supervisor or designated senior operator. Experiments will be approved' first as proposed experiments for one time application and subsequently as approved experiments for repeated applications following - a review of the results and experience of the initial' experiment implementation. Each experiment will be designated- as one of three classes. One class will consist of experiments such as-routine- reactor operation for calibration or instruction, and routine irradiations 'such as neutron activation analysis. This class of experiment will require only the reactor operator during the reactor operation or experiment set up. 'A few experiments may> require the' presence of both a' certified operator and the experimenter and will be designated as a seperate class lof' experiment. Another class
. of experiments will be specified for experiments 'that require large reactivity changes such as experiment facility movement, fuel or control rod movement, or significant cha'nges to shielding of core radiation. This class will require the supervision of a senior operator. ,
8.4 ACTIONS AND REPORTS 8.4.1 Operatina Reports. Routine annual reports covering the activities of the reactor facility during the
- previous calendar year shall be submitted to licensing 8-6
SAR 9/84 authorities within three months following the end of each prescribed year. Each annual operating report shall include the following information: i f: (a) A narrative summary of reactor operating experience including the energy produced by the reactor or the t hours the reactor was critical, or both. (b) The unscheduled shutdowns including, where applicable, corrective action taken to preclude recurrence. (c) Tabulation of major preventive and corrective maintenance operations having safety significance. (d) Tabulation of major changes in the reactor facility and procedures, and tabulation of new tests or experiments, or both, that are significantly different from those performed previously, including conclusions that no unreviewed safety questions were involved. (e) A summary of the nature and amount of radioactive effluents released or discharged to the environs beyond the e fective control of the owner-operator as determined at or before the point of such release or discharge. The summary shall include to the extent practicable an estimate of individual radionuclides present in the effluent. If the estimated average release after dilution or diffusion is less than 25% of the concentration allowed or recommended, a statement to this effect is sufficient. (f) A summarized result of environmental surveys performed outside the facility. (g) A summary of exposures received by facility personnel and visitors where such exposures are greater than 25% of that allowed or recommended. 8.4.2 Safety Limit Violation. Actions to be taken in the case of safety limit violation shall include cessation of reactor operations until a resumption is authorized by the licensing authority, a prompt report of violation to license authorities and management, and a subsequent follow-up report reviewed by the reactor committee and submitted to the license authority. The follow-up report shall describe applicable circumstances leading to the violation including causes and contributing factors that are known, effect of the violation upon reactor facility components, systems or structures, health and safety of personnel and the public; and corrective action to prevent recurrence. Prompt reporting of the event shall be by telephone and confirmed by written correspondence within 24 hours. A written report is to be submitted within 14 days. 8.4.3 Release of Radioactivity. Actions to be taken in the case of release of radioactivity from the site above allowable limits shall include a return to normal operation 8-7
J SAR 9/84
;9
!sf e or reactor : shutdown. until authorized by management if
' necessary;to ~ correct the occurrence, a report to management i ,
- -and license 1. authority, and a - review - of the event by - the
- reactor committee at the next scheduled meeting. Prompt reporting ofithe! event shall be by telephone and confirmed
- by written correspondence'within 24 hours. 'A written report is,to be submitted within 14 days.
8.4.'4 Other Reportable Occurrences. Other events that will be considered reportable - events are 711sted in this section. :A return toi normal operation or curtailed
. operation- until authorized by = management will occur.
-Appropriate reports. shall be- submitted to license-authorities. (Note: Where components or systems are provided ' _ in addition: to those' required by the technical specifications, the. failure of components or systems is not.
considered -reportable provided that the minimum number of components or_ -systems - specified or ~ required perform their
- intended reactor safety function.)
(a)-Operation with actual safety-system settings for required systems less conservative than the limiting safety system _ settings specified in the
. . technical 1 specifications.
(b)~0peration;in violation of limiting ~ conditions for. operation established in the technical specifications unless prompt remedial action is taken. (c) A-reactor safety system component malfunction which renders'or_could render the reactor safety system incapable of performing its intended oafety function unless the malfunction or condition is discovered during maintenance tests or periods of reactor-shutdowns.
-(d) Abnormal and significant degradation in reactor
, -fuel,.or cladding, or both, coolant boundary, or confinement boundary (excluding minor leaks) where applicable which could result in exceeding prescribed radiation exposure limits of personnel or environment, or both.
(e) An observed inadequacy in the implementation of administrative or procedural controls such that the inadequacy causes or could'have caused the existence or development of an unsafe condition with regard to reactor operations. , A-8.4.5.0ther Reports. A written report within 30 days to
^ the chartering or licensing authorities of:
(a) Permanent changes in the facility organization
. involving Director or Supervisor.
(b) Significant changes in the transient or accident analysis as described in the Safety Analysis Report'. 8-8
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)
, SAR 9/84
' 8.'5 L RECORDS -
g Records:-of the following activities shall be maintained
.The
- and ' retained f or the periods' specified' below [4].
-* records.may ~~ beLin the: form of logs, d ata . shee t s ,- or other
. suitable forms.- _The required _inf ormation may ~ be contained
~
~
in' single or; multiple records, or a combination thereof. 8.5.1 Records to be Retained f or 'the' Lif etime of the Reactor Facility:. (Note: Applicable annual reports, if they contain all of the required-information, may be used asL records ~in this section.) J (a)' Gaseous an'd! liquid radioactive affluents released' to the environs.
'(b) Offsite environmental monitoring surveys required by' technical specifications.
(c); Radiation exposure for all personnel monitored. (d) Updated drawings of the-reactor facility. 8.5.2 Records to-be Retained for a Period of'at- Least Five' Years or for the Life of the Component Involved Whichever is Shorter.. (a) Normal reactor facility operation (supporting documents such as checklists, log l sheets, etc.. shall be~ maintained for a period of at least one year). (b). Principal maintenance operations. (c) Reportable occurrences. (d) Surveillance activities required by. technical _ specifications. (" L (a)_ Reactor facility radiation.and contamination surveys where required by applicable. regulations. (f) Experiments performed with the reactor. (g). Fuel inventories, receipts. and shipments. (h) Approved changes in operating' procedures.
-(i) Records of meeting and audit'reportsLof the review and audit group.
~ '
.8.5.3 Records to be Retained for at Least One Training
' Cycle. Retraining and requalifications of certified operations personnel. Records'of'the most recent. complete cycle ~ shall be maintained at. all times the individual 'is employed.
i 8-9 e,
7_ _ _ t. SAR 9/84 Chapter 8 References
- 1. " Standard for Administrative Controls" ANSI /ANS - 15.18 1979.
- 2. " Selection and Training of Personnel for Research Reactors", ANS I/ ANS -- '15.4 - 1970 (N380).
3 " Review of Experiments'for Research Reactors", ANSI /ANS 15.6 - 1974 (N401)..
- 4. -" Records and Reports for Research Reactors", ANSI /ANS -
15.3 - 1974 (N399). i 8-10
SAR 9/84 Chapter 9 L QUALITY ASSURANCE PROGRAM Obj ec t ive s of quality assurance (QA) may be divided into two maj or goals. First is the goal of safe operation of equipment and activities to prevent or mitigate an impact on public health and safety. Second is the reliable operation of equipment and activities associated with education and research functions of the University. The risk or potential release of radioactive materials is the primary impact on public health and safety, and may be divided into direct risks and indirect risks. Direct risks are activities such as waste disposal, fuel, transport and decommissioning. that introduce radioactive materials into the public domain. Indirect risks are accident conditions created by normal or abnormal operating conditions that generate the potential or actual release of radioactive materials from the controlled areas of a facility.
9.1 INTRODUCTION
Characteristics of uranium loaded zirconium hydride fuel used in the TRIGA reactor provide substantial benefits to safe reactor operation. Many accident situations are simulated by normal operation of the fuel in either pulse mode or steady state mode. Other features such as fission
) product retention, stainless steel cladding design, facility engineered features, and periodic schedule of operation combine with routine operation procedures to decrease the consequences of failure of any reactor components. The limited scope of application of formal quality assurance criteria is due to the fact that most parts and procedures associated with operation of the TRIGA type reactor are not relevant to public health and safety.
Safety-related identifications for quality assurance are determined from license specifications. The specifications for safe operation include design features, safety limits and limiting conditions for operation. The application of quality assurance shall be considered for those structures, systems or components in the technical specifications that are either design features or required as limiting conditions for operation. Such systems should include the control and safety system, radiation monitoring system and other support systems. Although the quality assurance program is not applied to routine reactor operations and surveillance activities, the program shall be implemented for non-routine activities determined to be 9-1
SAR 9/84 safety related in ' nature or , effecting safety related equipment. Such activities ~ shall include design, construction, testing, modification and maintenance of safety related iters. (. Two additional conditions remain, however, that are important to the applJeation of at least portions of the quality assurance program. One is the safety to operation personnel and experimenters and the other is continuity of the operations. programs. Each of these cond,itions must be examined objectively relative to operation procedures and program expectations. In general, the application of good industry quality assurance prohtices is suf ficient to meet operational program goals. The quality assurance _ program s h a ll' be commensurate with the TRIGA ' type reactor,- The University of Texas administrative programs and the goals of quality assurance. This document provides requirements for establishing, managing, conducting and evaluating the QA Program. The QA Program applied to items or activities determined to be safety-related follows the guidelines of Nuclear Regulatory Guide 5.2 (77/05) [1,2]. 9.1.1 Purpose. Qualir,y assurance of certain activities associated with the University of Texas TRIGA reactor facility is important for the safe and efficient completion of tasks that are identified as safety-related. This document outlines the Beneral elements of quality assurance applied to safety-related strt_ictures, systems or components, and activities. Require 4ents are documented for establishing, managing, . conducting, and evaluating the QA Program. Although aspects of the OA Program may be routinely applied to many faci'lity actfvities, the formal implementation of the program is limiccd. to specific items or activities related to public health and safety.- 9.1.2 Responsibility. The University of Texas at Austin as owner and operator of the TRIGA reactor facility shall be responsible 'for a quality assurance program. The owner-operator shall a,1tablish andimimplement a progran consistent with the goals of,s qtia lity assurance for safety-related activitiec, structures, systems and components. Identification of safety-related items shall be the responsibility of the owner-operator and will include a description of the item and the applicable elements of the quality assurance program. Special _ qpality provisions, delegated functions of'the sprogram, and ' unresolved quality assurance problems shall ,31e be identified by the owner-operator. Table 9-1 lists'the responnibf11 ties and key personnel particip'ating in theqUniversity TRIGA QA Program. o
, I 9-2 >
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>:p Table 9.1 RESPONSIBILITIES'AND KEY ~ PERSONNEL
-~ Re s p o n s'ib ilit ie s ' ' Key -- Univ e r s it y Personnel l'.' Establish' program Director or Supervisor Implement program of1TRIGA facility A Safety-related' identification f2. Unresolved issues : President or Vice President for
-Academic Affairs andL Research L3. Delegated functions Faculty and staff
- 4. . Specialized-functions- Specified personnel
. 9.1. 3 ' organiz a t ion ._ The organization applied to quality assurance activities shall be part of the normal _ university
- administrative structure.- The facility Supervisor shall develop and implement the quality; assurance program and.
identify safe.ty related items. Unresolved: issues:of quality-assurance shall be reported-to'the Director of-the facility and the : appropriate- administrative Lvice president of the
- university. Execution of specific elements of the program may be, delegated to' persons.in the University organization or other organizations. as appropriate. University persons shall' include committees,- faculty, researchers or staff as required for specific; program applications. Non-university organizations ~ or persons shall supplement University personnel when specialized' qualifications are necessary for specific: quality assurance tasks. The University
-organization applied to reactor safety and quality assurance is.the academic administration represented by Figure 9-1.
9.1.4 -Documentation. All activities affecting safety-related items subject to the quality assurance program shall be identified and- documented f - ormally. . The
-format of Table 9-2 shall be used to identify applicable-e,lements of the Quality Assurance Program and identify documents, procedures, reviews, inspections, tests, or'other quality assurance features that are to be applied to a safety-related activity.
i 9-3
m- ,
,s-SAR 9/84 I'
office of the President . . University of Texas ., at Austin Vice President for ' i Research and Academic Affairs
'l g. t . ,
Dean of the College of _ Engineering _ Reactor' Committee. Chairman of the Department of Mechanical Engineering i 1 s Director and/or Supervisor of The Nuclear Engineering Teaching Laboratory-I License Certified Operators ACADEMIC ORGANIZATION a Figure 9-1 ( ,. 9-4
SAR 9/84 Table 9.2 FORMAT FOR SAFETY RELATED QA-CHECKS-Each. safety-related activity structure, system, or component will be given a letter symbol, such as A, B C, and be. - appended with the- 'following- designations (for example, A1.0):. 1.0 Title ~ Identification and description of safety-related item 1.1 Participation:- supplemental organization and functions l '. 2 Documents --applicable procedures or special measures . - 2 .1 ' Design Control 2 .1.' 1 ' Codes.. standards and regulations 2.1.2 . Method of verification
-2.1.3 Modifications. proposed
- 2. 2 - Procurement Control 2.2.1 Codes, standards.and regulations,
~
2.2.2 Quality assurance. specifications 2.2.3. Proposed changes enacted 2.2.4 Procurement conformance. method 2.3: Document Controli
- 2. 4' haterial-Control 2.4.1 Special, procedures-required.
- 2.4.2 Equipment-required -
, -2.4.3 Personnel.qualificatior 2.5 ' Process Control 2;5.1 Special procedures 2.5.2- Special equipment 2.5.3 Personnel qualifications
- 3.1 Inspsetion Program Description
. 3.2 Test Program Description
. 3.3 Measurement Equipment
~3.4 Nonconformance-Item.and Disposition
. 3.5 . Corrective Actions-Instituted i 4.0 TRecords List , 9-5 4 5 y : -- e nr~p- r -- -
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SAR 9/84 9.2 QUALITY ASSURANCE CONTROLS 9.2.1 Design Controls. Design controls shall consist of design specifications, references to applicable codes, standards and regulations, design verifications and document approval. Applicable codes, standards, regulations or other quality requirements will be identified and requirements incorporated into the design documents. Design document approvals should be part of the design document. Design approval will be by a person, other than the design originator, that is knowledgeable of the design and quality requirements or knowledgeable of the designers qualifications. Modifications of safety-related documents shall be subject to the same provisions as the original document. Approval of the design modification will be included with the design document and the modification identified. Verification of design adequacy shall be provided by either design reviews, alternate calculation, test program or other method, '+ determined to be appropriate. Verifications of the design should check characteristics such as compatibility of materials; suitability of application of inspection, maintenance and repair; proper interfacing of sub-systems, and proper acceptance criteria. Method of verification will be identified and documented by approval of the design document. 9.2.2 Procurement Controls. Procurement controls shall consist of procurement specifications, references to applicable codes, standards and regulations, procurement acceptance and document approval. Applicable codes, standards, regulations or other quality requirements will be identified and references incorporated into the procurement documents. Procurement document approvals shall be part of the procurement document. Procurement approval will be by a person, other than procurement originator, that is knowledgeable of the procurement and quality requirements or knowledgeable of the procurers qualifications. Changes to safety-related procurement documents shall be subject to the same provisions as the original document. Approval of procurement changes will be included with thd procurement document and the change identified. > Acceptance of procured items or services shall consist of evidence provided by the contractor, evaluation of the procurement source, inspection at the source or inspection upon receipt. Acceptance of the procurement should require measures such as quality assurance by contractor, inspection and test functions, or controls on materials processes and nonconformances. The methods of acceptance will be 9-6
~ SAR'9/84 a
identified and documented by approval of the procurement document.
~
'9.2.3 Document Control. Document centrol consists of monitoring the. development, revision, release and use of
; documents, drawings or specifications affecting safety-related activities; Document control shall include
_ assurance that- safety-related documents are identified as such,. and are . completed and . maintained properly. The
~1aboratory Supervisor shall provide control of safety-related documents that are specified according to the format of Table 9-2.
9.2.4 Material control. Procedures shall be written to establish material control. when special n}easures are necessary to assure material quality of safeby-related items. Controls ~ shall- be applied to a c tivitiie s duch as identification handling, storage, shipping, cleaning and preservation Procedures shall specify. equipment .and personnel required to accomplish the specified_ material
^
control. Applicable. codes, standards,- specifications or 4 personnel qualifications shall be documented. 9.2.'5 Process Control. Procedures ~ shall be written to
- establish process control when- special measures are
, necessary to assure process quality of safety-related items.
Controls 'shall- be applied to activities such as crimping, soldering. welding, painting, cleaning .and heat treating. Procedures 'shall specify qualifications .of. equipment 'and personnel required to perform the appropriate process control. Applicable codes, standards, specifications or
. personnel qualifications shall be-documented.
9.3 : INSPECTION AND CORRECTIVE ACTIONS 9.3.1 Inspection Program. An-inspection program shall be established for safety related items or activities. The inspectjon program shall- apply .to construction, procuraments, experiment equipment fabrication,- and modifications that effect safety-related structures, systems, or' components. Maintenance . persons delegated to
- perform inspections shall not be the same person involved in
- the- safety-related activity but may be from the same
; organization. ii The inspection program will consist of written
+ procedures that will include, as appropriate, procedures ! specifying characteristics. to be inspected, acceptance criteria ~and inspection hold points. 1 Procedures should provide for identification of ,; inspected, tested and non-conforming items. Procedures F 9-7 9
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SAR 9/84 shall: .also- .be written as. necessary for monitoring
- E activities.. ,
9.'3'.2 Test Program. A test- program shall .be. established for safety-related items or activities. The
, . test. program shall apply to .
prototype qualifications, installation proofs-and functional tests. Testing shall!be performed in accordance with acceptance criteria derived from design.or procurement documents. The - test program will - consist of written procedures that will. include,- as appropriate, procedures that specify acceptance criteria, monitoring requirements. . equipment required, personnel ' qualifications, environmental conditions,' data acquisition and documentation of results. 9.3.3 Measuring and Test Equipment. Measurement tools, guages, instruments, and other measuring or test devices that measure critical parameters of safety-related items
=
shall be identified. Provisions for identified measuring and test devices shall include availability. adjustment, calibration-and accuracy as required for each application. Test' equipment will be identified. 9.~3.4 Non-Conforming Material and Parts.
-Non-conforming materials and parts associated . with safety-related structures, . systems or components shall be identified._ The disposition such as acceptance, repair, rework or rejection of parts from safety-related functions will be determined by -the person responsible for document control. Repair or. reworked parts will be removed .or labeled.until accepted. Rejected parts will be removed-and (labeled. Significant non-conformances and the disposition will be documented.
9.3.5 Corrective Action. Documentation of specified quality control .or assurance . documents shall provide
. evidence of quality of safety-related items. Significant deviations from acceptable quality', repeated quality
_ problems or unresolved ' quality issues shall be noted and
,g\ ; reported in' writing to administrative management personnel.
'l
- s. . It should be recognized that a determination of a quality-r problem may be subjective and should include evaluation of
.the documented quality requirements relative to the impact on the safety-related nature of the item.
9.4~ RECORDS AND AUDITS . 9.4.1 Quality Assurance Records. Records that document f~ ' quality of. safety-related items or activities are identified according. to ' Table 9-2. The records identified consist of i. inspection .and test results, quality assurance reviews, quality. assurance- procedures and engineering analysis in 9-8 e_
=
SAR 9/84 l
- support of design modifications or changes. The records
. shall be' retained-with as-built drawings, manuals and other records of - important, facility and system information. The retention period.is to.be the life of the facility or system for-most, if not 'all, saf e ty-related items. The retention period ~ is indicative of _ the expectation that ' items which affect saf ety ' related to a TRIGA reactor are integrally
, related_to-the - reactor, instrumentation and facility design.
and'should persist for.the system or facility life. 9.4.2 Audits. An audit shall be-conducted to examine
- the records and function of the quality assurance program.
Audits will occur within two years of the QA- Program activities by- designated persons that were not directly
~
responsible for the audited functions. Written procedures, Table 9-3, f or the . audit will be considered .part of the
- Quality Assurance _ Program. A report.of the- audit ' results ,
actions to 'remolve deficiencies and evaluation of the program will be made to a facility operations committee - and university' administrative management, and maintained with other-Quality Assurance Program. documents. 1 t } 4 l rh g
- l 9-9
SAR 9/84
' Table 9-3 QUALITY ASSURANCE PROGRAM s
AUDIT PROCEDURES
- 1. Designate a person'or persons responsible to perform' the program audit.
- 2. Determine the date of_the previous audit.
3.. Review the Quality Assurance Program document.
.4. . Examine the list of safety-related items.
- 5. . Note additions to.the safety-related items.
6.. Identify records applicable to additional items.
- 7. Determine the location of all indicated records.
8.. -Review records for abnormalities and completeness.
- 9. Prepare statement that evaluates functions of Quality Assurance Program.
- 10. Report findings of audit and program functions to operations committee and management.
q. t !^ i f d s , 9-10 M. C_-. .
SAR-9/84
. Chapter 9 References 1.- " Quality Assurance Requirements for Research Reactors",
, . Nuclear Regulatory Guide .. 5.2-(77/05)
- 2. " Quality Assurance Program Requirements for Research Reactors", ANSI /ANS --15.8 - 1976 (N402)
Y A r , f t t 3 a l' 5; 9-11
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SAR 9/84 Chapter 10 RADIOLOGICAL PROTECTION PROGRAM Protection of personnel and the general public against hazards of radioactivity a_nd fire is established through the safety programs. of the University Safety Office.
-Implementation supplements of safety programs at the reactor facility the university programs so that appropriate safety. measures are established for the special characteristics of the facility [1,2].
10.1 RADIOLOGICAL MANMEMENT ORGANIZATION 10.1.1 Management and Policy. Radiological management policy shall- include a commitment to keep occupational exposures as low as is reasonably achievable to facility
-personnel and the general public. Other elements of the radiological management will include:
- a. instruction of personnel in awareness of the low as reasonably achievable commitment,
- b. identification of radiation protection personnel and their responsibilities,
- c. authority of personnel to communicate with management and modify or suspend activities for reasons of radiation protection,
- d. assurance of sufficient and appropriate training of personnel in radiological. safety, ,
- e. periodic evaluations of the program to determine possibilities for lower radiation exposures.
Suggestions and recommendations for modifications to operating and maintenance procedures and to reactor equipment and facilities shall be considered by management to reduce exposure to radiation. Implementation of modifications will occur if substantial exposure reductions are possible at acceptable cost. 10.1.2 Kesponsibilities. Radiation protection at the reactor facility is the responsibility of the Reactor Supervisor or a designated . senior operator in charge of operation activities. Responsibility shall include the authority to act on questions of radiation protection, the acquisition of appropriate training for radiation protection and the reporting to management of problems associated with radiation protection. 10-1
_c' - SAR 9/84 j
. 1
. 1 10.1. 3 7.Organiz ational = A'e ces s . The person, responsible
. f or '. radiation protection at the reactor facility'will have access to other- individuals ^ o r. , groups responsible. for radiological saf ety. at the = University. C o n t a c t -( w i t h the
' Radiation Saf ety i of ficer will occur on an.as needed basis
'and contac'. ;with 'the Reactor. Committee- will occur on a e
y periodic basis. [ L10.1~. 4 . Equipment ' and Supplies. Equipment and supplies 5 maintained for radiological. safety management.shall_ include:
~
a.) fixed area radiation monitors, b.) air particulate monitor,
- c.) gaseous efflaent; monitor, d.)' portable'_ radiation monitors, l
'e, Y detectors- for contamination measurement.
- f .- ) maintenance and calibration capability for equipment, 5-
_' g . ) laboratory _ counting and analysis equipment, h.). supplies for storage of contaminated equipment, i.) provisions for radioactive waste disposal, j'.)Ldecontamination facilities, k.) protective clothing. 1.) resp'iratory protection equipment, m .' ) and emergency response equipment. l, 10.1.~ 5 - Training and Safety. Each person in the restricted, . area of the reactor . facility .shall have , sufficient radiological safety t raining _- f or the purpose of access to the area or be escorted by a person with the 1, appropriate - training. Training will-be appropriate to the , activities 'of persons . admitted. to the area and will range-
~from simple instructions'of emergency-alarms and evacuation procedures to more complex implementation of the area emergency plan.
Training for' fac'ility perso'unel shall be'specified_by the Reactor Supervisor and shall provide sufficient training-in radiation: safety policies and-procedures, and in the use of radiation saf ety. equipment : located in the f acility - to control exposures during normal, abnormal and emergency M situations. Training will consist of: a.)~ radiological safety policien, plans and procedures.
-b.) radiation' hazards and_ health risks. '
~c.) use of protective clothing and equipment,
. d.) usenof portable radiation. monitoring equipment, E e;) and other documents such as the emergency plan and federal.or state' notices to workers.
An evaluation shall occur every two years to determine whether. . additional training of : personnel is -required. and that' 'the radiological safety- program is functioning
- adequat'ely.
10-2 4
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Safety ^ ' programs, . with the exception of. reactor. operations, are operated as .a function of the business administration of s the University and include a radiation
~
safety organization as. presented'in Figure 10-1.
- 10.2' RAD'I0 ACTIVE' MATERIALS' CONTROL Physical control .of ra d io a'c t iv e materials shall be provided ; as. an essential part o f. the radiological saf ety
' program.
Control shall' include identification of items.or storage in identified locations. . Controls such as shielding, -isolation, . containment and ventilation will be
~
. provided, as necessary. to". control radiation exposure to the inventory-of radioactive materials.
10.2.1 Reactor-Fuel. Irradiated. reactor fuel shall be maintained in the reactor core, reactor pool . storage racks [or reactor bay storage pits. Fuel elements will be removed
~from these .f acilities only f or - transport , measurement or experimentation. An ' a're a of the reactor facility will be
.' designated for the storage of a few single. fuel elements of unirradiated fuel.before the fuel is moved to other stcrage areas or irradiated'in the reactor.
10.2.2. . Reactor Components. Each reactor component
. removed' from the reactor pool shall be measured for activation levels and removable contamination.- All-components remaining in the pool shall. be assumed .to be
' Components removed from the pool will .be radioactive.
leaned ~ or . covered as necessary to control radioactive centamination. Components'that contain radioactive material will be labeled and stored in an area designated for such components. l'O.2.3 Experiment Facilities. F.xperiment facilities shall" consist -of.all. tubes or penetrations into the' reactor core or reflector that provide access to the reactor neutron flux _~for an experiment application. and shall include-facilities in which materials are exposed to beams i originating from'the reactor core. The cobalt-60 irradiator shall also be-considered'an experiment facility. Removal of' experiment facilities from'the pool'or the beams originating
~ f rom' the : reactor shall be subject to the same controls as
+ those forireactor components.
- 10. 2. 4' Activated Samples. Materials that are inserted
- into reactor experiment facilities or. reactor beams shall bh controlled -as- radioactive materials until disposed as radioactive . waste, transferred to an authorized user or decayed.to releasable-levels for non-radioactive materials.
g 10-3
-SAR 9/84 Office of the' President University of Texas at Austin Vice President for Business Affairs Radiation Safety Comittee University Safety Manager Radiation Safety Fire Marshall Officer Director and/or iupervisor of The Nuclear Engineering Teaching Laboratory BUSINESS ADMINISTRATION I Figure 10-1 i
, 10-4 _ . _. - - . _ . . . . - . _ _ . . _ - _ . _ _ . ~ . . , _ . _ , . _ - . . . . ~ . - _ _ _ . _ . _ _ . _ _ , _ . ~ _ . _ _ . - - _ _ . _ _ . . _ - - . _ . _ . . _ , _ . . , , . _
- _ . - _ - - - - _ . ~ _ , - - _ . -.
w SAR 9/84~ - c 4
---Samples ' exposed .in th'e cobalt-60 irradiator will not . ' be
~
considered . activated as radioactive materials. Specific
' locations that may depend on .the samples _ analytic : status shall-be design'ated . f or. the --- s torage of. activated samples.
' Locations- shall-.-be designated and . labeled for storage . of ~
- ' samples 'and- sample encapsulations, befo'te and after.
-analysis.. -Locations- should be designated. f or storage of sample or encapsulation materials that are decaying and that.
j- .are to be a.): analyzed,. b .' ) d
~isposed, c.). released to an-authorized user, d.)-released:as non-radioactive material, e.)-and.' retrieved-for subsequent use.
10.2.5 Radioactive Waste. Canisters shall be available
. and. lab'eled for. radioactive waste 'at locations where radioactive contamination from sample processing or other activities with contaminated m'aterials occur. A location shall,be designated for storage of solid wastes that are'to be released for. disposal. Liquid wastes shall be maintained in:a designated storage location until release criteria are
- determined such as decay,. dilution or processing. Specific p sinks and drains . in the f acility .that are designated for ,
radioactive materials shall be identified. Gaseous wastes
~ are.to-be vented through low volume-facility hoods according
, to - allowable release criteria. Appropriate monitoring'will. ) be applied as required. ' l _ _10.2.6 Other' Materials. O'ther materials that are to be
- identified-and. controlled by identification-and-location are- ~
encapa ulated . . isotopic radiation sources, radiochemica1' source _ materials and process- equipment or tools possibly
, contaminated with radioactive materials. Activity levels of 4
encapsulated and' radiochemical sources are-expected to vary widely. as will the handling and storage precautions. 1 Activity' levels associated with process equipment or tools ~ shall be identified so that appropriate handling and storage precautions ~can be instituted. t
- lO.3 RADIATION MONITORING '
Radiation monitoring shall consist of fixed, portable, or sampling type systems. Monitoring systems will be-applied to measurement, of radiation areas and high radiation ' areas around the reactor facility, significant contamination within.and' adjacent to the reactor facility and radioactive materials and.their concentrations in effluee s. Monitoring shall be considered for routine operations, abnormal i conditions and emergency situations. 10-5 Jr
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8 7 SAR.9/84 s-
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10.3.1 Minimum Procedures.
- l
- a. ). ' Zone 1 identification, access control and_ protective p . equipment _ .shall be: . der,ignated.,. Zone identification for
. radioactive materials and' radiation . areas are designated as specifie'di by- 10 ~ CFR, ' part 20 (Standards for- Protection '
Agains t' . Radia tion) . Access : control . f or zones shall be to. control 1 r a d i a t i o n e x p o s~u r e s- and physical security of the reactor _f acility _ and its material as " specified by- .10 CFR
- parts 19 -and 7 3 ,; (Notices, Instructions, and Reports to
. Workers; JInspections _ and Physical Protection of Plants and Materials)^.' Protective equipment f or routine abnormal and emergency : conditions shall ' include at least gloves, _ shoe 7 covers, coveralls, half' . mask _ air purifying respirators, tape, plastic bags, and absorbent paper.
. b.) Continuous ~ monitoring or control of radiation
' fields in the restricted area around the reactor shall occur
'whenever levels greater than 100 mrem /hr are produced ~ in -
Laccessible areas. The . radiation levels may. be caused by normal operation of the reactor or an experiment, deviations from normal operation, or easily changed ' shield configurations. Periodic measurement of accessible areas
. should occur in locations with significant radiation levels
-that do not require continuous monitoring. Personnel shall be informed of high. radiation levels and care taken to-prevant inadvertent _ increases in the- levels. Continuous monitoring may be replaced by periodic monitoring .for temporary conditions that do not violate applicable regulations or' license constraints.
_ .. c . ) Contamination areas .or areas that are routinely subject to contamination shall be marked clearly-and control points established to monitor for contamination of personnel or equipment that leaves the designated area. Measurements shall provide action levels for removable activities of 200 disintegrations per minute. Periodic monitoring of areas in which contamination is probable shall be of- adequate frequency to reveal significant changes in contamination levels. Decontamination .of personnel, equipment, and surfaces shall be appropriate to requirements'for control of -) radiation exposure and control of radioactive material containment ~ . - t d.) Airborne radioactive monitoring shall consist of continuous sampling _of. air particulate activity in the reactor- area. Warning levels- and action levels will be
~ determined relative -to allowable maximum permissible ,
concentrations. Measurements should be sensitive to one maximum -permissible ' concentration . change in one hour. Monitoring will occur during reactor operation or activities
nvolving i fuel, core, or experiment facilities and will
--provide measurements for routine, abnormal, and emergency conditions. Additional airborne monitoring equipment should 10-6
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l SAR 9/84 be provided forispecial? experiment needs or locations remote-
- to :the reactor. area particulate monitor.
<e.)' Effluent monitoring .shall. be provided for the -
- discharge of'the radioactive noble gas argon-41.' Monitoring
. . will consist-of either the~use of integrating dosimeters at ;
5- '
.a'locationJofLinterest.or sampling-of a point-in the release
-' path'. . . Measurements ~shall determine , tha t- the' dose at a
- location of int'erest is either less than ten mrem per. year
- . above '_ natural background or two percent. .o f the . allowable ,
maximum ' permissible concentration for the year. Liquid L
' effluents: shal1~ 'be monitored before release by sampling of gross beta-gamma activity. Specific isotopes..should be 4 identified ~ and dilutions : calculated .such that released i ' concentrations averaged over one year do not . exceed 1% of p the allowable maximum permissible concentrations. Other gaseous' or radioactive affluents are to be examined on a case to case basis.
.f . )--- Personnel dosime try shall be required for access to.
reactor. areas and some other facility activities. Monitoring ~ devices will typically be film badges with pocket dosimeters and thermoluminescent detectors for supplemental , [ measurements. Other personnel monitoring such as bioassays or whole body counting will be applied as determined by-the ,
. activity and conditions of radiation exposure situations.
' Personnel shall use supplemental dosimetry.during activities-
!. th'at deviate -substantially from routine operations with supplemental dosimetry also provided for persons visiting . . areas with potential radiation exposures. b 10.3.2 Monitoring Techniques.- Implementation of radiation monitoring to maintain the goal of as. low as reasonably achievable should c'onsist of: b a.) preoperation planning, e b.) operations techniques,-
- c.) and post operation analysis.
+
! 10.3.3 Management Surveillance. A review by management of. radiation exposures related to operations that cause significant radiations. exposures compared to routine , operations will be performed. The review should be applied
- . to determine whether. facility modifications or procedures f should be implemented to maintain radiation exposures as low as reasonably achievable.
h 10.3.4 Frequency and Accuracy. Monitoring frequency and l accuracy'of-activities will be determined by several factors
- -- related to personnel' access , requirements, probability and I
consequences of equipment failure, contamination potential, L
- periodicity of modifications and adequacy of current t
' monitoring. Accepted standards for measurement sensitivity j and accuracy should be appropriate to maintain radiation !
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SAR 9/84 exposures as low as reasonably achievable. Frequency and accuracy specifications should be specified by procedures or other documents when_ appropriate. 10.4 INSTRUMENTATION
-Instrumentation for the evaluation of radiation exposures from routine, abnormal and emergency situations aball consist of fixed area monitors, portable survey monitors and appropriate sampling methods. The minimum instrumentation available during reactor operation shall consist of fixed area gamma dose rate monitors, continuous
.meter, air particulate monitor, portable thin window GM tube survey portable neutron. sensitive counter and pocket
' dosimeters with charger. Other detection equipment that should be available includes alpha-beta proportional counter, multichannel gamma pulse height analyzer, thermoluminescent detector with reader, alpha scintillation detector, high and low range beta-gamma dose rate meters and GM tube. friskers.
10.4.1 Fixed Area Monitors. Fixed area gamma monitors shall the have remote-readouts with audible'and visual alarms at reactor control console. Local readouts should be provided in areas with significant radiation levels and routine personnel access.
.10.4.2 Airborne Radioactivity Monitors. A continuous air particulate monitor with audible and visual alarms shall be functional in the reactor vicinity during reactor operations. A gas monitor system for the noble gas effluent, argon-41, shall also be operable during operation or. sufficient data available to demonstrate a calculated release quantity.
10.4.3 Laboratory Instrumentation. Portable survey i monitors for alpha, beta, gamma or neutron radiation shall be maintained for area surveys of laboratory and experiment i areas. Supplemental measurements should be available with , alpha beta proportional counters or gamma ray pulse height analyzers. 10.4.4 Liquid Effluents. Liquid effluents shall be monitored by sampling methods to determine gross alpha-beta activity. Gamma spectral analysis should be applied for identification of isotope mixtures that require substantial dilution for disposal. Liquid effluents shall be released in batches after storage for decay and dilution determinations. Reactor coolant may be monitored for radioactivity in the coolant or purification loops as a supplemental indicator of water activity. 10-8
SAR 9/84 10.4.5 Range and Spectral Response. Instruments shall be available to measure the various types of radiation and the presence of-low and high levels of radiation. Several types of detectors should be available for measurement determinations. 10.4.6 Calibrations. Calibration methods, accuracy.
-frequency and functional checks shall be established for radiation monitors. Two classes of monitor calibration will be . applied. One class of calibration will consist of monitors applied to routine facility operation and surveys.
Maintenance, calibration and functional checks will be subject to reactor operation specifications. The second class of instruments should have functional. checks at annual intervals but may be calibrated infrequently or at the time of application. 10.5 RECORDS Records are specified for maintenance of radiological data that relate to reactor operation. These records shall include: a.) Personnel dosimetry including bioassays or other special measurements made, b.) Radiological control surveys required by facility specifications, c.) Gaseous and liquid radioactive effluents released to the environment, d.) Radiation survey records, e.) Instrument calibration records, f.) Radioactive material receipt and transfer records, g.) Solid radioactive waste disposal records, h.) Leak tests of sealed sources, i.) Data on radiological incidents. 10.6 EMERGENCY PLAN AND RADIOLOGICAL PROGRAM REVIEW An. emergency plan shall be established, maintained, and implemented by the Reactor Supervisor. The plan will exist as a separate document of the radiological safety program. A review of the radiological safety program and emergency plan should be integrally related. Some partial assessment of the radiological safety program should occur each year such that a complete assessment occurs during a two year period. The two year period shall also apply to the emergency plan. 10-9
- ).
SAR'9/84 c Chapter 10 References ,
- 1. - " Radiological Control at Research Reactor Facilities",
ANSI /ANS - 15.11 1977 (N628) 2.- " Design Obj ec tive s for and Monitoring of _ Systems-Controlling Research Reactor Effluents", ANSI /ANS - 15.12 1977 (N647)
- 3. " Nuclear Regulatory Commission", Chapter 10 U.S. Code' of Federal Regulations s
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SAR.9/84
' Chapter)11 FIRE PROTECTION P
The goal -of prov'ide~
~
fire protection shall. be to Ereas'on'able assurance that safety-related systems perform as intended ~and that other defined. loss criteria are met-[1,2]. For the purpose of fire, protection, . loss crit e ria . sho'u'id 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. I 11.1 FIRELPROTECTION COMPONENTS Each of the three components of : the fire' protection program shall be- applied to the- design, operation and modification of the' reactor facility and its components. 11.1.1 Passive Fire Protection Elements. Passive fire protection'should. provide a potential for fire safety that'
, does not. require physical operation or personal response to achievel the- intended function. Passive elements to be considered should include. inherent design.fcatures, building physical -layout,. safety-related. systems- layout,- fire-
. barriers and construction -or component materials. -Other passive. elemen'es that~ may be considered. for .special conditions;will be frangible walls-for overpressure' relief, curbs ~for containment of hazardous liquids, and drainage for control of. fire protection runoff water. Penetrations in-fire barriers ~- shall . have fire resistant . ratings compatible
-with the purpose of the fire barrier..
Safety-related systems. shall- incorporate passive fire
' protection f eatures that provide protection necessary for the design functions of the system.- Separation of redundant z
- system components, if applicable, and protection ' of distribution systems should be examined. Materials r of noncombustible or limited combustion properties should be used when practical. Materials such as sealing materials, l electrical insulation, structura? finishes, adhesives, and
-linings should be selected so as to minimize fire hazards.
Material selection may consider characteristics such as calorie content, ignition properties, flame retardation and rate of heat ~ release. 11-1 j
> SAR'9/84-11'.1.2- Active Fire Protection Elements. Active ' fire protection elements differ from . passive . elements in that--
active- features require automatic. operation,. manual iresponse,- or -personnel Jaction;;for the intended function.
- Active L elements to be considered should be automatic fire a, detection, automatic fire suppression,. fire information L 4- transmission, manual fire suppression.and other manual--fire -
control. or loss -control' measures. Automatic . protection sys t!ams considered will include smolde de tection , thermal
' d e t e'c t i o n , ' sprinkler action, spray. or . deluge and special protection such as gaseous. extinguisher systems. Fire doors,- dampers, . ventilation control, and. the. inspection, maintenan'ce and testing of. equipment to assure reliability
, and proper operation of the equipment are also considered elements of active fire protection.
Manual protection shall consist'of manual fire fighting 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 shall be utility control,' personnel control and evacuation. Preplanning. shall be an additional element applied to active protection by training f acility personnel and emergency personnel for the appropriate actions'in response of fire and the possible hazards involved. 11.1.3 Fire Prevention Elements. Fire prevention shall be implemented to prevent the occurence of a. fire or limit the probable severity of a fire .that might occur. Elemen'ts of fire prevention shall include control of ignition sources, availability of combustible materials and locations of combustible materials. Controls should "be-applied as necessary on such activities as cutting, welding, other flame operations, electrical equipment and smoking. 11.2 FIRE PROTECTION CONTROLS Management of the Nuclear Engineering Teaching i Laboratory shall be knowledgeable of fire- protection controls.. The controls will consist of actions ' of equipment, actions of laboratory staff and interactions with trained University personnel.
- 11. 2.1~ Facility Fire Protection Elements. Fire protection is recognized as an important element of the safe operation of the TRIGA reactor facility. Commitment by 'the University to fire protection is provided by the functions of the University Safety Office.
The organization for fire protection consists of the
' University Fire Marshall, a member of the University Safety 11-2
R x:. , 4, ' _SAR 9/84
. , w
,0f fice-!and 'the Reactor ' Supervisor, a member of E the Nuclear Engineering Teaching L ab'o r a t o r y .. ' Responsibilities .o f. the' e 1 Fire .: Marshall- are the , maintenance of fire protection equipment .and. inspections for fire prevention.
Responsibilities.of the Reactor' Supervisor-are knowledge of
; potential ~ hazards and implementation- of - fire protection
- recommendations.-
' Although fire protection _ is .provided, f or the general
- saf ety ' of J personnel. and preservation of property . special considerations.shall be provided'for systems designated as safety. related. >Primarily special _ considerations are
-applied to protection of'the reactor and shield structure, and f uel, s torage'. wells . Design features of.these facility components provide a maj or factor of the. fire protection.
Fire. protection for the instrumentation and control system, and L radiation measurement systems are. important for the initial reactor shutdown and the availability in emergency conditions. Fire- protection -of the reactor bay area
' boundary' is of importance to the extent of limiting-either internal conditions that would cause the release of
. hazardous materials or external conditions that would
' threaten the. release of hazardous materials.
N , Loss criteria for decisions on fire protection at- the reactor f acility. shall consist of preventing any injury to personnel, and minimizing the potential or actual release of radioactivity to the. environment. No injury or exposure to the(public should occur from the adverse effects of a fire.
. Laboratory personnel, particularly certified operators, shall.be instructed to observe continually conditions that might represent a risk to- fire protection. -
Appropriate m ' assessment of - the - risk ' should be provided by the Reactor Supervisor and will include consultation- with the Fire Marshall,when appropriate. Passive fire protection elements ef f ectively protect
- the reactor core, fuel elements and storage wells.- Inherent design of the . reactor bay- and reactor tank structure.
construction materials, building' layout and fire barriers are all applied to the protection. Instrumentation and-control systems and radiation measurement systems primarily:
.are protected by fire detection and alarm information.
_These systems are important_to safety only for the initial shutdown and -removal of personnel. Protection of other
-equipment and the reactor -bay boundary is accomplished' in part ~ by building design,- but primarily by detection .and
-alarm.
11.2.2 Facility Fire Protection Control. The Reactor Supervisor and the Reactor Committee shall consider the
' impact of maj o r facility modifications and experiment programs on facility fire protection. The University Fire 11-3
{'
SAR 9/84 Marshall will recommend- fire protection requirements .and provide for' inspection and test of fire protection components.
. Activities such as welding, cutting, open finmes, electrical ' loads, or other equipment that effect- fire protection'shall be examined.on a case by case basis by the Reactor Supervisor.
Laboratory staff shall be instructed in fire. response actions and notification of response personnel. A program to familiarize response personnel with laboratory equipment, material. hazards and physical layout is considered the major element for response of emergency response. organizations. 11.3 FIRE SAFETY ASSURANCE-At intervals of two years the fire protection program should be examined actively by the Reactor Supervisor, University Fire Marshall-and Reactor Committee. Evaluations of post inspections, tests or incidents shall be incorporated into an assessment o f. the fire protection evaluation. Recommendations if any should be identified and appropriate actions taken. t 11-4
SAR 9/84
' Chapter 11 Re'ferences
- 1. ~ " Standard for. Fire Protection- Program Criteria for Research-Reactors", ANSI /ANS - 15.17 (1981).
- 2. Nuclear Research' Reactors 1983, National Fire Protection Association. Inc., NFPA 802.
4 1 1 h t I i 11-5 i l!
SAR 9/84 Chapter 12-TRAINING AND CERTIFICATION OF OPERATORS
.This. section describes the program . applied to requalification, and qualification of: persons that are to be certified as operators by the licensing authorities [ 1,' 2 ] .
Some features -. o f the program are indicative of the educational nature of the' University institution and the 111mited size of the research reactor staff.
'12.1 TRAINING SUBJECTS Instruction will be given for subj ect matter at the level representative. of either- an operator or senior operator in the f ollowing subj ect areas:
(a) Nuclear Theory and Principle of Operation, (b) Design and Operating Characteristics, (c) Facility Instrumentation and Control Systems. (d) Facility Safety Systems and Engineered Safety Features, (e). Normal. Abnormal,.and Emergency. Procedures,
-(f) Radiation Control and. Safety, (g) Technical Specifications and Basis.
. Instruction shall occur over a two year period and will-consist 'of sessions scheduled for lectures, discussion or self-study. The program goal will be to cover one topic of
, the previously outlined subjects each: calendar quarter. At least f our ' subj ects shall' be covered each alternate year which cllows for the repeat of-one subject in a subsequent year. Lectures may consist of facility class- presentations that are scheduled 'or designated classes of organized university courses.
~
Discussion or self-study sessions may
- replace ' formal lectures when three or fewer persons are participating in the qualification or requalification program.
4 12.2 ' TRAINING EXPERIENCE Each operator or senior operator shall perform ten
-reactor startups or other significant reactivity c manipulations of the reactor during the term prior to a new 1 license or renewal license request. Experience for a new license shall include sufficient additional operating 12-1 t
,na t,
SAR 9/84 7 experience L to provide the trainee with a proficient . skill' and knowledge ' of the reactor operation' for the. type of license . . to be _- requested, j The experience requirement for a
- license . renewal ~of a senior operator shall allow directly y : supervised activities to substitute.for direct performance.
6 [ An annual review by. each operator of abnormal .and emergency procedures- shall occur. Changes . in. design procedures and licenses .or _ technical specifications shall also be reviewed.in a timely manner.
-12.3 EVALUATION Knowledge of an operator-or senior operator shall be evaluated by an annual examination over the material of that j
years _ training program. Competency of an operator or. senior operator 'shall be evaluated by annual observations .of a supervisor or management. . Oral questions to evaluate both knowledge and competency may also supplement the' evaluation. A written examination shall be administered to all operators or. trainees and shall cover the subj ec t s of. the training program for.that year. Each subject will be graded-1; 'on .a 100 point basis with an average of 80% as f the
. acceptance criteria. A n' overall score. of less than. 65%
shall require .an immediate - evaluation - of continued license duties. Proficiency by retraining -shall be demonstrated-F 'within 4 months or license duties shall be suspended until-proficiency is demonstrated. A person that scores between
- 7 y 6 5 %--8 0 % ' shall retrain as necessary in - those areas '
h_ demonstrated: by oral or written exams. The- person! who [ writes, grades, and~ administers the exam shall be a senior ! t-operator and shall not .be required to take the . exam. . An' 1 j accelerated program of less than two years duration will>be j applied to training of persons for new -r operator L certification. 1 n The reactor supervisor-shall periodically evaluate-the i i . performance and competency of each certified . operator. ! i; Evaluation shall include the individuals review of design, procedure and license changes, a review of abnormal and j emergency procedures, manipulation of reactor controls, _ t awareness of laboratory conditions and log 'or '
. entries. =A check 14st' competency evaluation shall be performed annually or as required by operator inactivity of periods
. greater.than four months.
D, Inactivity shall be considered an 1 absence from-all activities of reactor operation. L i L
- j. 12'.4 RECORDS '
- r
' Documents shall be - maintained that record results of qualifications M activities [3,4]. Records , shall include [ 12-2
- j. . \;
SAR 9/84 tests, scores,. qualification activitier such as startups or equivalent reactivity changes, performance evaluation, review of design, procedure and license changes, and: review of abnormal and emergency _ procedures. i t 12-3
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Chapter 12 Referencen ,, , ,
, )
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-1. " Standard for Admin $.strative Control's" srs
- ANSI /ANS - 15.18 x
1979.- - - _ g s
- 2. " Selection and Training pf,' Personnel for Research Reactors", ANS I/ ANS: - 15'.4 1970 (N380). i
, 3.. " Review of Experiments)o'rR'eseaychReactors", ANSI /ANS 15 . 6 -- 1974 (N401). ,m
- 4. " Records'and Re
.15. 1974 '(f) 9ports'for 9 ) . .. ,
Research Reactors", ANSI /ANS -
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> ~5 SAR 9/84 Chapter 13 STARTUP PROGRAM Startup and testing of the Balcones Research Center TRIGA facility shall be performed by personnel of The University of Texas with consultation of the reactor manufacturer GA Technologies. The University of Texas has accumulated more than 20 years operation experience with a TRIGA reactor prior to the new facility proposal. More than twenty TRIGA type reactors, eleven in the U.S., with power levels of one megawatt or more have been produced by CA Technologies.
Training of university personnel associated with startup activities at the new facility is expected to consist of the relicensing of at least two licensed operators from the current facility that have certified senior operator permits. Training of an additional operator or retraining of a current operator by GA Technologies should occur to provide an effective transfer of the manufacturer's experience to the owner-operator. One or more of the certified operators shall have a bachelors or advanced degree in a field of engineering. The startup program is to consist of five phases beginning with the storage of nuclear fuel on site to the reporting of observed reactor parameters. At each phase written procedures, check liste or other documents shall be developed for activities or measurements that will have significant importance to safety or operation. Documentation shall include information required by the various programs to be implemented at the facility such as operator qualifications, radiological protection, fire protection and quality assurance plus operating procedures and other requirements of license authorizations. The startup program is to be divided into the following phases: a.) Storage of fuel and acquisition of componenti, b.) Tests of systems before core loading, c.) Fuel loading and core criticality, d.) Tests subsequent to core criticality and e.) Acceptance of core operation. 13.1 STORAGE OF FUEL AND ACQUISITION OF COMPONENTS Provisions for the storage fuel and components for the reactor facility at the completion of the facility. construction shall require the limited implementation of 13-1
p 1 a 4 y SAR 9/84
/
administrative controls. A license authorization for the possession and storage of special nuclear materials and other radioactive components, such as the cobalt-60 irradiator will be'obtained and materials relocated to the facility. Storage of non-radioactive components, storage of other reactor components and instrumentation, and assembly ' of facility systems will be performed in the initial startup a phase. 13.2 TESTS OF SYSTEMS BEFORE CORE LOADING Facility systems, auxiliary systems and reactor systems or physical parameters shall be tested for the appropriate operating conditions prior to fuel transfer into the reactor core. Yuel may be loaded into the pool during this phase.
- Systems "<shall be tested according to designated spe cifica r tons , when applicable, and acceptable operation shall be ' established before core loading proceeds. Facility systems to be', tested should include security, fire, communication and ventilation systems. Auxiliary systems to be tested should include radiation monitoring, pool coolant, alarm au'd - in t e r lo c k systems. Reactor systems to be tested will include the instrument and control system and verification of physical specifications for assembly and operation of reactor components. Some systems or components that do not meet specifications and are not required for operation may be deferred for acceptance to a later startup program phase.
13.3 CORE LOAD FOR INITIAL CRITICALITY Continuous operation of coolant system, insertion of the neutron source, installation of the cobalt-60 trradiator, and mo v e me'n t sf Zu-1 into 'the core will begin the core load startup program t hase. Certain verifications of instrumentation and control sys' tem functions will be completed before initialization of an approach to critical experimant by standard reciprocal source multiplication f actor measurements. Rod worth values shall be estimated
- from the core loading procedureu.
i e 13.4 TESTS SUBSEQUENT TO CORE _. CRITICALITY i Rod calibration shall be determined by . positive period t measurements before reactor op'eration at power levels effected by the power coefficient. Next an intermediate power-calibration shall be made and evaluation of-the fuel tempera.ture as measured by an instrumented fuel element. ; Last, Ehe fuel loading of the core should oc adj uu t ed for i f ull .p'over operation, and operation of the cooling system , verified at , power. Any variation of core parameters ' 13-2 i
SAR 9/84-
-significantly. dif f erent than predicted by calculations or experience shall' be ' resolved during this startup program phase.
,13.5 ACCEPTANCE FOR. OPERATION
.The - final startup . program phase shall consist . of the
< resolution of all deviations from specifications.
Deviations should be resolved as specified- for quality a s s u r a n c e .. or other methods determined to be acceptable.
'Three months- _after completion of requisite initial startup and power-escalation testing of _the-reactor, or.nine months after initial criticality. a written report shall be submitted to licensing authorities. The report .shall
' include a. summary of the following:
(a) Description of measured values of operating conditions or characteristics obtained and comparison of these values with design predictions
.or specifications.
'(b) Description of major corrective actions taken to obtain satisfactory operation.
(c)-Re-evaluation of safety analysis where measured values indicate substantial variance from those values used in the Safety Analysis Report. Results of the startup program shall become a supplement to this' chapter-of The* University of Texas-TRIGA Safety Analysis Report. h e I f 13-3
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