ML19341C514
| ML19341C514 | |
| Person / Time | |
|---|---|
| Site: | 05000192 |
| Issue date: | 01/31/1981 |
| From: | TEXAS, UNIV. OF, AUSTIN, TX |
| To: | |
| References | |
| NUDOCS 8103030673 | |
| Download: ML19341C514 (204) | |
Text
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TRIGA MARK I i
SAFETY ANALYSIS REPORT Revised January 1981 e
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TABLE 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 2.1 Ceneral Location 2-1 2.2 Surrounding Activities and Population.
2-1 2.3 Climatology.
2-7 2.4 Geology.
2-11 2.5 Hydrology.
2-18 2.6 Seismology
. 2-20 References.
2-22 3.
REACTOR
. 3-1 e
3.1 Design Bases
. 3-1 s
3.1.1.
Reactor Fuel Temperature.
. 3-2 3.1. 2. P rompt Nega**ve Temperature Coefficient 3-30 3.1.2.1 Codes Used for Calculations.
. 3-33 3.1.2.2 ZrH Model.
. 3-33 3.1.2.3 Calculations
. 3-36 3.1. 3. Steady-State Reactor Power.
. 3-37 3.1.3.1 AP, Entrance Loss 3-41 f
3.1.3.2 AP, Exit Loss.
. 3-41 e
3.1.3.3 AP, Loss Through Portion of Channel g
Adjacent to Lower Reflector.
3-42 J.1. 3. 4 AP Loss Through Portion of Channel Adjacent to Upper Ref2cctor.
. 3-42 3.1.3.5 AP, Loss Through Each Increnent of the Channel Adjacent to the Fueled Portion of the Elements.
. 3-43 i
m
3.1.3.6 Acceleration Term.
3-4,
3.1.3.7 Friction Term.
3-46 3.1.3.8 Gravity Term 3-47 3.2 Nuclear Design and Evaluation.
3-56 d
3.2.1. Reactivity Ef fects.
3-56 3.2.2. Evaluation of Nuclear Design.
3-58
- 3. 3 Thermal and Hydraulic Design.
3-62 3.3.1.
Design Bases.
3-62 3.3.2.
Thermal and Hydraulic Design Evaluation.
3-63 3.4 Mechanical De 11gn and Evaluation.
3-64 3.4.1.
General Description.
3-64 3.4.2.
Experimental and Irradiation Facilities.
3-66 3.4.3.
Reflector Assembly and Grid Plates.
3-67 3.4.4.
Fuel-Moderator Elements.
3-70
- 3. 4. 4.1 Evaluation of Fuel Element Design.
3-73 3.4.5.
Graphite Dummy Elements.
3-75 3.4.6.
Neutron Source and Holder.
3-75 3.4.7.
Control System Design.
3-75 3.4.7.1 Control Rod Drive Assemblies.
3-75 3.4. 7. 2 Transient Rod Drive Assembly.
3-79 3.4.7.3 Evaluation of Control Rod System.
3-82 3.5 Safety Settings in Relation to Safety Limits.
3-83 Re fe rences.
3-84 4.
4-1 4.1 Design Bases.
4-2 4.2 System Design.
4-2 4.2.1.
System Operation 4-2
.....,............ s 4.2.2.
System Instrumentation.
4-6 4.3 Water System Design Evaluation.
4-7 5.
REACTOR ROOM.
. 5-1 5.1 Design Basis.
5-1 5.2 Room Design.
. 5-1 9
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l 5.3 Room Isolation.
5-5 i
5.4 Room Containment Evaluation.
5-8 5.4.1 Release of Argon-41 and Nitrogen-16 from Pool Water.
5-8 5.4.1.1 Argon-41 Activity in Reactor "com.
5-8 5.4.1.2 Nitrogen-16 Activity in Reactor Room.
5-17 5.4.2 Activation of Air in the Experimental Facilities.
5-21 References.
5-24 6.
INSTRUMENTATION AND CONTROL.
6-1 6.1 Ceneral Description.
6-1 6.2 Nucicar Channels.
6-1 6.3 Temperature Channel and Water Monitor Channel.
6-5 6.4 Reactor Operating Modes.
6-5 6.4.1 Manual Mode.
6-o 6.4.2 Automatic Mode.
6-8 6.4.3 Pulse Mode.
6-9 6.4.3.1 Monitoring Channels.
6-9 6.5 Reactor Control System.
6-10 6.5.1 Manual Rod Control Circuit.
6-10 6.5.2 Automatic Operation.
6-12 6.5.3 Pulsing Operation.
6-13 6.6 Reactor Safety.
6-14 7.
RADIOACTIVE WASTES AND RADIATION PROTECTION.
7-1 7.1 Radioactive Waste.
7-1 7.1.1 Design Bases.
7-1 7.1.2 Evaluation.
7-1 7.2 Radiation Protection 7-2 7.2.1 Shiciding.
7-2 7.2.1.1 Design Bases 7-2 7.2.1.2 Evaluation of Shielding.
7-2 7.2.2 Area Monitoring.
7-2 7.2.2.1 Design Bases 7-2 7.2.2.2 Evaluation of Area Monitoring.
7-3 iii
7.2.3 Personnel Monitoring.
7-3
- 7. 2. 3.1 Design Bases.
7-3 7.2.3.2 Evaluation of Personnel Monitoring.
7-3 8.
SAFETY ANALYSIS.
8-1 8.1 Fission Product Release.
8-1 8.1.1 Fission Product Invent o ry.
8-1 8.1.2 Fission Product Release Fractio *.e..
8-1 8.1.3 Downwind Dose Calculations.
8-4 8.1.4 Downwind Doses.
8-6 8.2 Loss of Reactor Coolant 8-8 8.2.1 Summa ry.
8-8 8.2.2 Fuel Temperature and Clad Integrity.
8-12 8.2.3 After Heat Removal Following Coolant Loss.
8-16 8.2.4 Radiation Levels.
8-21 8.3 Reactivity Accident 8-22 Re fe rences.
8-27 9.
ORGANIZATION.
9-1 9.1 Administrative Structure.
9-1 9.1.1 Nuclear Engineering Teaching Laboratory.
9-1 9.1.2 Radiation Safety Committee.
9-1 9.1.3 Radiation Safety Officer.
9-1 9.1.4 Reactor Committee.
9-3 9.1.5 Laboratory or Reactor Supervisor.
9-3 9.2 Operating Requirements.
9-3 9.2.1 Licenses 9-3 9.2.2 Procedures.
9-4 9.2.3 Other Requirements 9-4 iv
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LIST OF FIGURES 2-1 Texas 2-2 2-2 Travis County.
2-3
'4 2-3 Downtown Austin.
2-4 The University of Texas at Austin Campus.
2-5 2-5 Taylor Hall; First Floor Layout 2-6 2-6 Travis County Census Tracts.
2-9 2-7 Austin Census Tracts.
2-10 2-8 Wind Rose-Austin, Texas 1951-1960.
2-15
?-9 Texas Annual Tornado Density.
2-16 2-10 Tornado and Funnel Cloud Occurences Surrounding Austin.
2-17 2-11 Texas Earthquakes.
2-21 3-1 Phase diagram of the zirconium-hydrogen system.
3-3 3-2 Equilibrium hydrogen pressures over ZrH versus temperature.
3-6 3-3 Strength of type 304 stainless steel as a function of temperature 3-7 3-4 Strength and applied stress resulting from equilibrium hydrogen dissociation pressure as a function of temperature 3-9 3-5 Radial poser distribution in U-ZrH fuel element 3-10 3-6 Axial power distribution in a fuel element assumed for thermal analysis 3-12 3-7 Subcooled boiling heat transfer for water.
3-13 3-8 Clad temperature at midpoint of well-bonded fuel element.
3-14 3-9 Fuel body temperatures at midplane of well-bonded U-ZrH element after pulse.
3-15 3-10 Surface heat flux at midplane of well-bonded U-ZrH element af ter pulse.
3-16 3-11 Surface heat flux distribution for standard non-gapped fuel element after pulse, h
= 500 3-19 3-12 Surface heat flux distribution for standard non-gapped fuel 3-20 element after pulse, h
,= 375 3-13 Surface heat flux distribution for standard non-gapped fuel cicnent after pulse, h
= 250 3-21 3-14 Surface heat flux at midplane versus time af ter pulse for standard non-gapped fuel element 3-22 I
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h 3-15 Transport cross section for hydrogen in zirconium hydride und
. Y
.......... and 400'C average spectra in.TRIGA ZrH fuel element for 23 C fuel.
3-32 3-16 A comparison of theoretical and experinental neutron spectra in H,0 using free hydrogen and bound hydrogen models for the
calc ula tion.
3-34 3-17 Experimental and theoretical neutron spectra from ZrH 1 75 showing the ef fect of temperature variation.
3-35 3-18 TRICA prompt negative temperature coefficient versus average fuel temperature.
3-38 4
3-19 General system configuration.
3-40 3-20 Experimentally determined vapor volumes for subcooled boiling in a narrow vertical annulus.
3-44 3-21 Cross-plot of Fig. 3-20 for use in calculations.
3-45 3-22 Miximum heat flux for which DNS ratio is 1.0 versus coolant tempe ratu re.
3-53 3-23 Estimated reactivity loss versus power.
3-59 3-24 Estimated maximum B ring and average core temperature 3-60 versus power.
3-25 Typical Mark I TRIGA reactor.
3-65 3-26 Reactor Tank and Shield Structure 3-67 l
=
3-27 Typical core diagram.
3-69 3-28 TRIGA stainless steel clad fuel element with triflute end fittings.
3-71 3-29 Instrumented fuel element 3-72 3-30 Control rod poison container.
3-76 3-31 Rack-and-pinion control rod dri e.
3-78 3-32a Adjustabic transient rod drive operational schematic..
3-80 3-32b Adjustable transient rod drive.
3-81 4-1 Primary cooling system.
4-3 4-2 Water Purification system.
4-5 5-1 Taylor Hall 131 Floor Plan.
5-3 5-2 Reactor Tank and Shield Structure 5-4 5-3 Taylor Hall Floor Plan adjacent rooms to 131.
5-6 6-1 TRIGA control console, front view.
6-2 6-2 Functional block diagram.
6-3
~
6-3 Operating ranges of TRIGA Mark I pulsing reactor neutron detectors.
6-7 6-4 Control Panel.
6-11 a
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1 8-1 Maximic fuel temperature versus power density af ter loss of coolant for various cooling times between reactor shutdown and coolant loss 8-9 8-2 Strength and applied stress as a function of temperature.
U-Z ril ue u 1 and clad at same temperature.
8-10 l.65 8-3 Cooling times af ter reactor shutdown necessary to limit maximum fuel temperature versus power density.
8-11 9-1 Administrative Structure of Nuclear Engineering Teaching Laboratory.
9-2 m
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LIST OF TAP.LES 1-1 Principal design parameters.
1-2 2-1 1976 Population Density Distribution - Travis County.
2-8 2-2 Meteorological Data.
2-12 2-3 Meteorological Data.
2-13 2-4 Meteorological Data.
2-14 3-1 Physical properties of delta phase U-Zrli.
3-4 3-2 Hydraulic flow parameters.
3-39 3-3 Typical TRIGA core nuclear parameters.
3-57 3-4 Expected reactivity effects associated with experimental facilities.
3-57 3-5 Estimated control rod not worth..
3-57 3-6 Estimated fuel element reactivity worth compared with water as a function of position in core.
3-58 3-7 Comparison of reactivity insertion effects.
3-61 3-8 250 kW(t) TRIGA heat transfer and hydraulic parameters.
3-64 3-9 Thermocouple specifications.
3-74 3-10 Summary of fuel element specifications.
3-74 3-11 Summary of control rod design parameters.
3-77 3-12 TRICA safety settings.
3-83 4-1 Reactor coolant eystem design summary..
4-4 5-1 Saturated argon concentration in water.
5-10 5-2 Volumes and Thermal Fluxes of Facilities.
5-22 8-1 Noble Gas and Halogens in the Reactor.
8-2 8-2 Assumed Breathing Rates.
8-6 8-3 Average gamma-ray energy and internal dose effectivity for each fission product isotope.
8-7 8-4 Downwind doses from fission product release.
8-8 8-5 Calculateo Radiation Dose Rates for Loss cf i
Reactor Pool Water.
8-21 8-6 Reactivity Transient Input Parameters..
8-23 i
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