ML19341C514

RSAR for Univ of Tx Triga Mark I

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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v TRIGA MARK I i

SAFETY ANALYSIS REPORT Revised January 1981 e 1 i

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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 f

. . . . . . . . . . . . . . 3-41 3.1.3.2 AP , Exit Loss. . . . . . . . . . . . . . . . 3-41 e

3.1.3.3 AP g , Loss Through Portion of Channel 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

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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. REACTOR COOLANT SYSTEM . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.1 Design Bases. . . . . . . . . . . . . . . . . . . . . . . . . . 4-2 4.2 System Design . . . . . . . . . . . . . . . . . . . . . . . . . 4-2 4.2.1. System Operation . . . . . , . . . . . . . . . . . . s 4-2 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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1 LIST OF FIGURES 2-1 Texas . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-2 2-2 Travis County . . . . . . . ... . . . . . . . . . . . . . . . 2-3

- 2-3 Downtown Austin . . . . . . . . . . . . . . . . . . . . . . . . '4 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 element after pulse, h ,= 375 . . . . . . . . . . . . . . 3-20 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 average spectra in.TRIGA ZrH fuel element for 23 C fuel . ..... . . . Y . . . . . . . . . . and..400'C ... . 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 4

fuel temperature . . . . . . . . . . . . . . . . .. ... 3-38 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 versus power . . . . . . . . . . . . . . . . . . ..... 3-60 3-25 Typical Mark I TRIGA reactor. . . . . . . . . . . . . .... . 3-65

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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

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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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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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