ML20065K721

SAR for Univ of Wisconsin Nuclear Reactor

ML20065K721
Person / Time
Site:	University of Wisconsin
Issue date:	04/30/1973
From:	Cashwell R WISCONSIN, UNIV. OF, MADISON, WI
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ML20065K719	List: ML20065K712 ML20065K721
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NUDOCS 9012040113
Download: ML20065K721 (174)

Text

___ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ .

's SAFETY ANALYSIS REPORT FOR THE UNIVERSITY OF WISCONSIN NUCLEAR REACTOR .,

SUBMITTED TO THE UNITED STATES ATOMIC ENERGY COMMISSION Prepared by R. J. Cashwell Department of Nuclear Engineering College of Engineering The University of Wisconsin Madison, Wisconsin 53706 April 1973 90120ao1i c 901120 pen ADOCl . O'5000 L 5j p F Di.,

l INDEX Page CHAPTER 1

SUMMARY

AND PROPOSED TECHNICAL SPECIFICATIONS 1-1

1.1 INTRODUCTION

1-1 LISTING OF CIMNGES TO CHAPTER 1 1-2 CHAPTER 2 1-2 CHAPTER 3, 4, 5, 6 1-3 1.2 GENERAL DESCRIPTION 1-4 1.3

SUMMARY

OF REACTOR DATA 1-5 Experimental Facilities 1-5 Reactor Materials 1-5 Dimensions 1-5 Predicted Nuclear Characteristics 1-5 lh PROPOSED TECHNICAL SPECIFICATIONS 1f6 INDEX IN SPECIFIC h5FCTION N8 epa Unit) dgg mw gNcr es ws "

- _ - - - . ..-__.-___----.__-___-__-..--_--___--____---._-_--_____._--_--_-_---J

Page C HAPTER 2 FACILITY DESCRIFTION 2-1 I

2.1 REACTOR CORE 2-1 2.1.1 Core Support 2 - 2 .

2.1.2 Grid Box 2-3 2.1.3 Fuel 2-3 2.1.4 Reflectors 2-9 2.1.5 Safety Blade 2-9 2.1.6 Regulating Blade 2-9 2.1.7 Transient Control Rod 2 - 14 2.1.8 Neutron Source 2 - 14 2.1.9 Core Arrangement 2 - 14 2.2 DRIVE MECHANISMS 2 - 16 2.1.1 Safety Blade Drive 2 - 16 2.2.2 Regulating Blade Drive 2 - 20 2.2.3 Fission Counter Drive 2 - 20 2.2.4 Transient-Rod Drive 2 - 20 2.3 COOLING AND LIQUID WASTE SYSTEMS 2 - 24 2.3.1 Pool 2 - 24 2.3.2 Cooling Sys te-' 2 - 27 2.3.3 Pool Make-up and Clean-up System 2 - 29 2.3.4 Waste Disposal System 2 - 32 2.4 EXPERIMENTAL FACILITIES 2 - 34 2.4.1 Thermal Column 2 - 34 2.4.2 Beam Ports 2 - 36 2.4.3 Thermal Column and Beam Port Ventilation System 2 - 38 2.4.4 Pneumatic Tube 2 - 41 2.4.5 Grid Box Irradiation Facilities 2 - 41 2.5 CONTROL AND INSTRUMENTATION 2 - 43 2.5.1 Steady-Starn Operation 2 - 44 2.5.2 Square Wave Operation 2 - 46 2.5.3 Pulsing Operation 2 - 47 2.5.4 Blade Contro] 2 - 47

q. e i

Page 2.5.5 Automatic Control System 2 - 49 2.5.6 Transient Rod Control 2 - 49 2.5.7 Scram circuits 2 - 49 2.5.8 Alarm and Indicator System 2 - 51 2.5.9 Radiation Monitors 2- 53 2.6 SHIELDING AND EXPECTED RADIATION LEVELS 2 - 55 2.6.1 Basic Reactor Shielding 2 - 55 2.6.2 Pool Surface Radiation Levels-N 16 Activity 2 - 56 2.6.3 Demineralizer 2 - 57 2.6.4 Heating Effects in Shield and Th cmal Column 2 - 57 2.7 FUEL ELEMENT MEASUREMENTS 2 - 59 CHAPTER 3 LOCATION AND BUILDING 3-1 3.1 LOCATION 3-1 3.1.1 Site Description 3-1 3.1.2 Topography 3 - 13 3.1.3 Geology and Hydrology 3 - 13 3.1.4 Water Supply 3 - 15 3.1.5 Seismology 3 - 17 3.1.6 Climatology 3 - 17 3.1.7 Meteorology 3 - 19 3.2 BUILDING 3 - 26 3.2.1 . General Description 3 3.2.2 Heating and Air conditioning- 3 - 31 3.2.3 Ventilation 3 - 32 CHAPTER 4 PROTOTYPE PERFORMANCE CHARACTERISTICS AND REACTOR PARAMETERS 4-1

4.1 INTRODUCTION

TO PROTOTYPE TESTS 4-1 4.2 REACTIVITY CHANGES DUE TO REFLECTOR VARIATIONS 4-3 4.3 FUEL ELEMENT WORTHS 4-3

e- .o l

Page 1 4.4 STEADY STATE PARAMETERS 4-6 1 4.4.1 Critical Loading- 4-6 4.4.2 Power Coefficient 4-6 j 4.4.3 Fuel Temperatures 4-7 4.4.4 Isothermal Temperature Coefficient (Bath Coefficient) - 10 4.5 PULSE PARAMETERS 4 - 10 4.5.1 Period 4 - 10 4.5.2 Pulse Width 4 - 13 4.5.3 Peak Power 4 4. 6 - FLIP FUEL PERFORMANCE CHARACTERISTICS 4 - 18 4.7 MIXED FLIP-STANDARD FUEL CORE 4 - 18 CHAPTER 5 ORGANIZATION AND PROCEDURES 5-1 5.1 OPERATING ORGANIZATION 5-1 5.1.1 University Radiation Safety Committee 5-1 5.1.2 University Health Physics Office 5-1 5.1.3 Reactor Director 5-1 5.1.4 Reactor Safety Committee 5-3 5.1.5 Reactor Supervisor 5 5.1.6 Senior Operators (Alternate Supervisors)- 5-5 5.1.7 Reactor Operators- 5-5 5.2 OPERATING STANDARDS- 5-6 5.2.1 Limithtion of Experiment Reactivity Worth 5-6

~5.2.2 Operations Which Might Involve Changes in Core Reactivity Conducted When the Reactor is Shut Down 5-6

. f5.2.3 Shut-Down Margin 5 -6 5.3 OPERATIONAL PROCEDURES 5-7 5.3.1 Initial Test Program 5-7 5.3.2 - Routine Start-Up Operation 5-8 5.3.3 Re fueling 5-9 l.

5. . .

Page 5.4 EXPE. MENTS 5 - 12 FUELED EXPERIMENTS 5 - 13 CONVERTER PLATES 5 13 CHAPTER 6 SAFEGUARDS EVALUATION, 6-1 6.1 GENERAL 6-1 6.2 PRODUCTION AND RELEASE OF GASEOUS RADIOACTIVITY 6-2 6.3 SPILLAGE OF RADIOACTIVE . MATERIALS 6-3 6.4 REACTIVITY ACCIDENT-Ejection of the Transient Rod while at Maximum St.eady-State Power. 6-5 6.4.1 Fuel Tempere cures from Operation at the Serr.m Point 6-6 6.4.2 Temperature a f ter Pulse 6-8 6.5 FUEL ELEMENT CLADDING FAILURE 6 -

9 6.5.1 Fission Product Inventory in Fuel Element 6-9 6.5.2 Fission Product Release Fraction 6-9 6.5.3 Activity in Pool Water 6 - 12 6.5.4 Fission Product Release to Air Within the Reactor Laboratory 6 - 12 6.5.5 Release of Fission Products to Unrestricted Areas 6 - 14 APPENDICES TO CHAPTER 6 EMERGENCY PROCEDURES 1 REACTOR ACCIDENT FISSION PRODUCT RELEASE OR MAJOR SPILI OF RADIOACTIVE MATERIAL UWNR-150 LEAK RESULTING IN DRAINING OF POOL UWNR-151 SUSPECTED FUEL ELEMENT CLADDING LEAK UWNR-152 CALCULATIONAL METHODS FOR ATMOSPHERIC RELEASE OF RADIOACTIVITY A- 1

e. :n FIGURES Page CHAPTER 1 Figure 1 Open Pool Reactor 1-7 CHAPTER 2

-Figure 1 Core Suspension- 2-2 Figure 2 Triga Fuel Elements 2-4 Figure 3 Sectional. Views of TRIGA Standard 4-Rod Cluster 2-5 Figure 4 Fuel Element' 2-7 Figure 5 Instrumented Fuel Element 2-8 Figure 6 Reflector Element 2 - 10 Figure'7 Safety Blade 2 - 11 Figure 8 Shroud Assembly 2 - 12 Figure 9 Regulating Blade 2 - 13 Figure 10 Transient Control Rod 2 - 15 Figure 11 Grid Arrangement 2 - 17 Figure 12 Drive Mechanisms 2 - 18

' Figure 13 Transient-rod Drive Assembly 2 - 21 Figure 14 Pneuma tic-elec tromechanical Transient-rod Drive 2 - 22 y Figure 15 Reactor Pool 2 - 26 Figure'16' Cooling' System - 1 MW L

Operation 2 - 28 L Figure-17 Pool Makeup and Cleanup l System 2 - 31 1

! Figure 18 Waste Disposal System 2 --33 Figure 19 Thermal Column 2 - 35 Figure 20 Beamport 2 - 37

-Figure 121 Beam Port and Thermal Column Ventilation System 2 - 39 l Figure 22 UWNR Pneumatic Tube System 2 - 40 Figure 23 Nuclear Instrumentation 2 - 45 Figure 24 Fuel Measurement and Maintenance-Tool' 2 - 58 Figure 25 Bow and Elongation Sensors 2 - 60

e. . )

l FIGURES Page Figure 26 Puel Element Measurement Data Sheet 2 - 63 CIIAPTER 3 l Figure 1 Topography 3-2 Figure 2 City of Madison 3-3 i

Figure 3 University of Wisconsin Campus 3-4 Figure 4 Reactor Laboratory and Mechanical Engineering Bldg. 3-5 Figure 5 Reactor Laboratory and Basement Mechanical Engineer-ing Building 3-6 Figure 6 Reactor Laboratory and First Floor, Mechanical Engineering Building 3-7 Figure 7 Second Floor, Mechanical Engineering Building 3-8 Figure 8 Third Floor, Mechanical Engineering Building 3-9 Figure 9 Basement Engineering Research Building 3 - 10 Figure 10 First Floor, Engineering Research Building 3 - 11 Figure 11 Table of Population Distribution 3 - 12 Figure 12 Geology 3 - 14 Figure 13 Surface Wind Roses 3 - 22 '

Figure 14 Upper Air Wind Roses 3 - 24 3 - 25

' Figure 15 Basement Floor Level Layout 3 - 27 Figure 16 First Floor Layout 3 - 28 Figure 17 Laboratory Location South from q of Core 3 - 29 Figure 18 Reactor Laboratory Looking North 3 - 30

Pace Figure 19 . _ Reactor Laboratory Looking West 3 - 34 Figure 20 Reactor Laboratory Looking East 3 - 35 Figure 21 Ventilation System 3 - 36 CHAPTER 4 Figure 1 Fuel Element Bundle Worth 4 Figure 2 Fuel Loading Diagram-Prototype Reactor 4-5 Figure 3 Reactivity Loss vs. Reactor Power-Prototype Reactor 4-8 Figure 4 Measured Fuel Temperature Above Ambient vs. Reactor Power-Prototype Reactor 4-9 Figure 5 Reactivity vs. Pool Temper-ature-Prototype Reactor 4 - 11 Figure 6 Inverse Period and Inverse Width at Half Maximum Power vs. Prompt Reactivity Insertion-Prototype Reactor 4 - 12 Figure 7 Peak Power vs. Inverse Half-Width Squared 4 - 14 Figure 8 Full Width at Half-Peak Power vs. Period 4 - 15 Figure 9 Peak Power vs. Full Width at Half Peak Power 4 - 16 Figure 10 Peak Power vs. Reactor Period 4 - CHAPTER-5 Figure 1 Organization Chart 5-2 CHAPTER 6.

, Figure l-  % of Outside Fuel Radius 6-7 L

1 L

LIST OF TABLES Page CliAPTER 3 Table 1 Average Temperature for Madison 3 - 19 Table 2 Normal Precipitation for Madison 3 - 20 Table-3 Snow, Sleet, and Hail for Madison 3 Table 4 Miscellaneous weather Data for Madison 3 - 21 CHAPTER 4 Table 1 Triga Mark III Prototype Reactor Fuel Moderator Specifications 4-2 Table 2 Reactivity Changes Associated With Reflector Changes 4-3 Table 3 Power Peaking and Fuel Bundle.

Worth in Mixed Flip-Standard TRIGA Cores 4 - 21 C,HAPTER 6 Table 1 Fission Product Yield and Release Potential 6 - 10, 11 Table 2 Iodine-Thyroid-Dose Information 6 - 14

e .

l l Chapter 1

SUMMARY

AND PROPOSED TECIINICAL SPECIFICATIONS

1.1 INTRODUCTION

The University of Wisconsin Nuclear Reactor achieved initial criticality on 26 March 1961 as a 10 KW teaching and research reactor. The power level was increased to 250 KW on 7 December 1964, using the original flat-plate aluminum clad fuel.

The reactor was converted to a 1000 KW, TRIGA reactor with pulsing capability in 1967. Initial criti-cality with the TRIGA core occurred on 14 November 1967.

Since that time the reactor has been operated for more than 3 million kilowatt hours. At the present time a partial refuelling is n2cessary to provide adequate reactivity for the operating program.

This and subsequent refuellings will use TRIGA FLIP fuel containing erbium oxide burnable poison and a higher uranium 235 enrichment which will provide extended fuel lifetime.

Since submission of the Safety Analysis Report for operation with TRIGA fuel in July of 1966, a number of 4 changes in the facility have occurred. These changes were reported to the Commission in Annual Reports required under 20 CFR Part 50. In view of there changes and re-evaluation of safety analyses required by use of FLIP fuel, this document is submitted as a replacement for the entire Safety Analysis Report even though the present application is intended as a request for changes in Technical Specifications.

The remainder of this introduction section points out changes f rom the previous SAR. In addition to the changes indicated below, there were several places wher e references to 1.5 MW operation were omitted or where minor wording changes were made.

1- 1

_ _ _ - - - _ _ _ _ _ _ _ _ . _ . . . ._..___.________.__m_ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ __

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Chapt r 1. Introduction Section and Proposed Technical Specifications--Extensively revised.

General' Description and Summary of Reactor Data--

Changes made to reflect use of FLIP fuel.

Proposed Technical Specifications are completely revised to conform to new format and FLIP fuel addition.

Chapter 2.

Section 2.1.3 Modified to include description of FLIP fuel.

Section 2.1.9 Changes to reflect restrictiot.s on core arrangement with-FLIP fuel.

Section 2.3.1 Changed to indicate use_ of three fuel storage baskets rather than previous one-piece storage rack.

Reference to pH control of pool water removed as previously reported. -

Section 2.4.1 and Figure 19--Changed to show location of safety channel-CIC's inside thermal column as previously reported.

Section 2.4.4 Changed to show use of CO V*#

2 gas rather than helium.

Section 2.4.5 Changed to describe-hydraulic sample irradiation devices previously reported.

Section 2.5.1 Changed to include scram on fuel temperature. This is a change in support of new Technical Specif; cations.

Section 2.5.5 Includes description of servo control system-changes previously reported.

Section 2.5.7 Includes fuel temperature scram at LSSS.

Section 2.6.2 Changed to include installation of stairway from console area to to pool top as previously reported.

1-2

e .

Section 2.7 Changed to make : 'rences to fuel element bow and elongation consistent with proposed Technical Specification Limits.

Chapter 3.

This chapter is extensively rewritten to reflect remodelling in the Reactor Laboratory and surrounding areas and to reflect construction of the Engineering Research Building east of the Mechanical Engineering Building as previously reported.

Chapter 4.

This chapter has been revised to show UWNR data on curves of prototype behavior, and to indicate behavior expected with FLIP and mixed cores.

Chapter 5.

Chapter 5 is extensively revised to show (a) Procedures to be used in assembly of operational core using mixed FLIP and Standard fuel, and-(b) To indicate a change in review and audit functions as presented in the proposed Technical Specifications.

Chapter 6 3 Chapter 6 has been extensively revised as a result of the higher power density due to peaking in mixed and FLIP cores. -The sections on loss of coolant, pulsing while at full power, anf fission product release have been' revised to reflect the power peaking and more recent values of fission product release fractions in TRIGA fuel.

1-3

1.2 G ENE:- ,L DESCRIPTION Figure 1 is a pictorial view of the reactor. The reactor is a heterogeneous pool-type nuclear reactor fueled. with TRIGA and TRIGA-FLIP fuel modified to adapt to 4-element bundle assemblies. The coolant is light water which circulates _ through the core by natural convection. The core is reflected by water and graphite. Maximum steady-etate power level is 1000 KW, A 7 by 9 grid, _ surrounded by a core box, positions fuel, reflector, and control elements. Three shim-safety blades, a transient control rod, and a regulating blade control core reactivity. The control blades move vertically in two shrouds extending the length of the core. The grid box and control element drive mechanisms are supported by a suspension frame from the reactor bridge.

Cold, clean core excess reactivity is. ab0ut 4.9 per cent reactivity. The safety blades provide a shut-down margin of 6 to 14 per cent 6g eff*

The proposed technical specifications for the facility are listed below as Section 1.4.

1.3

SUMMARY

OF REACTOR DATA Responsible Organization The University of Wisconsin Location Madison, Wisconsin Purpose Teaching and Research Fuel Type TRIGA and'TRIGA-FLIP High Hydride in_4 element clusters Number of elements in standard 1000 KW 99 (25 Fuel Bundles)  ;

Core Control t

Safety elements Three vertical blades Regulati ng-servo One vertical blade element Transient control One rod i

1-4 '

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9 Exoerimental-Facilities Thermal column One, 40-inch square graphite, d th = 2 x-10 8 Beam Ports Four, 6-inchdiamyger d th = 1 - 3 x 30 ny at shield side of shutter; about 8 x 1011 at core end of port Pneumatic Tube One, 2-inch (sample size 1-1/4 inch diameter by 5-1/2 inches long),

d th = 10 12 nv.

Thermal neutron fluxes for isotope production include the above, plus large irradiation spaces outside the core with thermal neutron. fluxes of around 1.3 x 1013 nv.

Reactor' Materials Fuel-moderator element 8. 5 wt-% Uranium, 89. 9 wt-%

Zirconium, 1.6 wt-% Hydrogen, 1.5 wt-% Erbium in FLIP 20% U235 (70% for FLIP)

Cladding 20 mil stainless stee'

.R.. lector Water and Graphite Coolant Light Water Control Doral and stainless steel; Borated graphite for transient rod Structural material- Aluminum l-Shield Concrete and Water Dimensions Pool 8 x 12 x 27-1/2 ft. deep Standard 1000 10V core 15 x 17 x 15 inches high Grid box 9. x 7 array of 3-inch modules l 1-5

2ontrol blades 10-1/2 inches wide Puel Element Diameter 1.41 inches Leng th 30 inches Predicted nuclear characteristics 1 MW Steady state:

Maximum thermal 3 neutron flux 3.2 x 10 ny Maximum fast y3 neutron flux 3.0 x 10 nv 2000 MW Pulse Maximum thermal 16 neutron flux 6.5 x 10 nv Maximum fast 16 neutron flux 6.1 x 10 nv Prompt temperature coefficient of ~

reactivity -1.26 x 10 AK/ C Void coefficient of reactivity .2 x 10~4 AK/% void Prompt neutron lifetime 42 psec STD Fael, 18 psec FLIP Effective delayed neutron fraction 0.007 1.- PROPOSED KCHNICAL SPEChKICATIONS Numerica values given these specifi tions ma differ from casured value due to normal nstruction and .nu f ac turing clerances or rmal accuracy  ?

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Chapter 2 FACILITY DESCRIPTION 2.1 REACTOR CORE 2.1.1 Core Support The core is suspended from an all-aluminum frame, Figure 1, which extends from the grid box to a height of about one foot above the pool surface.

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suspension frame serve as guideg for nue car p ins.tsems ni.a timi- detectory. 7"f- DF MO The reactor bridge (mounted over the pool) supports the core suspension frame. The all-steel, prefabricated bridge was bolted together in the field and aligned with shims.

A locating plate, made of 1/?-inch steel, spans the upper end of the suspension frame.

It is bolted to the bridge and aligns the four control blade drive mechanisms and the transient rod drive with the core. The five mechanisms work through individual clearance holes, each mechanism being secured to tne locating plate.

The plate and mechanisms are not removable as a unit to prevent accidental withdrawal of the control elements. The fission counter drive is-mounted on a portion of the hand railing support structure.

An aluminum coolant header (not shown in Figure 1) mates with the bottom of the grid box and forms a transition to the coolant piping originally provided for future use with a forced convection cooling system. An opening in the side of the header, 24 inches wide by 12 inches high, allows cooling water flow for natural convection. A diffuser pump and jet above the core deflects the cooling water streaming from the ec.a to reduce Nlb activity above the core.

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e 6 2.1.2 Grid Box The core elements are supported and enclosed on four sides by the grid box. The grid box is approximately 28 inches long, 24 inches wide, and 36 inches high. The bottom is an aluminum grid plate with a 9-by-7.

array of square holes, spaced to conform with the basic 3-inch element module. The sides of the grid box direct the convection current of the cooling water through the core. Four corner posts attached to the lower end of the suspension frame support the grid box.

All parts, except for mechanical fasteners, are made of aluminum.

2.1 3 Fuel The fuel is of the TRIGA four-element bundle type developed to provide a simple means of converting reactors using flat-plate fuel to TRIGA reactors. Figure 2 shows a four-element bundle and a three-element bundle.

The four-element bundle consists of bottom adapter, top adapter, and four TRIGA elements.

The bottom adapter fits the existing grid plate as did the original fuel elements.

The end fittings on individual TRIGA elements are threaded into the bottom adapter until a flange on the element seats firmly against the adapter, providing rigid cantilever-type support. (See Figure 3. )

The top adapter serves both as a handling fitting and as a spacer for the upper ends of the fuel elements. A sliding fit between this adapter and the fuel element end fittings allows 3 for dif ferential expansion of the elements.

This top fitting can be removed with remote handling tools to disassemble the bundles for any required measurements or for shipping spent elements for reprocessing.

2- 3 1

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• t 2.1.4 Reflectors The reflectors are standard General Electric: Company reflectors as used in the Washington State University Reactor. The nominal 3-inch reflector elements are made of AGOT grade graphite clad with aluminum (see Figure 6). Reflector element lifting handles are diagonal to facilitate identifi-cation when viewing the core and storage racks.

2.1.5 Safety Blade Reactor control for startup and shut-down is accomplished by three blade-type control elements, Figure 7, with a total shutdown worth betweer. 6.9 and-11 per cent 4K eff.

The poison section is boral aheet (boron carbide and aluminum-sandwiched between aluminum - side plates) . It is 40.5--

inches long, 15 inches providing active

- control of the core and the remaining 25.5 inches connecting the poison section to the drive tube.

Each safety blade is guided throughout its travel by a shroud shown in Figure 8.

The shroud consists of two thin aluminum plates

. -38 inches high, separated by aluminum spacers to provide a 1/8-inch water annulus. .The shrouds can be removed, if necessary, by use of a grapple hook. Small flow holes at the bottom of the- shroud minimize the ef fect of viscous damping .cn1 the scram time.

2.1.6 Regulating' Blade The regulating blade, Figure 9, is a g stainless-steel sheet.>about 11 inches wide and 40 inches long', supported and guided in the same manner as the safety blades described U in Section 2.1.5. It compensates for-small changes of reactivity during normal reactor operation and may be actuated by a servo-control channel.

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i 2.1.7 Transient control Rod The transient control rod is borated graphite contained in a 1-1/4" diameter stainless steel or aluminum tube (Figure 10). The poison section is 15 inches long. This rod is guided laterally by the aluminum guide tube in a special three-element fuel bundle. A hold-down tube extends f rom this guide tobe to the top of the reactor structure and holds the three-element bundle in place during transient rod i movement. ,

2.1.8 Neutron Source  ;

The neutron source is a 100 mg radium-beryllium source irradiated to give an output i

greater than 10 7 neutrons /second. It fits into a source holder which, in turn, fits into a radiation basket occupying one grid module adjacent.to the active core. The source is usually left in for full power operation, and will, with the expected operating cycle, mainrain its output of about 107 neutrons /second.

2.1.9 Core Arrangement The use of the reactor as a training and research tool' requires flexibility of core

. arrangement. These arrangements are subject, however, to the following criteria

a. A mixed core must contain at least four FLIP eley.9nts b.. FLIP fuel must be located in a central contiguous region

c. -The core must be a close packed array except for (1) a maximum of.two, 3-inch square lattice positions which must contain an in-core experiment or experimental facilityt (2) single fuel element positions.

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d. The reactor may not be operated with a core lattice position vacant except for positions on the core periphery.

e. Calculations indicate that operation will be within technical specification limits on power generation por element and fuel temperature.

2.2 DRIVE MECHANISMS 2.2.1 Safety Diade Driva The drive mechanism for the safety blades, Figure 12, .right picture, includes a reversible electric motor with an integral worm-gear assembly to reduce speed and prevent drift of the control element. A mechanical slip clutch on the output shaf t limits the force on the contrcl-blade to approximately 75 pounds. A ball-bearing screw and nut are used to raiso and lower the control element.

Each blade-is coupled to its mechanism through an electromagnet that provides gravity scram when de-energized. In order to minimize friction and possible binding, no more guide bearings than necessary for proper alignment are used, and large clearances are provided on guide bearings.- All lubricants are_ sealed to-prevent leakage into the reactor pool. The working parts are enclosed in a housing, and limit switches'are protected with suitable-covers.

1 2 - 16

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2 - 22

The safety blade drive mechanism operates through a stroke of 16 inches at a normal speed of approximately 14 51nches per minute in either direction. Coasting of the mechanism is limited to less than 0.05 inch. Instruments give a continuous indication of position, accurate to 0.02 inch, and limit switches i at the ends of the stroke shut off the drive I motor and operate indicating lights. In addition, a limit switch within the scram magnet gives remote indication when the magnet engages the controi element.

The control shaft joining each drive with its blade is made up of an armature rod, connecting shaft, piston, and lower tube. The piston and armature rod arc screwed to the top end of the connecting shaft. The bottom end of the shaft is pinned into the lower tube. A polyethylene sleeve-bearing aligns the control shaft and limits radial play to 1/32 inch. The bearing guide tube is attached to the suspeasion frame.

A steel armature disc is welded to the top of the armature rod. During reactor operation, the disc is held by the scram  !

magnet on the drive. mechanism. When the reactor is scrammed, the magnet de-energizes and releases the control shaft and blade.

The release completely separates the blade from the drive mechanism. The element is free to fall within 60 milliseconds of presenec of a scram condition, and it then drops into the core under the force of gravity.- A dashpot above the guide tube bearing receives the piston and decelerates the shaft over the last five inches of its 16 inch travel. In case of power failure, L . scram is automatic.. To recover the control element after scram, the mechanism is run down and the magnet picks up the element.

2 - 19

2.2.2 Regulating Blade Drive The regulating blade servo-controlled drive (Figure 12, lef t picture) is driven by a servo motor through a spur-gear train and components otherwise similar to those in the safety blado drives. Remote position '

indication is again provided. A solid coupling replaces the holding magnett there-fore the regulating blado cannot scram. The maximum speed of travel is 17 + 0 5 inches -

per minute with a total stroke of 16 inches.

A control shaft joins the regulating blade to its drive mechanism. The shaft is aligned by a polyethylene sleeve bearing similar to that used for the safety blade shafts, but without a dashpot, with radial clearance between shaf t and bearings of 1/32 inch. The guide tube is attached to the suspension frame, Figure 1, 2.2 3 Fission Counter Drive

-The fission counter is positioned by a Morse chain-and-sprocket drivo giving a total motion of 60 inches. The drive has adjustable stops and covers the range from source level to full power with its four positions. Position is indicated at the console, and the control element drives cannot be withdrawn when the fission counter is in motion.

2.2.4 Transient-Rod Drive To allow transient operation, use is

• made of a pneumatic-electromechanical drive-system to eject a predetermined amount of-the transient rod from the core. This drive

. system, located on a special mount attached to the locating plate, is shown in Figures 13 and 14. J l

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The pneumatic portion of the pneumatic-electromechanical drive, referred to hercin as the " transient" rod drive, is basically a single-acting pneumatic cylinder. A piston within the cylinder is attached to the transient rod by means of a connecting rod.

The piston rod passes through an air seal at the lower end of the cylinder. Compressed air is admitted at the lower end of the cylinder to drive the piston upward. As the piston rises, the air being compressed above the piston is forced out through vents at the upper end of the cylinder. At the end of its stroke, the piston strikes the anvil of a shock absorber. This piston is thus decelerated at a controlled rate during its final inch of travel. This action minimizes rod vibration when the piston reaches its upper-limit stop.

An accumulator tank mounted beneath the removable floor plate of the bridge (Figure 1) stores the compressed air that operates the pneumatic portion of the transient rod drive. A three-way solenoid valve, located in the piping between the accumulator tank and the cylinder, controls the air supplied to the pneumatic cylinder.

De-energizing the solenoid valve interrupts the air supply and relieves the pressure in the cylirder so that the piston drops to its lower limit by gravity. With this operating feature, the transient rod is inserted in the core except when air is supplied to the cylinder.

The electromechanical portion of tha transient rod drive consists of an electric motor, a ball-nut drive assembly, and the externally threaded air cylinder. During electromechanical operation of the transient rod, the threaded section of the air cylinder acts as a screw in the ball-nut drive assembly.

2 - 23

l These throeds ongnga a action of balls contained in a ball-nut assembly in the drive housing. The ball-nut assembly is in turn connected through a worm-gear drive to'an electric motor. The cylinder may be raised or lowered independently of the piston and control rod by means of the electric drive. Adjustment of the position of the cylinder controls the upper limit of-piston travel, and hence controls the amount of reactivity inserted for a pulse.

A system of limit switches is used to indicate the position of the air cylinder and the transient rod. Two of those switches, the Drive Up and Drive Down switches, are actuated by a small bar attached to the bottom of the air cylinder. A third limit switch, the Air switch, is actuated when the piston is held in the cylinder by air pressure.

23 COOLING AND LIQUID WASTE SYSTEMS 231 Pool The aluminum-lined concrete pool,  ;

Figure 15, is'8 feet wide, 12 feet long, and 27-1/2 feet deep. It is penetrated by experimental ports. A pit, 6 feet by 4 feet by 6 feet. deep, with shielded cover, is provided for storage of fuel elements awaiting shipment to a reprocessing facility and for-storage of fuel in a shielded location, should maintenance of the in-pool portions of the reactor be necessary. Three aluminum fuel storage baskets (with aluminum clad cadmium poison plates)'

fit inside the pit and hold up to 60 fuel bundles for storage. The multiplication constant (K gg for the three storage baskets with full compTeme)nt of fuel elements is less than 0.8. Three additional fuel racks, each holding up to 9 four-element bundles, are attached.to the pool wall at about-the level of the grid box. The multi-plication constant for these racks, when fully loaded, is less than 0.8 2 - 24

The pool water is kept within the following limits:

Temperature (at Core cooling water inlet) . . . . <1300F Resistivity . . . .

. > ? x 105 ohm - em Activity . . . . . . <5 x lo-S uc/mi (see section 31.4 for further information on pool activity.)

The reactor core is cooled by natural convection of pool water through the core.

The 130 F temperature limit is imposed by domineralizer resin tolerance and by humidity control considerations.

The resistivity limit is set to reduce corrosion effects, extending the expected lifetime of the fuel elements and controlling water radioactivity. Routine checks of resistivity are made to determine the necessity of regenerating the demineralizer.

The radioactivity of the pool water is continuously monitored by an area monitor station located near the dominera11zer.

Should the pool water reach the activity limit above,the reading on this area monitor will increase during. periods when the reactor is not operating-(N 1D activity may mask this activity during operation). In addition, water samples are routinely analyzed for activity by other methods which give a more exact identification of quantity and type of activity present.

No problem is anticipated in maintaining pool water radioactivity below the indicated limit.

2 - 25

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The pool water is cooled by the system shown schematically in Figure 16 The secondary system is designed to operate '

year round, and is similar to other instal-lations that must operate at subzero ambient temperatures. The secondary system has a combination of manual (seasonal) and automatic (day to day operation) controls which maintain the temperatures about as indicated for 1.0 MW operation.

The intake and outlet diffusers are constructed to preclude draining more than 1 foot of water even in the case of a pipe ruptu re . This will maintain at least 19 feet of water above the active core.

The secondary side of the heat exchanger is operatec, at a higher pressure than the primary si<1e both when the pumps are running and under static conditions. A leak in the heat exchanger would result in secondary water encering the primary system. Such ,

leakage could be detected in two ways. First, the sucondary water influx will increase conductivity in the pool water. Second, the pool level float switch will be actuated by high water level should as much as 150 ,

gallons of secondary water enter the pool water e- system.

The cooling system will dissipate 1.0 MW year round with primary temperatures approximately as indicated in Figure 16. The system components have enough capacity to enable dissipation of at least 2 MW for wet bulb temperatures less than 750F.

The primary system continuously circulates pool water through the exchanger. Pump Number one' circulates secondary water from the pumps through either the tower or the bypass-line. -

The automatic diverting valve is controlled by sump water temperature and operates so there- is either full- flow through the tower ,

or full flow through the bypass. Pump. Number two circulates water from the sump through the heat exchanger. The automatic diverting valve in this loop is controlled by primary coolant temperature at the heat exchanger exit.

2 - 27

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The valve operates to allow the propcr amount of secondary flow through the heat exchanger to maintain primary coolant temperature at the desired point. Cooling tower f an speeds are manually controlled for seasonal variations.

Nitrogen 16 suppression is accomplished by a jet-type diffuser system. The system pumps about 80 gallons of water por minute from near the pool surf ace through a single nozzle having a 0.75 inch wide, 6.5 inches long opening. The nozzle is located 5.5 feet above and 0.5 feet east of the core, with the diffusing stream directed downward at a 45 degree angle toward the west end of the pool.

The pump for the N. suppression system is located on the outside east face of the reactor's concrete containment structure, about 8 feet below the pool surface. The system is constructed so as to preclude draining more than one foot of water from the pool in the event of a pipo rupture.

2.3.3 Pool Make-up and Clean-up System The pool make-up and clean-up system is shown schematically in Figure 17. Water is circulated from the pool surface, through the pump, through the domineralizer, and then into the pool under the core box and coolant header.

The pump m'aintains about 10 gallons / minute flow through the domineralizer. The domineralizer is a mixed-bed type with provisions for regeneration of resins or discharge of spent resin and loading with new resin. A water softener supplies softened water for regeneration of the domineralizer.

L Normally, make-up water is supplied by the still. The still delivers water to a storage tank from which it is pumped (by the pool recircu-lating pump) ir.to- the pool to maintain pool water level. An alternate and more rapid method allows water to be fed through the domineralizer into the pool . In either case, impurities in make-up water are reduced to less than 1 ppm before going into the pool.

2 - 29

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s Wastos from roganOrcticn of the domineralizer are discussed in the next section.

Flow from the domineralizer to the pool is through valve 11, check valve 709 which pre-vents back flow, and valve 719 into the 8 inch pipe loop and into the bottom of the grid box.

The 8 inch line is equipped with a siphon breaker at the top of the pool so that rupture of the line at the domineralizer outlet or of the 8 inch line outside the shield cannot drain the pool to a level that will uncover the core. A second 8 inch line is flanged off on both ends. The 8 inch lines were originally installed to allow a forced-convection cooling mode, but the lines are used only as indicated above.

A two inch line whose rupture could have caused loss of pool water has been permanently plugged inside the concrete shield and is pre-sently sealed off outside the shield. A pool drain line and valve have been climinated.

Should valve numbers 5 (shown in both figures 17 and 18) be left open upon placing the system in its normal operating condition, as much as 400 gallons of pool water could be pumped to the holdup tank.

16 further loss of water would then occur, since check valve 709 will pre-vent reverse flow from the 8 inch pipe loop to the domineralizer.

All operations involving the make-up and clean-up systems are performed by written checklist-type procedures designed to prevent draining of the pool.

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2.3.4 Waste Disposal System Figure 18 shows schematically the waste disposal system The hold-tank is a 2,000 gallon i stainless steel tank buried beneath the Reactor Laboratory floor. Waste water from regeneration of the demineralizer flows through valve number 5 (normally closed) into the tank during regeneration.

Valve number 713 is normally open, allowing pool overflow or pool curb drains to empty directly into the hold-tank. Valve 712 is normally closed, but it may be opened to allow draining flooded beam-ports. Once waste water is in the tank, it must be pumped out into the sanitary sewer system for ultimate disposal. The contents of the tank are not pumped out until it is determined that the concentrations and total quantities are below the applicable limits in 10 CFR 20. A senior operator's authorization is required before initiation of waste disposal into the sewer system. Since a valve must be opened and a- pump turned on to effect dischargo, accidental discharge is not to be expected.

Tank level is indicated in the demineral-izer area, and an alarm sounds should the tank be filled to its maximum capacity. Provision is made for obtaining samples of the liquid in the tank for analysis.

The tank is vented to the roof 'f the Reactor Laboratory. No air intakes i located nearby, so personnel will not be expt d to any fumes that might be given of f. The R  : tor Laboratory floor drain also feeds inte he hold-tank.

2 - 32

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. 2 - 33

o o 2.4 EXPERIMENTAL FACILITIES Facilities are provided to permit use of radiation from the reactor in experimental work without endangering personnel. Facilities provided with this rtJctor include four beam ports, a thermal column, and pneumatic and hydraulic irradiation systems.

2.4.1 Thermal Column The thermal column, Figure 19, is a graphite-filled, horizontal penetration through the biological shield which provides neutrons in the thermal energy range (about 0.025 ev) for irradiation experiments. The column, which is about 8 feet long, is filled with about 6 feet of graphite. A small experimental air chamber between the face of the graphite and the thermal column door has conduits for service connections (air, water, electricity) to the biological shield face. Detectors for the safety channels are located within the thermal column.

Personnel in the building are protected against gamma radiation from the column by a dense concrete door which closes the column at the biological shield. The door movcs on tracks set into the concrete floor perpendicular to the shield face.

A ventilation system maintains a low pressure within the thermal column so that

.) air flow is into the column when the door is I open. The door is gasketed so that air flow is very small when the door is closed. When the door is opened, however, en air velocity of about 40 feet per minute into the column prevents the A41 activity from diffusing into the Reactor Laboratory. Section 2.4.3 contains further information on the ventilation system for the thermal column and beam ports.

An annunciator is activated whenever the thermal column door is not fully closed.

In addition, an area monitor beside the thermal column door will give an alarm should the reactor be operated at a substantial power with the door open.

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Figure 20 shows the construction cf the beam ports, while Figures 11 and 15 indicate th) positions of the beam ports with respect to the grid box and shield. The ports are air-filled tubes, welded shut at the core ends and provided with water-tight covers on the outer ends. The portions of the ports within the pool are made of aluminum, while the portions within the shield are steel.

A shutter assembly, made of lead encased in aluminum, is opened for irradiations by a lifting device. When closed, the shutter shields against gamna rays from the shut-down core, allowing experiments to be loaded and unloaded without excessive radiation exposure to personnel.

shielding plugs are installed in the outer end of each port. The plugs, made of dense concrete in aluminum casings, have spiral conduits for passage of instrument leads.

I Since extremely high radiation levels could exist should the reactor be operated at substantial power levels with-the shielding plugs removed, a beam port monitoring system ,

is provided. The system consists of detectors mounted on the walls in line.with each beam port and a read-out device at'the console which gives an audible and visual alarm should a preset radiation level be exceeded. The system is set to alarm at a radiation level equivalent to a dose rate of about 60 mrem / hour at the beam port. openings.

2 - 36

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Tha thormni column-becm port ventilction system exhausts the beam ports through the pipes shown in Figure 20. With the beam port open, a linear flow velocity of about 40 feet per minute is maintained into the port opening.

This prevents A 41 activity from diffusing into the Reactor Laboratory. With the beam ports sealed up (the normal situation for most beam port use) very little air is exhausted from the ports, since the beam port drain valve is normally closed. The thin-walled aluminum cans in the in-pool portion of the ports prevents the higher A41 levels in these areas from diffusing throughout the beam ports.

2.4.3 Thermal column and seam Fort ventilation system Figure 21 shows schematically the ventilation system for the thermal column and beam ports. The blower is sized to maintain an air velocity of about 40 feet per minute into all beam ports and the thermal column, should all be opened simultaneously.

The effluent from the system is filtered and discharged into the Reactor Laboratory stack. The filter and blower are at roof level to prevent leakage of water should a l beam port rupture and fill with water.

The system is designed to sweep out the A 41 activity present in an experimental facility when the facility is opened. During ordinary operation the facilities are closed and there is an essentially zero rate'of discharge. When the facility is opened there is a slug of activity discharged.

The average concentration discharged will therefore be extremely low due to dilution ,

by the other blowers and the fact that no 1 activity is discharged most of the time, i Section 6.2 discusses the levels of activity discharged.

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I!IGH I'PESSUHb MLWER TFE;iC/,l,

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" NAP mat VFIGS F1 :11 % 2.1

' PJJ i 'D ' ~ .h , . ; t,Ui N v"'NTII.ATIC'!! :YSTE' 2 - 39

. . _ . _ _ . . . . . . . _ . . _ . __ _ _ _ _.._. _ ._ _ .- -m___.._____.-__...- .._m _ - . _ . .

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CHDCET FUMElOOD EXHAUST ---* - - )WIlOOD r

I.c I ~'l 1 1

, AIR'LINE-2 1/4" STEEL TUBING -+ ~,

ROOM 41 ROOM 43 SOLENOID

+ CABINET' CARRIER LINE -

NE 8.O* l COUNTING-9 5' ELEVATION C1%NGE -

ROCM N

1 0

1 e- CORE ' e .

{@

ELEVATICN CHANGE y -- ~E1.3* 4/

• f t , (2 e' ^

FROM TEE . CORE s -g..

V 21/4"AND "?TUBIDG'/'eo CONCENTRIC' u

y

- r m W w ->

RECEIVE-ONLY '

' ~

- STATION ANDr SWITCHING l E 17.5* DEVICE-

SIP)0N'DREAKERS-SOLEN 0ID VALVES; I

- AND CHECK VALVES

• FROM REACTOR. LAB-FLOOR LEVEL FIGURE 22 1

UWNR PNEUMATIC TUBE SYSTIN 2 - 40 1 I

e o l-2.4.4 Pneumatic Tube A pneumatic tube system conveys samples

, from a basement room to an irradiation position'beside the core (Figure 22). The

" rabbits" used in the system will convey sanples up to 1-1/4 inches diameter and 5-1/2 inches lens. The , system operates as a closed loop with CO 2 cover gas preventing gener-ation of A41 activity.

The controls and send station are

).ocated in Room 41 (see Figure 5, chapter 3).

Indications of " Blower On" and " Rabbit in Reactor" are provided at the reactor console.

An area monitor indicates radiation level and givea a visible and audible alarm should the radiation level exceed a preset level at each station. The preset level is selected ac-cording to the computed activity of the sample being irradiated, but is always less than 1 rem / hour.

The reactivity effect from a sample will be restricted to less than 0.2% <a .

Tests run with water and cadmium samples indicate that sanple reactivity effects will normally be less than 0.01% e . Static reactivity measurements will be run for c samples of fissionable material or particularly strong absorbers such as some of the rare ear th s .

Since the pneumatic tube penetrates the shield below water level, a leak in the tubing could drain the pool. To drain more than 8 feet of water, however, a siphon action would have to be set up. The siphon action is prevented by a solenoid valve controlled siphon breaker at the highest point in the system. The solenoid valves close when the blower motor is energized. When the blower motor is not energized, the solenoid valves open and check valves will then allow air to enter the system if a siphon action starts.

Normally these check valves prevent loss of cover gas from the system.

2 41

o o i The system will be operated using a written check-

-list type procedure to assure that the built-in safe-guards remair. ef fective.

dL"r_eeeive e .only " station-ia-inta-11cd-on-the grounde f 1 eve 1-fl'oor .- Thi a-a ta tionls--also_ installed -itr a-fume hd} p y hood. Both stations are ventilated by 1200 cubic foo t/

pp 3, T i 7 minute fans which draw air into the hood, past the sample gp outlet, and then exhaust it through absolute filters to 0- a stack. Sample activity is limited to a level which, should the sample rupture upon discharge from the system, will result in keeping concentration exhausted below Part 20 limits for unrestricted areas when averaged over a.24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.

2.4.5 Grid Box Irradiation Facilities Most irradiations of more than twenty minutes duration arc performed in irradiation facilities on the core periphery inside the grid box.

Radiation baskets are 3-inch square aluminum contain-ers which fit into the grid plate and may contain one or more samples.

Two types of hydraulic transfer systems are used, however, for most of these irradiations. The smaller hydraulic irradiation device consists of an irradiation terminal which fits into a vacant grid position and a loading device and control valve assembly just beneath the pool surface. This system is powered by bypass flow in the N16 diffuser system.

In the large hydraulic tubes, sample movement is powered by a separate pump located beneath t.he north side of the reactor bridge. The motor for the pump is electrically parallelled to the diffuser pump. Each sample tube extends from about one foot below the pool surface to the grid box, where its bottom end fitting positions it rigidly. A clamping device at the top of the sample tube provides further support and prevents inadvertent movement. The pump takes its suction just below the _ pool surface and directs its flow to a jet pump near the bottom end of the tube, causing sufficient flow '

down the. tube to move samples to the irradiation position and hold them in place. Samples which float return to the top of the tube where they are retained until removal by operating personnel. Non-floating samples can be removed with a 2 - 42

' 4S i

~I Jretriever! tool, or they~may be; installed with a retrieving st' ring attached.. Flow-direction and

" sample in" indicators and controls-are located at the

-pool' top?and control console.

' Rupture of piping connected-to the facilities .

will not result in loss of pool water due to its

' location in 'and immediately1 above the pool. Reactivity Jeffects of samples are much smaller than those associated with installation and removal' of conventional -irradiation baskets. . The remarks regarding_ reactivity _ effects

, ;forEsamples in-the pneumatic tube (Section 2.4.4) apply -

to-these facilities. '

- 2.S

- CONTROL AND ; INSTRUMENTATION .;

g. 7 The reactor will. operate.in three standard modes:

-Mode;.1; Steady . state operation at' power levels up to 1,000;KW.

Mode 2 Square-_ wave operation (reactivity 4 insertions-to' reach a desired steady state 1 power level escentially.instantan- . 1 eously) at power levels between 300 and 1,000 KW -

Mode.3 -' Pulsed operation produced by rapid transient-rod withdrawal that results in a step insertion ,of - reactivity; up _ to 2.1% Zh K/K

(peak' power.of 2,000:MW).

,z' Operation is from a;consolo displayingEall pertinent >

reactor' opera' ting _ conditions. A se'loctor switch is Lprovided 1 for ' steady-state, pulsing, L or square-wave modes ;of ' operation. -

1 1

1 2 - 43

.y

251 steady-state operation For steady-state operation the control blades are slowly withdrawn to obtain the desired power level. At this. level the reactor may continue to be operated manually or it may be switched to automatic control.

The automatic-control channel maintains t power level by servo control of the regulating blade, transient rod, or #2 control blade.

Figure 23 shows a block diagram of the control systen startup Channel As shown in Figure 23, the sensing element for this channel is a fission counter. The counter has a range from 2 nv to 100 nv. Since the counter is movable, its effective range is thus from about 2 microwatts to 2 MW. The pulses from the startup counter are amplified and converted-to a logarithmic count-rate displayed on a meter and recorded. The amplified pulses may also go to a scaler that is used for sub-critical measurements.

Log N - Period Channel This channel monitors the power level of the-reactor over the range from 0.1 watt to-full power. -The Log N - period amplifier detects the signal from a compensated ionization chamber and amplifies the signal to provide a-7-decade-logarithmic display proportional to power level. The amplifier _also extracts period information. The Log N signal is' recorded while the period signal is recorded, displayed on a meter on the console, and fed to the logic element.

2 44 I

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_ L2?dT UP  ; , LOG II  ; q 3AFETI , "tr!L "CL"RCLJREA RADIATICII. _

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AMPLIFIER F

• __ _, __ _ INDICATOR I (~

______.IAI m tl_-____

LINEAR EEAM FORT M)NITORS aueR RECORDEh _ h SQGCRS, LOG N & -- A '

000D FREA?G L. PERIOD PICO- PICO- SERVO AMFLIFIER & FA37 AMPLIFIER '.

_ A!EETER -

M Abf-!ETER AMPT. . INDICATOR RECORDER I l A11an I wo N 7arp __---_______________

REX:0RDER AMPLIFIER OoO AIR MONITOR GASEOUS TDT

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AFILIFIER RECORDEjl #2 SAFEIY TRANSI- '

REC E8

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CORE EAFETY MAGNETS DRIVE DRIVE POS.,1ND. PARTICULATE FUEL  !

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- , POS, ACTIVITY ELEFII.T  !

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, VOLTAGE NLTAGE l3 3 3I l COMROL PCWER MCNITOR TMMUPLE3 2 SUFPLY .' COI. TROL ~ (DE DRIVE 3 FIGURE 23 i ro t NUCLEAR INSTRUMENTATION  ;

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Safety Channels Two safety channels monitor reactor power level f rom about 0.1 watt to full power. The signal from each channel originates in a compensated ionization chamber. The chamber signal is fed into a solid state picoammeter.

The trip output signals from the picoammeters are fed to the logic element where they, along with the period signal from the Log N channel, determine whether power is supplied to the control blade magnets. Should any one scram signal or a combination of scram signals be present, the reactor shuts down. The power level scram trip point is set to 1.25 times the operating level.

Temperature Measurements Fuel element internal temperature is indicated at the console. It causes an alarm and scram at the limiting safety system setting.

The temperature of the bulk pool water is measured at the core inlet by a resistance thermometer. This temperature is indicated on a recorder and causes an alarm and a scram on excessive temperature.

Primary and secondary cooling syster inlet and outlet temperatures, and demineralizer inlet temperature are indicated on the system temper-ature recorder. An alarm on this recorder indicates excessive temperature at any of these points.

2.5.2 Square Wave ooeration This mode is provided for those applications which require that the power level be brought rapidly to some high level, held there for a period of time, and then reduced rapidly producing a square wave of power.

2 - 46

J L

? In_the squaro wave-mode the_ reactor;

-is? brought to a ' level- of 1 to - 1000 watts in the steady-state mode. The mode switch is-  ;

thenichanged to the square wave position.:

A preadjusted step reactivity change is then made to_ bring _the reactor to preset power levels between 300 and 1000101. The reacti -

I-- vity ' step change is made' with the transient '

rod.- Then the automatic-control systemi inserts additional reactivity required to

-maintain the preset-power _ level.as the fuel

' heats up. The; operator must manually augment-  ;

the reactivity; inserted by the servo. In this_ mode the period meter and scram are disconnected and thefsafety channel range switches must be set on their full power

-ranges.-iThe111near; power level _ scram is _

[

maintained.at 1.25 P max. and an interlock-prevents initiation of this mode if 1the range-

. switch'is not on the full power _ range setting.

2.5.3~Pulsinc i operation ,

TheLreactor is brought to a power level of less than-.1000 watts in steady state mode.

The modo switch is then changed to pulsingL ,

mode. When the-switch is in pulsing mode- 3 theonormalineutron' channels are' disconnected 1

--and a high" level; pulsing chamber is connected "

32 to read out the peak power' of the. pulse-on' a4 fast -recorder provided. for_ :that purpose.

~  ; Changing of - the imode switch to pulse. removes _ an >

? ' interlock- thattprevents application-- of f air?

tcr the transient rod unless - the : transient rod 3 l

-is in=the full "in". position. only: thel transient rod is1 automatically' reinserted J' after a_ preset time delay. Fuel temperature a

- 'is recorded during pulsing operation.- The

pulse channels are'also indicated on Figure 23.' i 2.5[4 Blade' Control 4 -

Th'e three safety blades are manually controlled by two switchesr one selects the blade to be moved; the other , a s pistol-grip switch with spring return to "off", ~has 2 - 47 i

.t

-___ .I ._ _ . . . . . . _ _ . . _ _ . . .- _ . _ , , _ , , - . _ _. , . _ , ._

a

positions of " raise",."off" and " lower" and controls #the selected blade.- Only one blade t' Jmay;be raised:at aitime.: A separate switch is-available which will lower all blades at the same time.- The position of each safety-blade is indicated by a-digital road-out,.and-

.the'indicatorLlights on the console show when.

each blade drive is at its "in" or."out" limit

~ ~

u and when the blade magnets are engaged with

-- the - a rmatu re s .

K WV The' safety blades will . scram from any- '

position during withdrawal and run-down.. In the: event of a scram, the manual controls are over-ridden 'and the blade drives run in to

.their "in" limits.. The following conditions s must be met before the -safety . blades can be withdrawn:

1. No scram' conditions ~present and scram relays resett.  ;

2. Count-rate'on'startup chLonel greater _

than 2~ counts per.second; ,

.3.-. Fission Counter not in. motion;

4. Console key switchLset to "on" position-

' i TheTregulating' blade hasgidentical< position j E, ' -indication and "in" and ."out" limitiindication.

It'is manually-controlled.byfa separato-

- pistol-grip--- switch and may; be driven concurrently;

. with ,oneiother controlielement. - The' blade-idrives may beitested-by use-of_a " test"E C

position on'ths key-switch.- Theiscram' relay l must be de-energized .before the ' drives can beLmovad while theckey switchlis' in the test 1

b position.

6 i

o 2 - 48 l-

. u .. -. , . a

O 4 2.5.5 Automatic control System

.. The servo amplifier' controls reactor power icvel in " automatic" and " square-wave" modes. The servo amplifier output drives either the regulating blade, transient rod, or a safety blade as selected by a servo element selector switch on the console. The servo amplifier _ responds tc a power level signal from one of the safety channel pico-ammeters and controls speed and direction of the servo element through a servo motor.

In " automatic" modo, the servo amplifier receives period information from the Log N -

period channel and limits servo element with-drawal to maintain a period longer than a pre-selected level. In " square-wave" mode, the period amplifier is disabled and a " servo error" circuit is employed. This circuit allows servo operation only when the servo error is less than 5%. Servo error is indicated at the console in both " square-wave" and

" automatic" modes.

Additional indicators on the reactor control cubicle are provided to indicate

" automatic on" and-scheduled-power.

2.5.6 Transient Rod Control Movement of the transient rod is controlled by the console-mounted, switch-light, pushbuttons which not only control movement, but also indicate in and out limits.

Position indication is accurate to one per cent.

2.5.7 Scram circuits The scram circuits initiate either relay ( slow) or electronic (fast) scrams.

Relay scram is accomplished by de-energizing the scram relays under one of the following condi tions :

3L - 49

i !,

a- .  % :. ;

i

1. . Fast. period-(adjustable between 5

- and -7 : seconds and set to 5 seconds) ;;

4 2; ' Manual l scram;-

'N

3. High1 voltage failure:inicontro1Hconsoler; 4

.I

~4. Temperature at. core coolant entrance 1 above 130 0F;

5. Power level greater.than-l'.25 P max;.  :;

m

.t'

6. Loss- of control power;

, 7. .

~ Log N - period amplifierLealibrate

-switch not in input position;;

8. . . Timed transient control rodLscram inL

-u pulse mode; 9.- Pool water. level low';-

1 10 . . . Fuel temperature 7above LSSS. j i

- > The control blades: are' free to fall- within -

120 milliseconds of the initiation 1of;a relay u Lscram..

Electronic-scraml-is. accomplished by biasing; transis tors . inithe . magnet- power supply ;into - A

.non-conductance. The eliminationaof frelays El

~

s cresults: in a shorteningf of; thei scram 5 delay-time t

~

11  ; to-:less:- than' 60 milliseconds . : ' For - theClowest -

range settings 1on:the picoammeters,Ethis; delay. I is increased- due to =the time constantcof the. 3 instrument 1at-these' low! settings- Electronic: .

scram;is initiated-by:.

'i

! , 11.  : Power level: exceeding 1 25 P-max; il  ;

-2. Period shorter-thanJ3.5 seconds; .;

1

3. ' Loss ofesignalvfrom picoammeters.

23 1

2 - 50

2.5.8 Alarm and Indicator System Nhen an abnormal condition develops, a horn sounds and a red light comes on. The operator may press the acknowledge button to silence the horn. When the condition is corrected, the green light goes on and the red light 3s extinguished.

The following conditions will actuate the alarm system:

1. Any scram;

2. Neutron flux exceeding 1.15 times the normal value;

3. Reactor period less than 10 seconds;

4. Safety blade disengaged from magnet;

5. Water level in pool two or more inches below normal (also gives an alarm at Protection and Security Headquarters) ;

6. Failure of high voltage power supply;

7. High area radiation level; s 8. Beam port monitor actuated by high radiation level;

9. Air particulate or gaseous activity above normal level;

10. Fuel element tempe rature above LSSS.

11. Count rate on startup counter approaching saturation level;

12. Core inlet temperature above 125 F.;

13. Pneumatic tube blower on:

2 - 51

14. Hold tank fullt

15. Chain switch across stair actuated (See Section 2.6.2);

16. Thermal column door open.

To provide operating information for the reactor operator, the following indicating lights are provided:

1. Scram;

2. Scram reset; 3

3. Safety blade magnet engaged;

4. Power on;

5. Control. elements in (distinct light for each) ;

6. Control elements out (distinct light for each) ;

.7. Regulating blade on automatic control;

8. " Rabbit" in reactor.

2- 52

4 +

- 259: Radiation Monitors; The radiation monitors _are arranged

.-into three_: systems;-the-primary area-monitors,_

beam port monitors,-and air activity monitor.

7 The primary area monitors are located as-- follows::

l.- Domineralizer area;

2. On the' reactor bridge about one foot above the water surface; 3.- Beside the. thermal column door;-

4. At the pneumatic tube send-receive station; ht n # <-

Un %<.w<d ' ES~E'

-Units _1 through 3 have ranges from 0.1-to-

'100 mr/hr, while units 4 and 5 have a range.

-from 1 to-;1000,mr/hr.

Unit l' supplies information on radiation-

_ level.from'the demineralizer. It willibe set at thei beginning: ofia_. reactor run to1 alarm at aLradiation--levelfjustLabove--that reached in-ainormal run. :Unitt2 willibe-set to alarm ~at--

% :a? radiation level just above.that. reached.at i full: power . operation, _

. Unit-- 3 ;is -located -beside - ,

the: thermal column. : It too--is setE to --alarm Jjust abovel normal operating = level.- This unit:

, _willsgiveLanfalarmLif-the-thermal column' door Jis left open when-the reactor is= operated att

-any; substantial _ power.

Unit;( 4 addv are set to alarm at,3 0 mr/hr-

^

-. unless{ the calculate _d- activity of- a sample :

r -results in-a. level above 10 mr/hr. In that-

. case'the~ alarm will;be set =at the calculated radiation le' vel. This unit has. audible and visual signals and radiation level indicators -

at the-pneumatic tube stations as wel-1 as at the console.

2 - 53

)

, _ .  ;- - . . . - . .m._,- ,. __

, 4

.. ==

LUnitsi l1throughi3 are connected to thefReactor! Laboratory evacuation alarm.-

- An alarm--from one offthese-units will-ssound -i the evacuation-alarm if it'is not acknow-ledgedlby the' operator within 30: seconds -!

r-- -

(see section 6.8 ) M4'A /T0 -1

The.--_ beam. port monitor is-an area radiation-monitor installed-to-preclude _the-possibility of unknowingly f generating high radiation -levels 1by : operating the reactor _ at -1 high. power : levels with the beam ports 'open-I The-sensors for this' system;are installed on the! walls of the Reactor Laboratory in direct  ;

line with the-beam ports. The system gives

, - visual and audible alarms'at the console if-

- theLradiation level exceeds a preset value.

- The; monitors are'normally set to alarm at a radiation:-level, equivalent to a dose rate of 50-100 mr/hr at the beam port : flange. -The setting varies froml beam port to beam port

- due-'to different~ distances-between=the-walls

. and:thelbeam_ port openings.

LThe air monitor measures both parti-culate. and gaseous activity cf the:- air die- o charged lfrom the stack. Particulate

activity;is'collecte counted-with a thin end-window-GM

.p on a filter tape and tube-and count-rateLmeter. ; Gaseous _ activity is

~

measured with a large1Kanne ionization ch ambe r ._-- The? system.therefore operates-by

detectingip-activity, fBoth_ particulate-add gaseous ~ activity} levels are; recorded, land, provide'annunciationtshouldipreset-levels becexceeded.

m c-2- 04 1

1

.a. ,c 4:: .

i-

-l The sensitivity 1of the -particulate -l activity monitor allows detection of-s - concentrations of-about.10-10 c/ml of;a~

material-with a- single p1 particle emitted

.per disintegration. The efficiency is higher _ i if more than one_p particle is emitted per disintegration. LThe: sensitivity of-the gaseous activity monitor is su concentration-of;about-1x10ghthata nc/mlfof 1 A41 att the stack discharge can be detected by the instrument.' The efficiency variesJ with the energy of- the p particle. associated i with the-isotope being detected. For B' energies h'igher than - that?of _ A41 the '

efficiency._variesextgemely1 energies 11ower than A 1 the efficiency. slowly;forB varies-more rapidly. The primary activity  ;

,y expected to?be presentzin.the stackLdischarge.

'. ' 'is A41 and?the instrument is' calibrated in .

terms of A41 activity.' .

I 2.6 JSHIELDING:AND~ EXPECTED:-RADIATION LEVELS

,' .2I6.1 Basic ~ Reactor Shield

The reactor is shielded by.-concrete 3 and water.- The coreLis-covered.by720 feet e 'of water? The= shield at core level-consists

_ of about 3; feet'.of : water plus 81 feet-- of

, sg ,

n, -ordinary concrete. Denser concrete:is used

' ?:

. in the1 thermal column 1 door Land beam " port l l plugs. . : Calculations andl measurements. .

0 , -ind ica te radiation:levelsitojbe-expected E

forfl000 KW operationiareE(excepting N

-activityJwhich-is discussed below):

~

SurfaceL of fshield,_ excepting beam port .

x and thermalTeolumn openings - less ~than 1.'5 mrem /hr.

Pool' surface' (leakage radiat' ion UNo N less than.15;. mrem /hr.

" Hot: spots" - measurements-have shown that higher _ radiation levels exist around the beam ports-and. thermal column. Ex trapola tion -

m from these measurements indicates the maximum i radiation levels at these " hot-spots" at 2

2 - 55

p . .

l-L 1000 KW will be about 10 mrem /hr around the beam ports and 40 mrem /hr at the hottest spot around the thermal column door. The dose one foot away from the hot spot will be about 5 mrem /hr.

6 2.6.2 Pool Surface Radiation Levels - N Activity The expected radiation level due to N activity at the pool surface directly above the core when operating at 1000 KW is 120 mrom/hr.

The diffuser jet system will normally be used, and the radiation level would normally be considerably less than the level indicated above. These radiation levels will be low enough that no hazard will exist to personnel outside the Reactor Laboratory or in normally occupied' levels within the Reactor Laboratory.

Radiation levels on the walkway surrounding the pool are expected to be around 20 mrem /hr while the reactor is operating at 1000 KW without the diffuser operating.

' he entire Reactor Laboratory is posted as a radiation area. A chain and switch-arrangement is positioned on the north stairway to the pool surface so that an alarm will be -

sounded should entry to that area be made while the reactor is operating, thus assuring that personnel will'not enter the area without knowledge of the reactor operator.

The south stairway, leading from the console area to the pool surface does not have a chain and switch arrangement,- as does the north stairway. Access to these stairs is gained'only through-the console area and is thus well monitored. No difficulty is expected in maintaining radiation doses to individuals below those doses permitted in 10 CFR 20. i 2 - 56

o .

2.6.3 Demineralize r

'Tha dose. rate in the demineralizer area due to activity removed from the pool water by the demineralizer may be as high as 100 mrem /hr for short periods of time at points close to the demineralizer. The location of the demineralizer is such that shielding can be easily provided'should it prove necessary. Since it-is expected that the radiation level will normally be much lower than the value quoted above, the area around the demineralizer will initially be roped off and shielding will be provided only if it appears that radiation levels will routinely approach the value quoted above.

?

2.6.4 Heating Effects.in-Shield and Thermal Column Heating effects caused by abuorption of gamma radiation and fast neutrons- a.re i within allowable limits. For all calculations, itLwas assumed that the pool water was at ,

1. G the 130oF temperature limit, and the reactor was operated continuously at 1 5 MW.

The_ heating in the concrete shield j is approximately 20% of the maximum suggested by,Rockwell*,

+: e Analysis.of the heating rate in_the lead shield for the thermal column indicates that. the maximum temperature of the lead -

will be less than 217 F. Calculation'of the graphite temperature in the thermal column indicates a maximum of 2440F.

Rockwell, Theodore, Reactor Shielding Design Manual, McGraw Hill, 19ft.

2- 57

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2 . 'T FUEL ELEMENT MEASUREMENTS Required measurements of fuel element bow and elongation are made with the measurement and maintenance tool (shown in Figure 24 with a dummy element installed). This tool is used for dis-assembly and assembly of fuel bundles as well as for fuel measurements. It is used under about 13 feet of water where it hangs from a fuel storage rack in the reactor pool. A fuel-element handling tool also operates the socket and crowsfoot wrenches. An air-operated clamping device reproducibly positions the bottom end box of the fuel bundle.

For disassembly or assembly of fuel bundles, the maintenance and measuring tool holds the bundle, provides a reference plate for the crowsfoot wrench, provides storage space for four individual fuel elements, and restrains the individual elements af ter they have been screwed out of the bottom end box. 1te handling-tool then may be used to remove individual elements and place them in the storage positions.

While it is possible to' disassemble and re-assemble the fuel bundles, it is tedious. The crows-foot wrench must be used for the' initial loosening of each element and for torqueing each element upon reassembly. While _the elements are loose enough to turn freely, a socket wrench can be used on u hexa-gonal portion of the top fitting. Disassembly or assembly of an element takes about 30 minutes.

Because of the excessive time and added handling required to disassemble the bundle and .

measure each element in a separate measuring tool, the tool was designed to make the measurements without disassembly.

Figure 25 shows the three sensors employed (a portion of the housing is removed in this view).

Each uses a differential transformer as transducer to give a remote electrical output proportional to disp 3acement of the sensors.

2 - 59

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The X and Y sensors employ spring-loaded aluminum wheels attached to the differential trans-former cores. When the bundle is lowered into the tool the wheels are forced back and they then ride on the fuel element clad surface. These sensors are adjusted to give a zero signal for a standard fuel element dummy.

The length sensor differential transformer is actuated by one lobe of a cam. A second lobe of Ebis cam is rotated into contact with the top edge of the fuel element cladding by a leaf spring Attached to the operating rod. The cam pivots out into the measuring position only when the operating rod is fully withdrawn. The length sensor is also adjusted to zero output for the standard fuel element dummy.

A readout box located at the pool surface allows.the operator to select X, Y, or length readout.

Differential transformer core position is indicated by a conter-zero meter. A recorder output is provided to the horizontal axis of an X Y recorder. Polarity isi set so an increase in element length or a bow away from the center-line of the bundle gives a positive meter ' indication or recorder readout.

A guide at the pool surface is positioned directly over the-measurement tool and clamped in place. An extension connected to the operating rod

. fits.through a bushing in this guide and engages-a gear-driven vertical position readout device. The vertical position. readout signal is supplied to the vertical axis -of the X-Y recorder.

The dummy fuel bundle has one dummy element exactly 0.100 inches longer than the other three elements. This element also has a section in which the radius has been reduced by 0.060 inch. By using this element the attenuation and zero controls on the readout box may be adjusted to give a calibrated readout of bow and length.

Af ter calibration of. the tool, measurement can be made on standard fuel elements. Figure'26 is a standard data sheet used to record measurements.

The standard setup used gives a 1 cm horizontal displacement for 0.060 inch transverse bend (bow) and a meter reading of , units for a length increase 2 -61

. _- __ . _ _ _ - . . ~ . - _ . - _ . - _ _ _ _ _ . . _ _ . _ _ _ . . __ _.

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of-0. LOO _ inch. A_-trace:is drawnLfor both_the X and 'i Y: sensors._while the: length measurement meter reading is D recorded: in- the . center' section ;of- the form. 'A  :

complete - setL of measurements for _ all four elements -l'

'in a bundle 1can be completed in-'about twenty minutes.

7 since the1X: and- Y readou'ts are 90 apart, the O maximumLpossible bow will be the square root:of the-

- sumiof - the squares of. the bows indicated by direct l measurement. 'As-long-as noither-measured bow exceeds 0.088 inch, no calculations or other measurements are necessary.- If either-bow measurement exceeds 0.088 .;

inch,,then'the; square root _offthe sum.of the squares-lofJth'e measured bows must be calculated to~ determine-I

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whether. or .notothis resultant 'is less than--l/8 inch.

If-the calculated number-is-less Unan 1/8- inch, the:

ielement is within technical specifications. Should

• 4 the calculated.bowLexceed.1/8 ' inch, the crowsfoot wrench may be used;to rotatenthe, element being1

. measured so that7 the> reading of one bow sensor is maximized and the~true bow may-be determined directlys toisee whether it exceeds technical specification-limits.

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2 - 63

Chapter 3 LOCATION AND BUILDING-3.1 LOCATION 3.1.1 Site Description The reactor is _ housed in a building termed the Reactor Laboratory, on the campus of the University of Wisconsin in Madison, Wisconsin. The location of this building is shown in Figures 1, 2, 3 and 4. The building adjacent to and-surrounding the Reactor Laboratory on the east, north and west sides is the Mechanical Engineering Duilding. The Engineering Research Building is adjacent to the Mechanical Engineering Buildi ng on the east. Drawings of the Reactor Laboratory, the Mechanical Engineer-ing Building and the Engineering Research Building are contained in Figures 5-10.

Thn-nearest residence is over 500 feet-from the reactor location, and the population distribution _ surrounding the reactor'is as shown in Figure'll. The nearest classroom is at least 60 feet from the reactor location

h several walls of-various materials separating the two.

Immediately to the south of the Reactor Laboratory is a service road and a parking lot and immediately to the north is a-courtyard. To the west are Nuclear Engineer-ing Laboratories which flank the Reactor Laboratory on both the basement and - first floor levels. - - To the east is a Nuclear Engineering Laboratory and a Heat Power Laboratory flanking the Reactor Laboratory

-on the basement and firct floor levels re spec tively .

The Heat Power Laboratory is a si ngle room, wi th 10 foot partitions separating different experimental areas, and covers the entire first and second floors on the cast 3 -1

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Appendix CALCULATION METHODS FOR ATMOSPHERIC RELEASE OF RADIOACTIVITY

References:

(1) Meteorology and Atomic Energy, U. S. Dept.

of Commerco Weather Bureau, Govt. Printing Offico, Washington, D. C (July 1955).

(2) F. A. Gif ford, Jr. , Atmospheric Dispersion Calculations Using the Generalized Gaussian Plume Model, Nucienr Safety, December 1960.

(3) Calculation of Distance Factors for Power and Test Reactors, (TID-14844), USAEC, Mar. 23, 1962.

A. Models_Uned for calculations For Sutton's diffusion model, the maximum concen-tration (Xmax) at any point downwind is given ass (1) Xmax " on 2

, whero (Ref. 1) 9

• maan wind speed Q = release rate, Ci/ soc h = stack height For the generalized Gaussian Plume Model, the maximum I concentration is given by the same equation (Ref. 2, eq. 8) .

l For calculations in this report, the following values are used:

U = mean wind speed a lowest monthly average

= 3.54 meters /sec.

h = stack height above ground = 17.1 meters and (2) X,, = 2.26 x 10-4 0 Wc/ml.

Reference 2 presents a method applicable to release from buildings with zero stack height to approximate release from buildings with zero stack height to approximate release from leaks in a containment structure. This A- 1

e

• l relation is given by equation 4 of the reference as (3) X= 0 (to o + CA)y where X, 0, and y are defined as above, and C is an empirical constant with a range from 1/2 to 2, and A is the minimum building cross section.

For calculations in this report we will neglect the atmospheric dispersion term in the equation, giving (4) X C

Inserting values for this facility, and using a value of 1 for C,-

~

(1) ( 30 f t . ) (44 fT.~) 9.29 x,10-2m#/ f tz ) [3. 54m/sec . )

~

(6) X = 2.3 x 10 QpCi/ml.

B. h, ample calculations Supporting Section 6.2, The maximum release rate for A 41 activity is 13.3 pC1/sec.,

and the resulting concentration is calculated to be, from equation (2)

~

(7) X =

(13.3 x 10-6) (2.26 x 10 )

~

= 3 . 01 x 10 pCi/m3.

Calculation by equation (6) gives the value

~ ~

(8) X = (13.3 x 10 ) (2. 3 x 10 ) pCi/ml.

~

= 3.06 x 10 PCi/ml.

Although the two methods used to calculate the above values cannot both. be applicable, the actual value will likely be between the two values above. The more conservative value is used in the text.

A-2

r o are derived as shown belows (10) X

1. 03 " (.0083 ci) ( 2. 26 x 10-4) 4238

-10

= 4.43 x 10 uci/ml.

The remaining values are calculated in the same manner.

The activity release was also evaluated through use of equation (6). (This calculation would be applicable particularly to leaks of the activity with the ventilation system not operating.)

83 Again using Br for an example, and assuming the same release time, (11) X Dr = .0083 ci) 2.3 x 10 b uC1/ml.

p3g

~

= 4.50 x 10 WCi/ml.

This value is a factor of 10.17 greater than that evaluated by the Gaussian Plume Model. All similar valuce in Table 1, columns 11 and I may be multiplied by this factor for a more conservative case.

I A- 5 t.

N .

C. Calculations Supporting Section 6.5.4 (a) Whole body exposure The activity concentration of the insoluable volatiles in the reactor room air was determined by dividing released activity by room volume:

^ " ~

=f*xy*0 9 c3 = 2.95 x 10 UCi/cm Since 3.7 x 10 dpm = 1 UCi,-h-= 109 y/sec-cm .

The maximum dose rate is calculated by assuming the room is equivalent to a hemisphere with a radius of 782 cm. In addition, the average gamma energy is 0.7 Mov, the attenuation coefficient for the air is 3.5 x 10-5 cm-1 and the 2flux to Jose rato conversion factor is 4.2 x 104 t/cm see per mr/ min.

Using the relation DR =

Os i- ~Rb g

where DR = dose rate in mr/hr S = volumetric source strength in y/sec em 3 R = outer radius of hemisphere I = attenuation coefficient for air C = flux to dose rate conversion factor yields a dose rate of 60 mrom/ hour.

(b) Dose to the Lungs l The dose to the lungs was calculated by first I

assuming uniform dispersal of the released volatiles-in the laboratory volume, giving a concentration of 6

A S.54 x 10 Ci uCi

~G~ " 2 x 109 cm3

• cm 3 A-3 1

, c ,

Since the "standa d man" breathes 3.25 cubic meters of air per active hour, he would breathe approximately 0.210 m3 in 10 minutes. If this number is increased to 0.300 m3 to allow for excitement, then his lungs would be exposed to an activity of Activity 5 volume 1 e = (0.00275 uci ) (3 x 10 cm )

em

= 825 UCi .

The dose to the lungs is then calculated from the following expression to be 1 rad.

0 -

Dose (rad) = ACR I EiEi(1 Ai -c Mt) i=1

-where A = activity exposure (825 PCi)

C = conversions factor 4 -6 3.7 x 10 S/soe uci) (1.6 x 10 era (100 erg /g-rad) MeV R = lung retention factor (0.125 is customary) m = mass of lungs (1000 grams)

Fi = fraction of total activity Ei = energy of beta for nuclide 1(Mov) 11=radioactivedecayconstant+biolggical release constant (6.7 x 10-8 sec~)

T = time of exposure (assumed infinite)

D. Sample Calculations Supporting Section 6.5.5 Release rate, Q, for an isotope is the total quantity released to air (column E of Table 1, Chapter 6) divided by the assumed release time. The release time used in further calculations is the time required for the exhaust fan to make a complete air change, i.e.

(9) Trelease=2000m/0.472mfsec = 4238 sec.

Using the generalized Gaussian Plume Model (i.e.,

equation 2) , and demonstrating with data for Br83, concentrations released-to unrestricted areas (Table 1)

A- 4

!) ,

e TABLE OF POPUIATION DISTRI!qTION FIGURE 11 Radius of Circio Permanent with Reactor at Center Population **

500 ft. 0 1,000 ft. 686 1,500 ft. 1,687 2,000 ft. 4,497 5,000 ft. 26,259 5 miles (approx.) 105,000 Total-Population, City of Madison 173,258 Total Population of Suburbs 42,082 Total Urban Population 215,340**

Total University Student Population in Madison during academic year 32,806 Maximum number of Spectators contained by Camp Randall Football ,

Stadium 78,280 Maximum nunber of Spectators contained by. Field House 13,140 Estimated maximum number of Students attending classes at any one time in the East Wing of Mechanical Engincoring Building during last year 300 NOTE: Of this 300, 150 were in class-rooms on the third floor of the wing.

Estimated maximum number of Students attending classes or doing research at any one time in the Central Wing of the Mechanical Engineering Building during the last year 50

• Taken from 1970 consus data

• • Permanent population does not include students.

• • • Includes the cities of Madison, Middleton and Monona, the villages of Cottage Grove, McFarland, Maple Bluf f, Shorewood Hills, Verona and Waunakee; the entire towns of Blooming Grove, Cottage Grovo, and Madison; and parts of the towns of Burke, Dunn, Fitchburg, Middleton, Pleasant Springs, Springfield, Verona and Westport.

3 - 12

i o ,  !

wing of the Mechanical Engineering Building except for a few of fices along the north and south walls. The south end of the Laboratory is being used for graduate student research and the north end is being used for undergraduate experiments. Eventually the entire Lab will be used by undergrads. The basement area north of 1 the Nuclear Engineering Laboratory is being used l for storage for the Heat Power Lab and for some I undergraduate experiments.

3.1.2 Topography '

l The campus is set on a narrow neck of land i

between Lakes Mendota and.Monona. Lake Mendota (15 square miles) lies northwest of Lake Monona (5 square miles) and the lakes are only 2/3 of l a mile _ apart at one point. Drainage at Madison  !

is southeast through the Yahara River into the Rock River which flows south into Illinois and then l 1

west to the Mississippi.

Madison is a glaciated area, and the topography

, tends to be irregular. In general, the terrain is hilly although the-hills are not large.

3.1.3 Geoloov and Hydrology In the vicinity of the reactor, a glacial deposit exists which consists mostly of-sands and gravels but may bc quite variable and contain clay and large boulders. Although this deposit may be as much as 100 feet thick, it is probably less than twenty. The reason for the variation in thickness is that the bed-rock sandstone which

. underlies the deposit is very uneven. The bed-rock consists of Cambrian sandstones which are 700 to 800 feet thick and which are permeable-to water. No borings are available at the specific reactor site, but sandstone was found very close to the- surface within about 1500 feet of the site.

Below the sandstone-is impermeabic basement tock.

A chart showing the geology at a well about 2,000 feet from the reactor site is shown in Figure 32.

The Cambrian sandstone layer constitutes a water aquifer, and the ground water flow from the reactor site is generally toward Lake Mendota-3 - 13 h

m .

Yahorn Rivor - Loko Monona system.

Thus, the general flow is toward the east and south. However, as discussed below, Madison obtains its water from wells drilled into the Cambrian sandstone. Consequently, local areas of depression of the water surface caused by pumping of the city wells could cause flow from the reactor site toward the wells.

3 1.4 Water Supply Madison obtains its drinking water supply from several wells drilled into the Cambrian sandstone described above. The location of these wells is shown on Figure 2, and they supply the University as well as the city. All of these wells are cased from ground level into the sandstone so as to keep

.out water from the glacial deposit. The closest well to the reactor site is about 2,000 feet southeast.

An analysis was made of the possibility that loss of water from the reactor or from  !

the hold tanks could affect the city water supply, and a negative result was obtained as indicated below. There are two methods by '

which such water may leave the reactor room: ,

(1) by flow through the floor drains and (2) by loss through the-floor and into the ground.

In so far as.the first method is concerned, it was ascertained that the flow through the floor drains would empty into a sanitary sewer main *.

-From there it would travel through mains via a pumping station to the main sewage plant, located south of and outside the corporate limits of the city. From there, the sewage travels through mains an additional five miles to the south before it empties into an open ditch. On the way, any water from the reactor would become considerably diluted since the minimum flow-rate into the ditch is 7,000 gpm whereas the L *The Reactor Laboratory floor drain empties into L the hold tank. Should the-entire contents of the l pool be let out into the room, however, some water could escape into the sewer system through a drain thimble into which waste water is pumped from the hold tank.

l L

3 - 15

, 9 CITY UNIT WELL NO. 4, MADISON, WIs.

Rondall cnd R3gant street W. H. Cater, Contractor, 1930 Elevation 853.6 D o-5 F5 / <- N Soil, dark grey (fill)

R 5-15 10/ . . ,

\ sand, fine, gray, glacial,dolomitic F 67 15-67 52 sand, medium, gray, glacial,dolomitic 9

M F 67~75 8' x grwel st nes t 1/2 inch

\

75-65 ,loj ,l s ' \ sandstone, medium to fine . lt . crv . de l . glac ,

R , ss . fine to med. , white, part yel-gy,dr1.

A 85 15 30- ~ sandstone, medium to coarse, white N 73 116-120 5/ ,- NxSAtldpigne . coarseJt_Xgl-cv. cart del b ' 120-125 57 N sanastone, fine to med. ,1t. gray, dol . _

R 125-140 15/ < 3 \ sandstone, fine to med, some yel, dol.

! 70 140-210 70 ,

'.\ sandstone, medium, white B s ,

A m i C 210-220 lo , " .

~ sandstone,med, to fine, lt. gray, dol.

H K 0-2.5 D/ ,~ X Dolomite, it. gray, sandy 225-230 5/ .

shale, greenish gray, red, dol.

A 230-335 105 U

i sandstone, fine to medium, it. gray,

,% - dolomitic C S L 'D 335-360 25 r, N

$ sandstone, fine to med.,1t gray, pink R

3 x dolomitic layers 360-375 15 , x sandstone, fine to medium, white 375 420 45 ~'

, sandstone, medium, white, some 20 430 10' --

\ ~ dolomitic laven sandstone, med.tum to fine 240 430-450 20 MgE sandstone, med1, white, pink dol., layers 450 460 10' sandstone, medium, white

'9:

\ '4 7 460-535 55 s sandstone, medium, white, a few dol, s s(, layers s 515-545 30' xe. sandstone, fine to coarse, white .

I 545-565 20' :

~

sandstone, med.to fine, white, some dol.

o 565-:>80 151 , y,\ '

sandstone, medium, white 580-610 30 3 ., 9 N

sandstone, coarse to medium, white 610-650 40' gs sandstone, very coarse to med.,1t. gray 650-690 40\ L ( / sandstone, coarse to fine, light gray 265 690-715 25\$1s / sandstone, very coarse to medium, light p

3s

x. s

'22 715-737 22 vvvv Rhyolite, weathered, red

-C dvvV Formations: Drift; Franconia: Dresbach (Galesville); Eau Claire; Mt. simon; pre-cambrian FIGURE 12 - GEOLOGY 3 - 34

i ,

expected to be low enough that no hazard exists.

3.1.5 g,eismology Reference to Figure 1 of " United States Earth-quakes", U.S. Department of Commerce, Coast and Goedetic Survey--Washington, by R. A. Eppley, shows that there have been no recorded destructive or near destructive earthquakes in the State of Wisconsin through the year 1971. Further, a survey of " Preliminary Epicenters" Reports (daily to weekly publication by the U.S. Dept. of Commerce Environmental Research Laboratories) printed since 1971, indicates that there have been no such earthquakes since 1971. The records do indicate that three such earthquakes have occurred in northern Illinois within 125 miles of Madisont two occurring in 1804 and 1912 had intensities of VII to VIII; a third occurring in 1909 about 75 miles from Madison had an intensity of VIII to IX. The latest tremor felt in Madison occurred Sept. 14, 1972.

Information obtained from Mr. Waverly Person of the Environmental Research Laboratories in Boulder, Colorado indicated the epicenter of the quake was located about five miles south of Davenport, Illinois (about 125 miles from Madison). The quakes magnitude was 4.5 on the Richter Scale; its intensity in Madison was estimated to be no greater than VI-VII on the Modified Mercalli Scale. Even considering the four quakes mentioned above, it is apparent that Wisconsin is not the center of strong earthquake activity.

3.1.6 climatology Madison has the typical continental climate of North America with h large annual temperature range, and with frequent short period temperature changes.

The absolute temperature range is from 107 0F.to -30 0F.

Winter temperatures (December-February) average 20 0-and the summer average (June-August) is 700F. Daily _

'mean temperatures average below 320 for 108 days and above 420 for 210 days of the year.

Madison lies in the path of the frequent cyclones L and anticyclones which move eastward over this area during the fall, winter, and spring. In the summer the cyclones have diminished intensity and tend to pass farther north. The most frequent air masses are of 3 - 17 l

probable maximum rate of entry into the floor drain would not be more than 10 to 100 gpm.

These points, coupled with the fact that stringent administrative precautions will be taken to ensure that water contaminated beyond established tolerance levels is not released to the drain, tend to preclude that the city water supply could be adversely affected by this method.

The possibility that the city water supply could be affected via the second method is also negligible. The base of the reactor is about 8 feet below ground level, and water cannot be dissipated via surface run-off. Since the walls of the building surrounding the reactor are made of concrete up to ground level, significant water loss through the floor could result only if the concrete was breached. In fact, it would appear that the only mechanism by which contam-inated water could enter the soil would be the result of an earthquake sufficiently severe to rupture both the reactor tank and shield, as well as the floor of the building, at a time when the reactor pool water was radioactive beyond tolerance levels. Such a set of coinci-dental occurrences is considered extremely rcmote. Further, even if it did occur, there is no assurance that the water supply would be adversely _ affected. For example, the nearest city well is about 2,000 feet from the reactor site, and it has been estimated

• that water would flow through the sandstone from the reactor to the well at not more than 0.1 foot per day. Thus, as long as 55 years might be required for the reactor water to reach the well.

- Should the hold tanks rupture, a similar l analysis indicates no adverse effect on the well. Furthermore, the quantities of water likely to be lost are small and activities are

• Note: The basic geological and hydrological information in sub-Sections 3.1.2, 3.1 3, and 3 1.4 has been, supplied by Mr. c. Lee noit, Jr.,

l District Geologist, and Mr. Denzel R. Cline, I Geologist, Ground Water Branch, U. S. Geological l Survey.

1 32 16

+ c 317 Meteorology Tables 1, 2, 3, and 4 present data on temperature, precipitation, and snow, as well as on miscellaneous factors. The data for the first three tables were obtained from " Local Climatological Data with Comparative Data, 1955, Madison, wisconsin',

U. s. Department of Commerce, weather Bureau publication. It may be seen that Madison's weather is' reasonably typical of the weather in this region of the country. Figures 13 and 14 present surface and upper air wind rose data.

TABLE 1 AVERAGE TEMPERATURE FOR MADISON

• Average Daily Average Daily Average Maximum Minimum Monthly.

Temperature Temperature Temperature January 26.7 0F, 11.80F. 19 3 F.

February 29 4 14.3 21 9 March 40.0 24.8 32.4 April 54.6 37 1 45 9 May 67 1 48.3 57 7 June 76.4 58 9 67 7 July 82.2 64.0 73 1 August 79 8 62.0 70 9 September 71 3 54.4 62 9 October 60.0 43 4 51 7 November 42 9 29 4 36.2 December 30.0 17.0 23 5 Year 55 0 38.8 46 9

• Normal values based on the period 1921-1950, and are means adjusted to represent observations taken at the present standard location (North Hall) on the campus.

l l 3 - 19 (f

o e polar origin. Occasional outbreaks of arctic air affect this area during the winter months.

Although northward moving tropical air masses  !

contribute considerable cloudiness and precipi-tation, the true Gulf air mass does not reach this at ut in winter and only occasionally at other seasons. Summers are pleasant with only occasional periods of extreme heat or high humidity. Noon relative humidity averages 52 percent in July and 54 percent in August.

The average summer has only eight days with temperatures above 90 0.

There are no dry and wet seasons, but 58 percent of the annual precipitation falls in the five months of May through September.

Cold season precipitation is lighter but lasts longer. Soil moisture is usually adequate in the first part of the growing season. During July, August, and September, the crops are dependent on current rainfall which is mostly from thunderstorms and tends to be erratic and variable. Average occurrence of thunder-storms is on fifteen days during July and August.

• Mr ch and November are the windiest months.

Tornadoes are infrequent. The average occur-rence for Dane County is about one tornado in every three to five years.

The ground is covered with an-inch or more of snow about 60 percent of the time from December 10 to February 25 in an average winter.

The soil is usually frozen from the first of December through most of March, with an average frost penetration of 25 to 30 inches.

The growing season averages 175 days. The most probable period (50% of the years) for the last killing freeze in spring is April 17 to May 2 The first killing freeze in autumn is'most probable from October 6 th through 25th.

The latest recorded killing freeze was on May 25, 1925 and the earliest in fall was on September 16, 1916.

3 - 18

$ e h I

d TABLE 4 MISCELLANEOUS WEATHER DATA FOR MADISON e

Average number of days per year with precipitation of 0.01 inch or more . . 115 Average number of cloudy days per year . . . ~145 Average number of partly cloudy days per year . . . . . . . . . . 120 Average number of clear days per yeer . . . 100 Average daily maximum summer (June-August) temperature (OF) . . . . . . . . 77 5 Average daily minimum winter (December-February) temperature (OF) . . . . 14.4 Record highest temperature, (July, 1936) . . 107 Record lowest temperature, (January,1963). . -30 Maximum monthly precipitation (September, 1915) (inches of water). .

10.69 Minimum monthly precipitation-(october, 1889) Trace Maximum monthly snow, sleet, hail

, (January, 1929) . . . . . . . . 31.8 3 - 21 ,

Y

r 4 ;

I

~

TABLE 2 NORMAL PRECIPITATION FOR MADISON *  !

g mal Total Precipitation January 1.47 inches February 1.27 March 2.03 April 2.49 May 3 21 June 4.02 July 3 40 August 3 07 September -4.11 October 2.00 November 2.20 December 1,44 Annual 30 71 inches

• Normal values based on the period 1921-1950, and are means adjusted to represent observations taken at the present standard location (North Hall) on the campus.

TABLE 3 SNOW, SLEET, AND HAIL FOR MADISON Mean Total Maximum Monthly Year January' 9 6 inches 31.8 inches 1929 February 7.8 21 9 1898 March 79 28.4 1923 April 1.6 13 0 1921 May 0.1 50 1935 June Trace Trace 1954 July 0 0 August Trace Trace 1949 September Trace Trace 1953 October 03 50 1917 November 31 14 9 1940 December 75 23 1 1909 Year 37 9 inches 3 - 20 l

l-

a- . I nu l

1 20 s '10 15% 2 10

)5 . ,

5 15%

Avg. Speed ,

Avg. speed 8.5 mph . 7 9 mph July August  !

l l

l s

5 10 15% 10

, 5 15%

~ >

Avg. Speed Avg.' Speed ,

9 7 mph

• 10.1 mph September- October f

r- 5 10 15% 1 5 10 15%

Avg.-Speed \ Avg. Speed 12.6 mph 11.6 mph November December-FIGURE 33, CONTINUED 3 - 23

o Q f

a 4 IN I) l 1h :5 10 D 15% '

l5 10 igg avg. Speed -

Avg. Speed 11.2 mph 11 9 mph l

' January February

% h 9 =5 '10 15%'

~ -

10 ,5 '

lo 15%

w 1

( i Avg. Speed Avg. Speed <

13 8 mph 13 3 mph March April s

i L

J 5 10 15% 1 5- 10 15%

Avg. Speed , vg. Speed 11.4 mph 10.4 mph May . June

%-Calms . _

h-12 11-11 >Tl 0 C 30 15 0-3 mph Miles per hour Percent Frequency FIGURE 13

- SURFACE WIND ROSES, 1949 - 1954, Madison, Wisconsin 3-22

, m.-- -w-- i.y- - ,.-,,-v--m., ,ey,.r_.,_,--,, v - - , , _m -7 ,,,,--,,,-.y.,m ,- 3. , , - , . , - - , -w-,c- ,y- . - - < -

Au

_ . _ li 5 10 15 20 25% _; a ,5

10 15 20 25s Atiy August i 20 2g 20 25%

September OctobcI

,f '

u.-

.3 8- is lo 15 2o 25% _ _.

- m- 6 :5 lo 15 20 25%

PIcuPI 1 .' 3 - 25

b 11

/ ~

= 10 15 5 20 25% = ~

5 10 15 20 GJ 25%

/ )

January February r

1

/

5 10 15% 9 - 5 10 15%

March April 10 15% 10 15%

May June I 10 31dl > 11 , Calms 0 10 15 20 25 Miles per hour Percent frequency

_ FIGURE 14 UPPER AIR WIND _RO,SES; ELEVATION 4000 FT., JULY 1956 - JUNE 1957 l

MADISON, WISCONSIN 3 - 24

U UF h  ?:ORIH UNEICt.GED '

UNEICAVAED

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mur n TER 10 GE

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AU10M)TIVE LAB 0iE10EI

, FIGUE 15 EA.SEENT FLDOR IEVEL LAYOUT

o e i

3.2 BUILDING

-3.2.1 general Description r

The Reactor Laboratory is shown in Figures  ;

15 through 20. The Laboratory is about 43 by 70 '

feet with a ceiling height of approximately 36 feet in most of the room. The portion'of the ceiling above the consolo area is at a height of 22 feet.

The floor of the room is concrete laid on the ground. The walls are concrete and brick.

The roof is a 1-1/2 inch steel deck with 2 inches of= rigid insulation and a 4-ply, built-up surface.

- The console area is located in the southwest corner'of the Reactor Laboratory. It is separated-Lon - the north and east sides from the laboratory-proper-by.cire reinforced glass. This acts effectively to reduce noise originating from

.i surrounding _ laboratories and equipment. Reactor Laboratory air is circulated through the console area-for ventilation.-

The Reactor Laboratory has L ree windows  ;

which face _the parkingLlot and stadium. 1here are five single doorst.two opening on the west l wall.into the Nuclear Engineering Laboratory at l' ground level,_one opening-on the west wall into' the basementilevel of the Nuclear Engineeringi l Laboratory,1one1 opening into'the parking area, ,

and one~ opening intoLthe Heat Power Laboratory.

One double door opens into;the Nuclear Engi'neering.

x l LLaboratory to-the east of:the; Reactor Laboratory.

~

All doors have glass panes'" 4ch-are covered 1 .with_ expanded, steel. gratin

.The Nuclear Engineering Laboratory to the east of the Reactor Lab. includes Room 5 of:the Mechanical Engineering Building and. extends down

~

"4 the corridor of the Engineering Research' Building

-(connected by double doors) and includesLRooms, q

, B-154, B-157 and B-160 of he Engineering.Research' l Building. Room 511s used primarily forf t'eaching 1

-laboratory courses and contains a suberitical assembly, aireactor simulator, aEhot lab, and' various detection and counting instruments.

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The basement icvel of the Nuclear Engineering Laboratory to the west contains small rooms having concrete walls on the order of 8 inches thick. The small rooms house the pneumatic tube equipment and dispatch station, a radiochemistry laboratory, and both general storage and radioactive storage areas. An instrumentation shop is located on the cast end and a machine shop area for the Reactor Lab is located on the west end.

The ground-floor level of the Nuclear Reactor Laboratory on the west houses graduate student research and office areas and an of fice area for the Reactor Laboratory.

Plans exist for building a room inside the Reactor Lab which will include the upper level along thc north wall. It would be used either for offices for the Reactor Laboratory or for an electronics shop.

The Reactor Laboratory is a restricted area. All doors are kept locked at all times except when authorized personnel are in the room. Keys are issued to a small number of authorized persr.nnel.

3.2.2 Heating and Air conditioning The Reactor Laboratory is heated by the exposed steam pipes running along the east wall of the room and by steam heated convectors located in the laboratory. The convectors circulate the air in the room, and do not cause exchange of air with other areas.

Two air conditioning systems are provided. The first, of about 5-ton

, capacity, is controlled by a thermostat located in the control console and provides cooling air and humidity control for the 3 - 31

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1The-ventilation 1 system is' designed =

to preventL the spread of airborne radioactive j material into occupied areas.- - Accidents 1 4

which might result in dischargesofo radioactive

~

. ;:c material from- the stack' are discussed 'else- a where~in this report', and remarks may be found there indicating tho' concentrations-

- which might be' expected. Justification of a

such deliberate discharge,.especially from -

use of the large fan, is also indicated.

• Figures 19 and! 20 show theslocations' 'k of the system components. Figure-21 isla ischematic diagram cWE this system. -

The-;high-efficiency.. filters useds

~

inLthis-system and in the fumehoods in which: pneumatic tube stations: are : located -

are all ? " absolute" filters manufactured Lby. Cambridge.or; Flanders.- These filters '

provide 97.97% efficiency-for 0.'3fmicron-

' particles. -All but thefemergency exhaust? -

-; fan 1 filters have prefilters..

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- instrumentation. Part of the cooled air:

from!this-unitLis exhausted into the Reactor Labi The system brings in no_ l outside air, nor does it discharge-any-air

= from_the laboratory. The second air conditioning _ unit, of about_20-ton capacity,-

cools and alleviates-humidity problems-in the remainder of the Reactor Laboratory.

E It consists of an air-cooled condenser _ and--

compressor located on - the roof of the building, and evaporator-convectors loceded-in the Reactor Laboratory.- This _ unit caes not: cause air flow into our out lif the -

Reactor _ Laboratory._

" i Condensate, water from both' air

-conditioning systems-goes to the building sewers. l

. Samples may be'obtained for assay.. -l L

iU 3.2.3 Ventilation  !

.T?i u;

' The ventilationisyste* for the. Reactor N1 1

' Laboratory consists of two_ fans with.high-- 1 Lefficiency_ filters,.ductwork, backdraft

dampers, andla. stack dischargingfabove the.

roof of the: east wing of"the Mechanical Engineering Buil' ding.

The ' smaller fan < (the room ' exhaust f an)._

{

has a: free air capacity of:about-1200 cfm.

^

It-is.,normallyf"on" to assureEthat:-any air.:

~

j

flow;isifromjadjacentVareas into;the Reactor- m

• [

-- Labora tory .

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The'largerLfani.(the emergencyLexhaust

_(

fan) is'for use;;in those cases z innwhichLit' "C -

1 isLeonsideredidesirablehto-completelyLchange  ?

j, :theJair . in ;the' Laboratorys to' prevent; spread mi c of" contamination to- adjacentL occupied pareas.

4 W Itchas; a free ' air capacitycof '12;000 cfm.-

+ tThis1 fan may also be1 operated to? force:large

' equantities ofLoutsideLairithrough the. stack-

~

.33

to dilute a~ir exhausted from the stack to 93 permissible. concentration 1offradioactive

materials.

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tchapter 4 1

. PROTOTYPE PERFORMANCE CI'ARACTERISTICS

-AND REACTOR PARAMETERS -

l l

Although the University of Wisconsin Nuclear.

Reactor will differ in-appearance.from the TRIGA Marki III

^

proto type, its; behavior.will be quite similar. Among theDdifferences will be_a; rectangular array of elements e . consisting'.of'about:99 fuel-moderator elements instead: 'i of the 91 ' elements -in - the prototype 1 and .four blade type.

control elements instead of the ; rod type control . elements ' l of the; prototype. .The fuel element diameter'is smaller-in L the UWNR core in'l order .to maintain the - proper metal-

~ ~

-y a to-water ratio in'th'e' core, i

4

, A l- INTRODUCTION,TO PROTOTYPE TESTS '

TheLTRIGA Mark-III prototype,-located j at.' General ~ Atomic!s SaniDiego: facilities,;was 1 first loaded to critical with stainless-steel-clad high-hyd rida fuel-moderator. elements in

. December .,of ~ 1961. .It has been operated since : that -

- timec:at steady-state power levels up-to 1 5'MW  ;

• ' ~

and pulses.have'been; performed:at; peak-powernlevels-3 -

+ ' upftoL6500 MW. The . core- h as accumulated 'some E15001 l

.V-' iMW-hr.3of operation.'a~nd morenthan 5000-transients. - ,;

3 1

- Specification'for.'thes fuel?ini the prototype :are', given- 4 i ik Tablet 1.- The: elements!toibeiused'intthe?UWNR:

i#

w >

facility have a;high'er ~U 352 2 ioading1(8'.5 weight:1per- -

' "l. ,

m.  : cent)~inJ the ZrH,1 as well:as the slightly smaller' diameter citsed iabove.  ;(Seel fueli descriptionE in Chapter?2.)5

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STACK

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CAMBRIDGE TO. LABORATORY-PREFILTER-AND 1A-1000-ABSOLUTE FILTER' ,

TO LABORATORY.

l 9

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~ Figure 21 - VENTILATION SYSTEM 3 - 36 y

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, - -- . . . . .. - . . . . . . . . - ~ . - - - . . .- , -

is. e.

. 4.2=- REACTIVITY CHANGES DUE TO REFLECTOR VARIATIONS 4 Since the University. of Wisconsin uses- '

the existing experimental facilities originally used' for the U-Al flat-plate- type core, the reactivity. cffects ' of reflector variations have been measured. Table 2 lists these reactivity worths.

TABLE 2

}

REACTIVITY CHANGES ASSOCIATED WITH REFLECTOR CHANGES Result, Condition  % Reactivity Flooding >ot all 4-f beam porta + 0.0005 est.

Flooding of: pneumatic tube +.0.0002' t Pneumatic tube samples water + 0.0002 cadmium - 0.0005 Adding.1; graphite 1 reflector

on center of one side.of core + 0.230

? Dropping fuel eleme'nt on top

'of':operatinif core + O.5

~ Adding fuel-element on side-of core -+ 0 77 -

4.3- FUEL ELEMENT WORTHS Because of the different core geometry .

, rectangular parallelopiped rather than cylindrical) and the fact that each ' bundle contains four fuel-

.. moderator -elements like those in the prototype TRIGA Mark III, the = data from the- prototype are not applicable-to the UWNR core. However,-based on.

measured . data from -the UWNR .TRIGA core startup, .

fuel element worths are indicated in Figure _ l.=

4-3

. e: o +

-1 TABLE-1. -

-TRIGA MARK-III PROTOTYPE REACTOR-' FUEL MOD _ERATOR ? S_PECIFICATIONS

~

Overall; length,. inches 28 37 Outside diameter..inchaa 1.47 Fuel outside diameter, inches -1.43 Fuel ~1ength, inches 15 0 Fuel composition -

-U_ZrHy,7 Weight of U235, grams-- 36 5-(avg.)  !

Uranium-235 content, wt% 8

-Uranium-235-enrichment,,$t 20-

- Hydrogen-to- zirconium -- ratio. 1.681(avg.)

Cladding Material = c304-stainless-steel--

cladding' thickness, mils. .

20 .

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0 000 000 000 00 0

0 0' 0 O@ 0 000 h -

0 g Ohh 0 0@0@0 000 00000 Transient rod

( control rod Fuel-moderator element FiR. 2 --- Fuel loading diagram - prototype reactor 4-5

e 7.' . .

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CIC: CIC

1. 1 42 TfGO'AL COLU:G -

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WORTH %4 K/K PEUi2 TIC WBE POSITION H,0g REF. GRAPHITE Rf; A 2.60. h.o B 1.95 3'.h6 C 1. 22 2.18 FIG M b C' l.16 1.15 I.. D 0.77 W3L ML2GNT BU'lDLs UORT!!

E 1.76 2.76 l F I.85 I.55 l F' l.01 1.08 44 H 0.h7 0.63 i

r.

L

4 O The loosened cladding reduces thermal conductivity and results in a higher fuel temperature. The change in reactivity loss at various powers is most rapid during the first few pulses, After approximately 100 pulses, the reactivity reached a value nearly as large as that value quoted for operation after 2400 pulses. Extrapolation of the reactivity loss at 1 MW versus number of pulsea indicates that only after a very large number of pulses (>8,000) will th.e reactivity loss approach 2.8%.

Figure 3 shows the reactivity loss versus reactor power for the prototype Mark III core after 2,400 pulses.

The power coefficient will be slightly smaller in the UWNR due to the larger number of fuel elements.

4.4.3 Fuel Temperatures Measurements of fuel temperature were made in various positions in the prototype core. The highest temperatures were measured in the "B ring" -- the central part of the Core.

Figure 4 shows the "B ring" temperatures for power levels up to 1.5 MW. The maximum temperature expected at 1 5 MW is about 410 c (7380 F) above ambient pool temperature. With-the 130 F limit. imposed on core inlet cooling water in UWNR,- the maximum temperature of the fuel would be -- 870 0F. Since the UWNR core will be larger and will have essentially the same' hot-spot factor, the temperature in the central element will probably be about 700F lower at 1 5 MW. However, both these tenperatures are well below the temperature limits for the fuel materials.

4-7

4.4 STEADY-STATE PARAMETERS Measurements have been made of various core parameters in the 91 element prototype core shown in Figure 2. Important differences expected between the prototype core and UWNR core will be pointed out at the conclusion of each section.

4.4.1 Critical Loading The critical loading for the Mark III prototype was 79 fuel moderator elements.

This loading was one incorporating a poisoned follower section on the transient rod. The critical loading contained 2.86 Kg of U-235 The UWNR core will require a larger loading for criticality due to the reactivity loss caused by.the control blade shrouds.

It is expected that, for a number of fuel elements greater than 60, the UWNR core will have about l_per cent less reactivity even with the slightly higher _ loading of Uranium 235 in each element. It is expected that the critical core will contain about 3 1 Kg of U 235, 4.4.2 Power Coefficient Measurements of the " power coefficient" (loss of reactivity at various power levels) have been made at several intervals during the pulse history of the core (2,500 pulses to date). Some change in values occurs with pulsing. The loss in reactivity at 1 MW was measured to be + 1.96% 6K/K for theLnew core with less than 10 pulses; 'this value increased somewhat to a value of -v 2.27% at 1 MW af ter 2,400 pulses of 1800 onVamplitude. The changes observed are due to a slight loosening of the clad due to thermal expansion of the fuel meats in going from 200C pre-pulse tem-perature to -w 400 C after-pulse temperature.

4-6 l

].I L I

~l Figure 4 -- Measured Fuel Temperature .l above Ambient vs. Reactor Power - 4 1

1 Prototype Reactor  ;

UWNR Data Indicated j 1

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e 6 4.4.4 Isothermal Temperature Coefficient (Bath Coef ficient)-

The coolant water temperature in the

. prototype was varied over wide ranges to measure the resulting reactivity change.

Figure 5 is a plot of the change in available excess reactivity (relative to excess at 20 C) for bath temperatures from 20 to 60 0C. The measurements were made at power levels of less than 10 watts. The coefficient is slightly positive with a not gain in available reactivity of 0.077%.

The average coefficient,0.0019%/0C, is small enough that it is essentially negligible for normal operating conditions.

The effect of the water gap left in the shrouds when the control blades are withdrawn is expected to increase the temperature. coefficient by about 20% in the UWNR, giving a temperature coefficient estimated ato.0024%/0C. This value is small enough to be considered negligible for normal operating conditions.

45 PULSE PARAMETERS Measurements were made of the various parameters relating to pulsing operation of the prototype. The most important of these are given below for- step insertions of reactivity _ up to 2.1% AK/K.

4.5 1 Period During pulsing operation the reactor is placed in a super-prompt-critical condition. The asymptotic period is inversely related to the prompt reactivity insertion.

Figure 6 shows the results of plotting the reciprocal of the measured period versus the

-prompt reactivity insertion. Since the 4 - 10

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. graphic. recording of'the reactor power versus time : at ? a portion of the pulse- before fuel 4 temperature. limiting effects have begun,

~ the: accuracy of the measurements is not so

< .-good as for other parameters. 'The-scatter i m of points about a straight line in Figure 6

- 1 . is due entirely to this difficulty. As can- -

'35 be seen,_the minimum period obtained for reactivity insertions of 2.1% 6 K/K is msec. ,

A 452 Pulse Width The - _ width of_- the- power pulsef is most ,

' conveniently described as the time. interval i between half-power points.- Also_shown in Figure 6 is a_-plot. of the reciprocal of the. a

, measured width _versus prompt reactivity <

S insertion, and Figure--7 shows the-linear i relationship between peak power andI(1/ width)2

- 4.5 3. Peak Power __,

Figures 8, 9,cand-lo show the. inter-relationship between maximum transient power,

m. pulse widths,_;and period.. When considered togetherj these; plots serve to describe 1the general features of ~ theE Mark III core per-formance in pulsing / modes. ~For'aigiven core ~

configurationk the: peak poweri: integral power 4

.in the prompt burst,pand.widthLof the= pulse g areidetermined byathe' reactivity: insertion' I

'made. "It;canibeiseenlfromrthe: plots.thati 'd

- the Jpeak power?-ieL controllableL over ,a rather a wide range.since this parameterJis:very nearly-

~ proportionalito (AK/K - O."/%)2 Pulse width 1 and-cintegral powers,-!on the-other4 hand,Lare H f .approxiniatelyJ11near functions :of-_ reactivity '

l iinsertionsJabove prompt critical- so that

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. Reactivity-Insertion - Protype Reactor UWNR Data Indicated ,

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