ML20204F531
| ML20204F531 | |
| Person / Time | |
|---|---|
| Site: | 05000113 |
| Issue date: | 10/19/1988 |
| From: | ARIZONA, UNIV. OF, TUCSON, AZ |
| To: | |
| Shared Package | |
| ML20204F348 | List: |
| References | |
| NUDOCS 8810210595 | |
| Download: ML20204F531 (89) | |
Text
{{#Wiki_filter:.. _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ t . SAFETY ANALYSIS REPORT UNIVERSITY OF ARIZONA TRIGA MARK I RESEARCH REACTOR OCTOBER 19, 1984 O hDR810210595 estog7 p ADOCK 05000gi3 PDC
m L T' SAFETY ANALYSIS 1EPORT Table of Conunts a 4 Page LIST OF FIG URES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , 111 LIST OF TA B L ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv
- 1. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 II. SITE CHA RACTERISTICS . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 A. Geogrrphy and Demography ..................... .2 B. Nearby Industrial. Transportation, and Military Facilities ..... 3 C. Me te rolog y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 D. Geology and H ydrology . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 E. Seismo log y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 11 1. REACTOR DESIGN ................................9 4
A. Reactor Description .............................9 B. Core and Reactor Physics ........................11 i C. Auxiliary Jystems .............................22 1 D. Nuclear Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 E. Thermal Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 IV. FACILITY CONSTRUCTION .........................30 i A. Building Constructl.sn . . . . . . . . . . . . . . . . . . . . . . . . . . . 30 B. HVAC ....................................32 C. U tili ties . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2 D. Fire Protection ...............................33 V. FACILITIES AND EXPERIMENTS . . . . . . . . . . . . . . . . . . . . . . 33 A. Pneuma'ic Transfer System ......................33 B. Rotary Specimen Rack ..........................33 C. Fast Irradiation Facility, Neutron Radiography Tube, and Graphite Thermaliser Block ...................,..33 D. Central Thimble .............................36 E. Demountable Fuel Element .......................36 l F. Fuel Storage Racks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37 VI. CONTROL AND INSTRUMENTATION . . . . . . . . . . . . . . . . . . 37 A, Control System Summary . . . . . . . . . . . . . . . . . . . . . . . . . 37 l B. Protection System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , 44 , C. Reactor Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . 46
- D. Other Control Room Equipment ....................48 4
1 I L
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, i 9
SAFETY ANALYh!S REPORT Table of Contents (continued) VIL SAFETY EVALU ATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 49 A. Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 49 B. Basis for Safety Analysis Calculations . . . . . . . . . . . . . . . . . 49 C. Fuel Temperature Rise from Loss of Ccolant . . . . . . . . . . . . 53 D. Radiation due to Low of Reactor Pool Water . . . . . . . . . . . . 54 E. Release of Fission Products from a Fuel Elen ent .........57 P. Inadvertent Transient ...........................60 O. Production o( Argon-41 la the Irradiation Facilities and Pool Water . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 62 H. Production of Nitrosen-16 . . . . . . . . . . . . . . . . . . . . . . . . 68
!. Other Radioisotopes in Reactor Pool Water .............68 VIII. CONDUCT OF OPERATIONS . . . . . . . . . . . . . . . . . . . . . . . . . 70 A. Organization and Responsibility . . . . . . . . . . . . . . . . . . . . . 70 t
I i t 1 i i ii
. I LIST OF FIGURES Figure Page Figure 2.1 Map of the State of Arizona . . . . . . . . . . . . . . . . . . . ....4 t Figure 2.2 Map of Eastern Pima County, St te of Arizona . . . . . . . . . . . . 5 Figure 2.3 Map of University of Arizona, Tucson, Arirona . . . . . . . . . . . 6 Figure 2.4 Geological Map and Cross-Sections of Arizona . . . . . . . . . . . . 8 Figure 3.1 Elevation of Reactor and Pit ......................10 Figure 3.2 Reactor Core and Reflector Assembly . . . . . . . . . . . . . . . . . 12 Figure 3.3a Standard TRIG A Fuel Element . . . . . . . . . . . . . . . . . . . . . . 13 Figure 3.3b Instrumented TRIGA Fuel Element ..................13 Figure 3.4 UARR Core Position Diagram ..... ..............15 Figure .1.5 Rack and Pinion Control Rod Drive Assembly . . . . . . . . . . . 17 Figure 3.6 Adjustable Transient Rod Drive Assembly . . . . . . . . . . . . . . 18 Figure 3.7 Schematic Diagram of Adjustable Transient Rod Drive Assembly . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 Figure 3.8 UARR Vapor Compression Refrigeration Unit . . . . . . . . . . . 23 Figure 3.9 UARR Water Purificatic> System ...................25 Figure 3.10 UARR Standard Core S.21 ....................... 27 Figure 4.1 Section of North Wing of the Engineering Building Showing the Reactor Laboratory ..................31 -
Figure 3.1 UAR". Pneumatic Transfer Sptem ..................34 Figure 6.1 UARR Operating Console . . . . . . . . . . . . . . . , . . . . . . . . 38 Figure 6.2 Simplified Rea: tor Control System Diagram . . . . . . . . . . . . . 39 Figure 6.3 UARR Rod Drive. Motor Control, and Indicator Lamp Circuit ......................... 43 Figure 8.1 UARR Organizational Chart . . . . . . . . . . . . . . . . . . . . . . . . 71 111
1 LIST OF TABLES Table Page Table 3.1 Operating Characteritties of UARR Standard Core S-21 .... ...........................28 Table 3.2 Pulse Mode Characteris'ics of UARR Standard Core S-21 .................................28 Table 7.1 Calculated Gamma-Ray De Rates with no Water Shielding the Reactor at One ;Sur After Operat i on . . . . . . . 57 Teble 7.2 Dose Rates in Reactor Room in mrem /hr due to Fission Product Concentrations after B ring Fuel Element Failure . . . 50 Table 7.3 Dose Rates in mrem /hr due to Fission Product Concentrations in Unrestricted Areas with Exhaust Fan Rate of 1250 cfm . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 60 Table 7.4 Isotopes Measured in Reactor Pool Water at 100 Kilowatts Power ..........................70 iv
. T 1
I. INTRODUCTION The University of Arizona Research Reactor (UARR)is a TRIGA swimming pool type reactor designed and constructed by General Atomic Division of General Dynamics Corporation (now GA Technologies of San Diego, California). The reactor was constructed at the University of Arizona in 1958 and went into operation in December of that yetr. The licensed power wts 10 kW thermal with operatic,n at 30 kW possible for short times. The original core loading consisted of 61 aluminum-clad fuel elements. Subsequently, two more cluminum-cbd fuel elements were obtained to allow operation et higher power and the licensed power was increased to 100 kw. In No' ember,1962, a stainless-steel-clad demountable fuel element was obtained and in October,1968, a license amendment allowed the use of a fast neutron irradiation facility in the reactor core. In Novemter,19fc, the expiration date of the facility license was extended to November 20, 1988. In May,1972, a new TRIGA control cottsole, control rod drises, and bridge were inst:.!!ed. In June,1972, a license amendment, based on a revised Safety Analysis allowed the receipt and possession ol' additi ons! fuel for a complete changeover from aluminum-cbd to stainless-steel-cbd fuel. In December, 1972, 87 partially used stainless sel-clad TRIGA fuel elements were obtained through an AEC grsnt. These stainlesuteel-clad fuel elements permit operation in the pulsed mode. The original Arizona TRIGA control console, bridge, control rods and control rod drives, and all the Auminum cbd fuel elements were transferred tc the University of Utah after renovation of the conso!c, this equipment is now a part of the TRIGA Rea: tor Facility in operation at that institution.
1 t t In February,1973, initial criticality with stainless steel-clad fue! was - obtair.2d with 71 fuel elements containing approximately 2.4 kilograms of U-235 and l ~, grsphite reflector elements in non fuel pocitions in the F-ring. j T in June,1973, a neutron radiography tube was installed in the reactos pool. j In December,1975, a motor driven reactivity oscillator was first p' aced in the . i i reactor core. In January,1976, a fuel element instrumented with thermocoupten was
- purchased and installed in the reactor core. In August,1973, an aluralnum-clad ,
] graphite thermaltzer block was installed in the reactor pool. ( t in October,1973, a license amendment increased the maximum reactivity : I l insertion la pulse mode operation fsom $2.10 to $2.50. In January,1981, a new top ' l l' grid plate ws: Installed to allew vertical Oux mapping and void coefficient ! i measurements in the reae'or core. The origlaal University of Arizona TRIGA reactor was the first TRIGA to 1 i be sold by General Atomic (it was the secord to be built; the first being at Generst J i l Atomic's labora'ory in San Diego). Since the updating in 1972, however, the i University of A?izona TRIGA has many of the featurve of more recent TRIGA [ reactors. Only the renector, the reactor pool itself, and the refrigeration and neer j pur.fication systems are original equipment. [
- 11. SITE CilARACTERISTICS ,
l A. Geogrsphy and Demography ( l i lhe University of Arizona Research Reactor is located on the campus of the i I j Unhersity of Arizona. The 325 acre esmpus is located in the city of Tucsen, Pima ( i e County, Arizona, in an area roned resider.tial and sms11 business. r
- The city of Tucson is in southeast Arizona approximately 65 miles from the i
I hirxlean border. It is 2,410 feet above sea level, and situated in a high desert valley ( surrounded by the Santa Catalina hfountains to the north, the Rincon hfo'ntains to I i ; I the east, the Santa Rio hfountaim M the south, and the Tucson hfountains to tne } j i j i I i . _____.b
. t 3
west. The valley is drained by three rivers or washes: the Santa Cruz Riser Rillito River and Pantano Wa',h. The University student enrollment in the fa'l of 1987 was 32,505, and employed 12,405 faculty and f*1ff. De Tucson metropoli an t area's 1988 copulation is 648,492 and is piojected to arcw to 943,268 by the year 2000. The location of the University Reactor is !!!ustreted in Figures 2.1 through 2.3. B. Nearby Industrial, Transportation, and Military Facilities Agriculture and copper mining have lont; been the base of southern Arizona's economy. Davis Monthan Air Force Bue and the Unisersity of Arizona are also major employers in the area. Private companies with large facilities in Tucson include International Business Machines, Hughes Aircraft Company, and Garrett AIResearch. Tucson International Airport, located 6.5 miles south of the university, is used by 13 major airlines. Interstatt 10, a major highway from southern California through Florida, borders the west side of Tucson's downtown district. Rdtroad lines . serving Tucson are Southern Pacific and Amtrak. Davis Monthan Air Force Base is the only major milits.ry facility in the vicinity of Tucson. Located 5 miles southeast of the university, and employias 5.725 military and 1,499 civilian personnel, the Air Force base provides tactical air control for the western United States, personnel training, and obsolete aircraft maintenance and storage. C. Meteorology he climate of Tucson is classified as a wist coast desert climate. Tucson's coldest month is January when the average high is 65.5'F and the awrnge low is 36.4* F. The hottest month is July when temperatures reach an average high of 99.l*F. Tucson is approximately 325 miles from the Pacific coast and therefore hurricane and tropical depression energy directed towards the city is
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Figure 2.3 Map of l'niversity of Arizona, Tucson, Arizona
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i depleted by the time the disturbance reaches the city. Typical wind direction is
, from the southeast in the morning shifting to a serieral west-north-west direction in the afternoon. Wind speed is usually between 5 to 10 mph. The average annual i rainfall in Tvesen b 11.2 inches.
Tornadoes in Tucson are rare. According to the Institute of Atmospheric l 3 Physics at the University, on the average a tornado sighting in the Tucson vicinity occurs once in two years and one may tonch down once every ten years. D. Geology and Hydrology . ] The geology of the Tucson area is typical of the southwestern desert region, i.e., alternating broad v;t!!eys and mountains. Figure (2.4) is a U.S. Geological Survey map of the area. Tucson is located near the center of a valley almost .; completely surrounded by mountains. The valley gradually rises towards its center, and the talversity is located near this high point. The surrounding area is drained by two rivers and one major wash, the P.iilito River to the north, Pantano Wash to the east, and t' e Santa Cruz River to the west and south. All three are normally dry and flow intermittently during spring meltit., of snow on the higher mountains and i during occasional rainstorms. Pantano Wash flows into the Rillito Rivtr, the Rillito 1, River flows into the Santa Cruz River and the Santa Cruz River joins with the Gila j River to flow into the Colorado River. The valley is filled with cenozoic deposits composed of alternating sands, j silts, clays, and gravels, The surface, or most recent, deposits are gravels similar to the rock composition of the nearby meantains. The University of Arizona Research . Reactor is located on the Pima soit series, which is characterizeo by resistant caliche i deposits near the surface. The Tucson ester supply is derised wholly from wells drilled into the cenuzoic fill deposits. By 1991 the Central Arizona Project is I expected to augment Tucson's water supply with water from the Colorado River. ] The University of Arizona has thirteen wells located on campus supplying its own
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e I water system. The water table at the university is approximately 200 feet below the surface. E. Seismology Tucson is located in an area considered to be a low earthquake activity area. No major earthquakes have been recorded in the Tucson area. Tremors which have been measured have been in no greater than the feeble shock (intens!iy IV) range. III. REACTOR DESIGN A. Reactor Description The reactor is located near the bottom of a circular pit as showti in Figure 3.1. The pit contains a steel tank resting on a 1-ft-thick concrete slab. The insHe diameter of the steel tank is 7 ft, and the wall thickness is 1/4 in. Approximately 8 in. of poured concrete surrounds the outside of the tan except for a window 4 ft wide by I ft 10 in, high which was left in the concrete to allow for the insertion of a thermal column at a later date, and a 3 inch diameter circular opening which wu intended to accept a van de Graaff generator beam tube. The steel tank served as the inner form for rauring the concrete and the outer form was a corrugated steel cylinder, which was left in place after pouring. A depression I ft by I ft by 2 in, deep has been provided in the concrets at the bottom of the pit to facilitate complete draining of the tank by a portable pump and hose should this become necessary. The inside of the steel tank is covered on the sides by a layer of Gunite approximately 2 in. thick and on the bottom by a layer approxh..ately 4 in. thick. The entire inner surface of the Gunite is coated with Arr . :n ep?xy-base paint). The pool is 20 feet deep and contains approximately 5000 gallons of purified water. Shielding above the top of the reactor core is provided by a minimum of 14 feet of water. The reactor pit has been designed to ensure containment of the witer.
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The Gunite and its coating protect the steel tank against corrosion by water, and if a ' small defect in the coating should occur, the steel tank will provide a secondary containment. Shielding between the reactor core and the van de Graaff vault in the basement consists of approximately I ft of graphite,1-1/2 ft of water, 2 ft of concrete, and 12-1/2 ft of soil. D. Core and Reactor Physics The core forms a right circular cylinder snd consists of a lattice of I cylindrical fuel elements in water. Figure 3.2 shows the reactor core and reflector j assembly. The fissile volume of each TRIGA fuel element (see Figure 3.3a) is 1.435 in, in diameter by 15 in. long and is a solid homogeneous mixture of hydrided uranium-zirconium alloy containing 8.5 wt-% of uranium, enriched to 19.99% in 888U. The hydrogen-to-zirconium atomic ratio is approximately 1.68. There is a 1/4 inch diameter axial hole in the center of the fuel alloy, which is filled with a zirconium rod. A thin aluminum wafer at each end of the active ft.el cor,ttains samarium oxide, a burnable poison. Each fuel element is clad in 0.000-in. thick 304 stainless-steel tubing. Sections of graphite with a ler.gth of 3.4 in, are located above and below the fuel to serve as top and bottom reflectors for the core. Stainless-steel end-fixtures are provided on both ends of the fuel element. The overall fuel-element length is 28.37 in, r The elements are spaced by top and bottom grid plates so that about 33% of i the core volume is occupied by water, as calculations indicate that, for the I aluminum-clad fuel elements in the original loading, the critical mass of the core would be a minimum with this fuel-to-water rado, In December, 1958, 59 aluminum-ctsd fuel elements were required to reach initial criticality. Seventy-one I fuel elements plus two fuel-follower control rods were needed to achiese criticality l after the core was replaced with the partially used stainless steel-clad elements in
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February,1973. Ninety-one positions are available in the top grid plate. The center i (A-1) position cannot contain a fuel element, being designed to awept a central irradiation facility. The other 90 positions are occupied by fuel elements, control rods, irradiation facilities, experiments are filled with water. The instrumented fuel element shown in Figure 3.3b, may be used to measure fuel temperature in any free position. A core position disgram is shown in Figure 3.4. De fuel elements are supported and spaced by means of top and bottom grid plates of 6061 aluminum. The bottom grid plate is 3/4 in, thick, with holes to receive the lower end-fixtures of the fuel elements and holes to allow convective flow of coolant upward through the reactor core. A lower end-fixture is a 1/4-in.-dismeter cylindrical projection on the bottom of the fuel element. A 5/8-in. shoulder is provided on the end fixture, and the hole in the bottom grid plate is countersunk by a corre3ponding amount. The weight of the fuel element rests on this shoulder and not on the bottom of the end fixture, which is used only to position the fuel element as it is being put into place. The pneumatic transfer system terminus, irradiation tubes, and the F!r may be located in any of the fuel . element positions. The top grid plate is also 3/4 in, thick, with I-1/2-in.ID holes for the fuel i elements and the control rods. The top grid plate does not support the weight of the fut' elements. The holes serve only to determine the lateral position of the fuel elements and to permit withdraw 21 of the fuel elements from the core. Space for the passage of cooling water through the top grid plate is provided in part by three spsees machined in the top end-fixture of each fuel element. Cooling water may I also flow through holes in the top grid plate between fuel elements, and in a narrow gap between the top grid plate and the reflector. The three control rods are loested near the center of the core and include a I transient rod in position C-10, with travel limited by a mechanical stop to provide a l 3
PNEUMATIC TRA!iSTER SYSTEM
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reactivity addition on pulsing less than $2.50; a shim rod in position D-10, with re ctivity worth of $3.10; and a regulating rod in position C-4, with worth of $3.94. The holes in the bottom grid pbte at these locations are 1.505 in. in diameter to accommodate fuel follower-type control rods. Thimbles attached under the grid pbte at positions C-4 and D-10 provide space for the fuel follower when the rod is inserted. The bottom grid plate hole at core position D-1 is also 1.505 in. In diameter to allow the installation of an additional control rod if needed. An aluminum plug is currently installed in the bottom grid pbte at this position to allow the installation of a fuel element. An aluminum plug is also installed in the lower grid pbte h position C-10 below the transient rod which does not have a fuel follower. The aluminum guide tube for the transient control rod is held in place by the top and bottom grid pbtes. The active neutron ahorber in the control rods is of sintered boron carbide (D.C). Fuel-fo!! owed rods are cbd in 304 stainless-steel tubing and the transient rod j is clad in aluminum. The active neutron absorber cannot pass completely through l a the reactor core since it can only drop approximately I in. before coming to rest on the bottom of a thimble or a plug in the bottom grid plate. The drive asremblies for the control rods are fastened to mounting plates
'l 1 j located on the center channel. The shim and regubting control rods have electrically driven rack-and-pinion drives (Figure 3.5) with withdrawal speeds of 19 and 24 in./ min, respectively.
The transient control rod is actuated by an electro-pneumatic system, which ; is controlled from the reactor console, ne drive system permits the transient control rod to be used in the steady state mode as well as in the pulse mode of ( operation. In the pulse mode, the drive sptem is adjustable so that any size pulse may be selected up to the maximum reactivity determined by the limit switches and mechanical block. The transient rod drive system (Figures 3.6 and 3.7) consists of a
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OMDON TO f 80TTOM LIMIT - = CONTROL ROD PlSTON R00 L l i l .j j Figure 3,7 Schematic Diagram of Adjustable Transient Rod } .l Drive Assembly I ; i , i ,
single-acting pneumatic cylinder with associated electrical and mechanics.: components, an air accumulator, and a three-way solerold-operated air valve. ne reactor core is located inside a cylindrical graphite reflector i ft thick and 22 in, deep, as shown in Figure 3.2. The graphite reDector is completely clad in aluminum. A well to accommodate the rotary specimen rack is provided in the reflector such that the rotary specimen rack and the reflector are each individual water-tight assemblies. The reflector material above and below the core is compued of the 3.4 in. long pieces of graphite in the top and bottom ends of the fuel elements. The reactor core is approximately 33% water, by volume, ne physics of the reactor has been studied in considerable detail at General Atomic and elsewhere. The information in this section is derived from experiments at various TRIGA facilities. The General Atomic prototype reactor originally attained criticality with 54 aluminum-clad fuel elements, or about 1.9 kg of 888U. The critical mass of the University of Arizona reactor with stainless steel-clad elemena is apprcximately 2.4 kg of 888U. Experiments performed by General Atomic personnel at the Brookhaven National Laboratory have shown that zirconium hydride has very special moderating properties for slow neutrons ' These e. perimental results can be explained by assuming that the hydrogen stom lattice vibrations can t'e described by an Einstein model with a characteristic energy hv = 0.i3 ev. nis description is consistent with the theory that the hydrogen atom occupies a lattice site at the center of a regular tetrahedron of titconium atoms. De basic consequences of this model, which have been experimentally verified, are that l l
- A. W. McReynolds, M. Nelkin, M. N. Rosenbluth, and W. Whittemore, Neutron
! Thermalization by Chemically Bound Ilydrogen, General Atomic Report GA-349. l l l l t
- 1. Neutrons with energies less than hv canaot lose energy in collisions with zirconium hydride.
- 2. A slow neutron can gain energy hv in a collision with tirconium hydride with a probability exp (-hu/kT), which increases very rapidly with temperature.
Since hv >> kT, zirconium hydride is not effewiive in thermalizing neutrons but can speed up neutrons already thermalized by water by transferring to them a quantum of energy hv. An increase in fuel temperature will result in a relativel) large decrease in reactivity. This large prompt negative fuel temperature coefficient results from the following effects-
- 1. The uranium in the fuel elements is approximately 20% as:U and 80%
***U. The capture resonances in as U are Doppler broaden *d by an increase in fuel tempirature. This causes a decrease in the self-shielding of the resonances and therefore an increase in the resonance capture probability. The temperature coefficient calculatad f this effect is -2 x 10-8/*C, and is expected to be nearly independent of temperature. !
2, When the fuel temperatu;e it. creases, the zirconium hydride temperature follows it essentially instantaneously. This increases the probability of previously thermalized neutrons gaining energy hu from the hydrogen atom vibrations. The increased fraction of speeded up neutrort results in an increased thermal and epithermal leakage from the core and increases the relative number of neutron captures in the nter. The reactivity to-hydride-temperature ratio should vary as t f
exp (-hu/kT). This corresponds to a temperature coefficient which doubles between 10* to 450*C. Experimental values of the prompt temperature coefficient have been obtained from steady-state reactivity compensation experiments on the prototype reactor. These experiments indieste that the prompt temperature coefficient varies from about -7 x 10-'/*C at a fuel temperature of 20*C to about -9 x 10-8/*C at 100*C, An increase in coolast (bath) temper 3ture alone results in a slight increase in
. reactiv ty in the University of Arizona TRIGA. The isothermal temperaturo coefficient is slightly positive adding 40 cents of resetivity when the water temperature is raised from 5'C to 30*C, and it is negative above 30*C
- Reactivity effects associated with water temperature will have essentially no effect on either the normal operating c'a aracteristics or the transient bebvior of the reactor because the fuel temperature coefficient is negathe and much larger in magnitude.
C. Auxiliary Systents Cooling System The reactor fuel is cooled by natural convection of the pool water. liest is removed from the pool water by a 7.5-ton vapor compression refrigeration system utilizing Freon-22 as the refrigerant, The compressor and air-cooled condenser for this refrigeration system are located on a concrete slab outside the building and are pictured in Figure 3.8. The Freon is circulated through an aluminum evsporator (cooling coils), located in the water as shown in Figure 3.1. The evaporator is made of two concentric coils of aluminum tubing in an array approtimately :i-l/2 ft ID,
- G. D. Spriggs and G. W. Nelson Exterimental Determination of The Total Isothermal Rescrivity Feedback Coefficient for The University of Arizons TRIGA Research Reactor,1RIGA Owners Conference IV Salt Lake City, March 3,1976 Report TOC-7, Gener.: Atomic Company, San Diego,
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I i i i i i < i 1 I ' I J 1 ,I l r i I i i J l Figure 3.8 UARR Vapor Compression Refrigeration Unit ' i i [ f i
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9 in. th!ck, and 4 ft deep. This shape provides cicarance for the removal of reactor core components if necessary. The neutron ond samn a Ouxes at the location of the evaporator are too low to either produce appreciable activation or radiation-induced decomposition of the Freon. I ne now of the woling water is downward along the outer radial region of the pool from the evaporator to the bottom of the reactor tank, up through the reactor core, and the central region of the pool, then back to the evaporator by natural circulat!on. Purification System The purification system (Figure 3.9) consists of a pump, fiber cartridge filter, mixed-bed type deminer-lizer, nowmeter, and surface skiramer connected by piping and valving. A probe for measuring the conductivity of the water is located ) in the reactor pool. t ] The major purification system components are described below. ' Purification System Pump. The water system pump is of a centrifugal type and is driven by a directly-coupled induction motor, c Filter. The filter element is a fiber cartridge that removes insoluble < j particulate matter from the reactor water system. The cartridges are replaced when l they beccme clogged, rather than being back-Oushed and reused, j Demineralizer. The function of the demineralizer is to remove soluble impurities frem the water in order to maintain the conductivity of the water at a sufficiently low level to prevent corrosion of the reactor components and to prevent activation of impurities in the water. The demineralizer is of the mixed-bed type, which removes both positive t and negative ions from the circulating water. The negative ions are replaced by c hydroxyl(OH) ions e.nd the positive ions by hydrogen (II) ions. The Oli and H ions combine to form witer. Consequently, any ions in the water are concentrated on the l
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resin and replaced by pure water. Elements dissolved in the water which might otherwise absorb neutrons and become radioactive are therefore concentrated in the resin bed. D. Nuclest Design r
- 1. handard TRIGA Core l De design and opersting chsracteristics of standard TRIGA cores are well I known, as is the inherent safety characteristic of this type of fuel. The first standard core loading of the University of Arizona TRIGA consisted of 61 aluminum-cisd fuel elements. As the maximum steady-state power level was upgraded and in-core experimental facilities were added, additional fuel ws:
required. When the reactor fuel wts replaced by stainless-steel-clad fuel in February,1973, a greater number of these f.nel elements wts required to achieve i criticality due to the artster thermal neutron absorption of the cladding and U-235 ; burnup during previous use of the fuel. With all availsble fuel space utilized, core l S-21 (Figure 3.10) allow 1 use of the pneumatic transfer s>1 tem and the Fir, and still [ has sufficient excess resetivity to reach criticality with the transient rod fully inserteu. Table 3.1 lists the operating characteristics of UARR standard core S-21. Table 3.2 lists the pulse mode sharacteristics of this core configuration. I l l l l f i l ( I I l
PNEUMATIC TRANtTER SYSTEM
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Table 3.1 Operating Characteristics of UARR Standard Core S-21 , Maximum Steady State Power Level 100 kW j Core Mass 2940 grams U-235 l Minimum Critical Mass 2653 grams U-235 , Total Control Rod Worth $9.48 Exem Reactivity [
$2.70 :
Shutdown Marsin at One Watt 56.78 Shutdown Marsin With Hishest [ t, Worth Contrel Rod Withdrawn 52.84 - l Power Defect (100 kW) $0.78 l Maximum 5:endy-State Temperature (100 kW) 109'C i Prompt Negative Temperature Coefficient .$016/*C Maximum Pulse Temperature (52.44 pulse) 309*C , Maximum Pulse Reactivity lasertion 52.44 j I Table 3.2 I Pulse Mode Characteristics of UARR Standard Core S-21 Reactivity Transient Rod Energy Maximum ! Insertion Withdrawal Peak Power Release Temp ( ($) (laches) (Mw) (Mw-sec) ('C) ; l.25 6.47 26 5.7 104 1.50 7.14 85 8.7 156 1.75 7.80 182 11.2 193 2.00 8.47 328 13.9 235 2.25 9.16 $07 16.3 277 2.44 9.72 690 18.4 309
'2.83 - 1100 25.5 380
'3.13 - l488 29.6 435
- Extrapolated Calculations The Limiting Safety Sptems Setting (LSSS) in the pulse mode is 1100 Mw. The extrapolated calculation at the bottom of Table 3.2 shows that a pulse of reactivity input necessary to reach the Limiting Safety Sprem Setting in pulse mode operation
' t 29 -
will not permit the nnxbum fuel temperature to exceed 1000*C. which is the Safety Limit as specified in the facility Technical Specifications. The LSSS is l protected by a taechankul block attached to the transient rod drive system that ilmits the withdrawal distance of the transient rod to a reactivity insertion of less than
$2.50.
- 2. Er.ternal Neut.vn Source -
I The start-ug. source used in the UARR is a 4.7 curie americium-beryllium neutrott source dou >ly enecpsulated with 0.045cch type M4 stainless-steel. The i source is attached to a threaded stud at the end of ar, a!utninum rod approximately la inches long. The top of this rod IAs a 5 inch 4!vminum disk topped with an end fitting 1:!:ntical to that of e, standard TRIGA fuel eteraent. The standard end fitting allows um of the TRIGA fuel handling tool for positioning LM wurce in core when l i necessary. The source is nottasily positiot:ed outside the graphde reflector in the southeast quadrant of the cere. From this position, sufflcient seutron count rate is present on the startup <hannel t2 satisfy the minirnum countrate specified in the l i facility Technical Specifkations. I f E. Therrnal Design f r The UARP opentes at 100 kW thermal steady state and is cooled by natural h convection. Cooling water enters the core at the Lottorn and lenes at the top by l t natural circulation. A 7.5 ton vapot compressbn refrigention sptem is capable of I remosing about 25 kW of heat frort the reactor pool. At inaximum steady state power (100 kW thermal), the net pciot temperature rise is et cratimatei" 4*C per f t hour. Administratise procedures limit the maximum reactor pcol water temperature i to 45' C. i I t n
IV. FACILITY CONSTRUCrlON A. Nuclear Resetor Laboratory 1 t ne University of Arizona Research Reactor is in the Nuclest Reactor Laboratory, w'4leh :. located on tk first Door of the north wing of the Engineering , Building. The Engineering Building is of brick and reinforced concrete construction, including most noors and ceil:ngs. Three adjacent rooms in the building are permanently established as the Nuclear Reactor Laboratory and are designated a cortrolled access area. These are a) Room 122, the control room, containing the control console and readouts for the I area monitors and water activity monitor, b) Room 121, t;.e reactor room, containing the TRIGA reactor, the particulate air monitor, four locked Door pits for storage of radioactive materiaa, one unlocked Door pit, one Door pit whleh contains a 1.5 curie cobalt 60 source for Icw-dose-mt Irradiations, and a contilever hoist for moving shielding plugs and the beam stop for the neutron radiography tube, c) Room 124A, l 2 the experiment set-up room, contains storage space for reactor laboratory and reactor calibration equipment, shielded storage for irradisted samples, a naturst-uranium, light water suberitical assembly, and a large graphite assembly used for laboratory experiments involving neutron Out mapping. l Room 216 is the room directly abose the reactor room which was originally j t L
- used to receive a beam of neutrons from the reactor. A 9 inch diameter hole (
- penetrates the Door of this room directly above the center of the re:ctor core, and a 30-inch by 36 inch hatch to the roof abose is directly oser the hole. Little use was i 1
made of this beam capability, so during the refurbishment of the reactor in 1972, no provision was made in the new bridge for a hole to accommodste the beam titbe.
- At this time the hole in the Door is espred and locked, and the room is used for i
storage of reactor supplies and departmental records. L f i i
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FiFure 4.1 Section of North Wing of the Engineering Building Shoving the Reactor Laboratory
I s All three rooms on the first floor and room 216 on the second Door are part of the Reactor Laboratory security area and have security locks. Figure 4.1 shows a section of the first floor of the Engineering Building and the locatioe of rooms 122, 1:4.and 1:4A la the Nuclear Reactor Laboratory. i B. HVAC . The Engineering Building, including the Reactor Laboratory, is air conditioned by forced convection air handlers. The chl!!ed water and heating steam circulated through the air handlers is provided ~oy a refrigeration and steam plant. The Reacter Laboratory is provided with a window-mounted exhaust fan with a minimum flow rate of 1000 cfm. When an airborne contamination alarm is generated by the continuous air monitor, the normal exhaust fan will stop and as emergency exhaust fan will start. In this mode air u drawn through a high efficiency particulate air filter and is exhausted via a vent on the building roof located a minimum of 50 feet above the level of the ground surrovsding thi Engineering Building. C. Utilities Potable t .er for the Engineering Building is provided by the uni ersity water system whose source is 13 deep wells on campus. The university water system is backed up by the City of Tucson Water Utility. Demineralized water for makeup of the reactor pcol is either purchased or distillesi in a separate facility and carried i to the Reactor LaNratory in bottles, f Electric power for the facility is provided by the unhersity electric power I distribution s>1 tem Ahich is supplied by the Tucson Electric Power Company. Transformers toested outside the building step down the campus distribution frcm f I 4160 v to 10/200 v. ( l [ r i
D. Fire Protection The Engineering Building and its fire protection system conform to national and Arizona fire protection codes. The fire mah, pressurized by the university water system and backed up by the City of Tucson water system supplies the sprinkler systems and hose stations in the building exclusive of tht Reacter Laboratory. Foruble fire extinguishers located in the Engineering Building hallways are of the preuurized water and carbon dioxide types. '.?arbon-dioxide portable fire i extinguishers are mounted in the Reactor t.aboratory. Fire protection s>1 tem monitoring panels are mounted on the north wall inside the main entruce, and outside on the south side of the building. V. FACILITIES AND EXPERIMENTS l A. Pneumatic Transfer Sptem The pneumatic transfer system (rabbit system) Figure 5.1 is used for neutron ! irradiation of single sample *. It consists af two tubes connecting a se.sder receiver ' i ttation in the F ring of the reactor core with a terminal station on the reactor j l bridge. The sample is psekaged in an irradiation capsule which moses in one tube. ! A blower provides the prtssure difference for mosing the capsule. . B. Rotary Specimen Rack The rotary specimen rack (lary susan) is used for neutron irradiation of i many samples sirnultaneously. In this facility forty eventy spaced aluminum cups ; i which serve as holders for irradiation capsules are located in an aluminum ring which can te rotated around the core. The ring can either te drisen by an electric motor or rotated manually from the top c( the reactor pit. Any cup can be aligned i i with the isotore-remos21 tutt (lary susan access tube) (cr inserting and removing specimens. An indetitig device is provided to enable positioning of the cups. L i The rotary specimen rack is completely enciesed in a welded aluminum bot. t k The aluminum ring is located at approximately the lesel of the tcp grid plate. The i l I 1
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l l h l specimen cups eatend from the ring down to about 4 in, below the top of the active l lattice in the radial direction, the centers of the cups are about 13 in from the center of the core. Since the boa enclosing the rotary specimen rack introduces a i considerable void into the graphite reflector, this boa has been designed to ensurs that it will remain watertight. Flooding of this told will increase the reactivity of f L l the reactor slightly, i i t C. Fast irradiation Facility, Neutron Radiography Tube, and l Graphite Thermalizer Block i in addition to the thermal neutron irradiation facilities descriled above, i additional facilities are available which were designed and constructed at the i Unive.sity of Arizona and thus are not standard with other TRIGA reactors, ne 'Mr' is a fast neutron irradiation facility which is inserted in a standard ! fuel element position. It is a tube whose terminus is lined with boron, cadmium, ! i and gold, to absorb for thermal neutrons while permitting fast neutrons to pass into l the sample. Samples are lowered into the tube in an aluminum can on a nylon cord, i The F!r permits studies of fast neutron damage to materials without inducit.r. the radioactivity caused by absorption of thermal neutrons. Similar tubes without the j t tining of thermal neutron absorbing materials are used for incere irndiation of single samples and to contain neutron detect $rs for esperiments measuring reactor parameters. i ne neutron radiography tube is a 12' diameter beam tube which permits the streaming of a near. parallel, low.flut team of thermal neutrons from the reflector ! area of the reactor. His beam is ased to expose neutrun radiognphic pictures. When not in use, a scattering blak and two shield plugs prohibit neutron or gamma radiation from reaching the surface of the reactor pool. De graphite thermalizer block is enclosed with a water tight coser of ' I 1/8 inch 6061 aluminum and permanently positioned west of the reactor core. The [ I i
s
- 36 Inner and outer curved sudace of the block match the radil of the graphite core redector and reactor pool, respectively, such (n- i stock occupies the avaihble space between the core redector and the pool wall. The block is the same height as the core renector and has three aluminum Irradiation thim' oles set into its top surface such that the bottom of the thimbles are approximately 5 inches below the vertical J
etnterline of the res tor core. The centerlires of the t'11rnbles are in a radial line i due west of the reactor core approximately 4,7 and il inches from the edge of the i graphite core renector. To access either of the thimbles, an access tube with O-ring l j seals is slipped into the thlmble from abow the pool surface. The tube is then pumped dry, allowing sample access to the thermalizer block. ] 3 D. Central Thimble The central thimble is provided to permit irradiation of experiments at the ; t center of the core in the region of maximum neutrcn Out. It can also te used to provide a highly collimated beam of neutrons and gamma rs>1 when it is emptied of
- witer. The thlmble is an aluminum tute with an inside diameter of 1.33 inches. It ,
l ! 4 entends from the top of the tank through the two grid plates and terminates in a ) plug at a point approximately 7.5 laches below the lower grid plate. The tute is l normally filled with water, but the water can t< replaced with alt or a gas such as l helium so it can serve as a team tube. l l E. Demountable Fuel Element i l The demountable fuel element is a special facility designed for Out mapping , and reacter parametet determination within a TRIGA fuel element, it consists of one inch long,1.435. inch diameter sections of Zrit with approximately 12 grams of uranium, enriched to 19.99 percent in U.235. The demo.intable fuel element has ! I graphite and renectors and is clad with the same stainless steel material as standard f r i TRIGA fuel, The top of the demountable fuel element may te unscrewed in order l 4 l to install or remove fohs between the fuel sections. In order to limit personnel l. I I i
... . - - . .. . . .=
exposures due to handling foils or fuel sections, procedures limit the product of power times time for Irradiation of the demountable fuel element in one day. The i demountable fuel element is normally stored in a locked storage pit, and is used only as approved by a Senior Reactor Operator. F. Fuel Storage Racks nirty-position fuel element storage racks on the floor of the reactor pool and 13-position holsters on the sides of the pool provide storage of fuel during fuel inspections and approach to critical experiments. VI. CONTROL AND INSTRUMENTATION A. Centrol System Summary The operating sud protection system instmmentation for the reactor, exclusive of the sensors
- and actuators, is contained in a console shown in Figure 6.1. Figure 6.2 shows the basic control and instrumentation system diagram. The controls and information displays are located on the console. There is stea on the desk surface to allow logging of data during operation.
'A single preamplifier is remotely located on the reactor bridge.
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_ _ _ _ _~ SIMPLIFIED REACTOR CONTROL SYSTEM DIAGRAM y - "M" Y 'D" IT"#l s e 1,.- j l--l l-l _ l = l l = l !.=.y p
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The left-hand vert' cal section of the console contains meters which display a percent-power channel (left safety channel), the wide innge log-power channel, and the reactor period. Visual annunciators on this panel below the meters indicate the status of the protection system components mounted in this section of the consale. The center vertical section of the console contains a dual-pen strip chart recorder with annunciator windows on each side. Below the recorder are the rod position indicators. The mode control switch is located to the left of the position ) indicators; the flux demand control, part of the regulator system, is located to the I right of the position indicators. The lower center section mounted on the sloped desk portion of the console contains the mariual rod control switches in the center, To the left of the rod control switches are the transient rod control switch, key s-vitch, and immedintely L below it, the power-on indicator switch. To the right of the red control switches is the multirange linear picosmmeter range selector switch. i The right vertical section of the console contains meters which display a percent-power channel (right safety channel), the fuel temperature channel, and the pool water temperature channel. Visual annunciators on the panel below these meters indicate the status of the protection system ccmponents mounted la this section of the console. Contg 1 The control system consists of the following basic subsystemr i
- 1. Multirange linear pleosmmeter 1
- 2. Log channel and period circuit. (Wide mnge counting and ;
Campbelling channel) j
- 3. Flux regulator l
4 Rod control I i i l _- . . ~ . ._.._._ _ - ,______ _ . _ _ _ , _ . _ , _ . , _ _ _ . . __ ,__.__, . . _ . _ _ _ - , -. . . - - ~ . _ ,
- 5. Mode contrel
- 6. Interlocks Multirange Linear Picosmmeter. The multirange linear picosmmeter is a de current measuring instrument specifically designed for use with nuclear reactors. The channel receives an input signet from a compensated ion chamber located in the neutron flux in the water outside the reflector. The linear picoammeter input stage feeds an operational amplifier which produces an output voltage that is pro':ortional to the input current. Full scale output of 10 volts may be obtained with input currents between 10-18 amperes and 10-8 amperes in fifteen sanges. Test and calibration signals are built into the channel and may be selected from tte range selector switch. The multirange linear picosmmeter is located beneath the rod i control panel in a shielded aluminum box. The linear amplifier, range switch, ranging resiston, and output buffer amplifier are constructed as an int gral unit and may be removed for maintenance if required. The output of the multirange linear channel is displayed on the "red pen" of the dual-pen recorder. This output is also input to the flux regulator as a feedback signalin the automatic mode of operation.
Log Channel. The wide range log power channel measures power over the complete neutron flux range from below source level to above full steady-state power level. It receives its input signal from a fission chamber. The output signal is displayed both on the green pen
- of the dual-pen chart recorder and on the wide range log power meter. The channel uses a pulse log-count-rate iechnique for the lower five decades of the range and a log Campbell technique for the upper five decades to produce a single output signal for over ten d-cades. Neither technique is strongly affected by the gamma ray background of the TRIGA. Test and calibration circuits provide appropriate signals into the input of the preamplifier for checking six calibration levels. The log channel contains a period (rate of change of power) circuit with front panel meter display and bistable trip units, which are used for control system
- ~.
interlock functions. One bistable trip acts as a rod withdrawal prohibit (RWP) if the source level as detected by the channel is below a preset minimum. The other bistable trip prohibits pulse operation if the reactor power level is above a specified minimum level. Flux Regulator. The flux regulator is a feedback control system that maintains a constant value of neutron Dux at the compensated ion chamber or, during power level changes, maintains a nearly constant period during the transition from one 1 power level to another. 'The minimum period allowed by the flux regulator during
- power transition is set at 10 seconds. The desired power level is set by the operator on the aux demssd control located to the right of the dual pen recorder. The percent demand control determines the percent of range, while the multirange j
- picoammeter switch determines the range. The flux regulator maintains constant flux during routine operations and constant period during power level transitions.
- Rod Control. Figure 6.3 is a simplified diagram of the motor control circuit shown i
with the rod drive between its up and down limits. During normal operation, points M and N receive power through the normally closed control rod UP and DOWN pushbuttor.s. Depressing the UP button opens the direct circuit to point M and
- permits line current to flow through the DOWN pushbutton to point N and through the I pF phase shift capacitor to point M. This connection will cause the motor to rotate clockwise. Counterclockwise motion is obtained in a similar manner when the DOWN pushbutton is depressed. Simultaneous operation of both pushbuttons i
interrupts connection of ths motor to the line and the rod will remain in position or l slowly move downward, inserting the control rod into the core. Figure 6.3 also shows the rod position indicator lamps, their control switches, and the automatic drive down control circuit. Three indicators, which are part of { i the rod control indicator-switch assemblies, illuminate the UP light when the rod is l fully withdrawn, the DOWN light when the rod is fully Inserted, and the l l
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44 CONT light when the armature is in contact with the magnets. The individual lamps receive power from a 12V stepdown transformer through 500 series dropping resistors. Illumination is controlled by using the position switches to bypass current around the lamps. Mode Control. The mode control switch, which is mounted to the left of the recorder on the center panel, has four active positions and modes. These are AUTOMATIC, MANUAL, PULSE LOW, and PULSE HIGH. The switch assembly is divided into three isolated sections to assure the independence of the operation system and the two redundant portions of the protection system. The mode control activates relays in the three divisions of the system, and these relays modify the system configuration and logic to meet the requirements at the respective operation modes. The use of each of these modes is described in a subsequent section of this 1 report. Interlocks. Two operating system interlocks are provided. One is a rod withdrawal prohibit, RWP, to prevent rod withdrawal until a minimum signal is obtained on the wide range log power channel. This interlock prevents rod withdrawal if a minimum neutron source level is not present. The second interlock is a pulse mode permissive which requires the reactor power to be less than a specified maximum in order to activate the transient rod in the PULSE modes. Activation of the circuit enabling pulse operation is signified by illumination of the indicator !!sht marked READY. B. Protection System The protection system consists of two protective subsystems, the Reactor Power (Percent-Power) Channels and the External Channel. When any scram condition is sensed, the control rod magnet power is interrupted, dropping the rods into the core.
The magnet power supply is connected to the magnets through two redundant logic networks. The contact of each of a redundant set of protective channels is cc .- cted in each leg of the magnet circuit to provide a one-out-of-two scram logic. Scrams may be initiated by any.of the following:
- 1. Left hand safety channel (110%) in steady-state mode operation
- 2. Right hand safety channel (110%) in steady-state mode operation
- 3. Pool water level less than 14 feet above the reactor core
- 4. Manual scram bar
- 5. Magnet power supply failure
- 6. A safety channel switched to the "calibrate" or "zero" position
- 7. Power failure
- 8. Earthquake sensor
- 9. Timer after pulsing operation.
- 10. I10% Peak Power scram in Pulse Mode Reacter Power. The right and left safety channels derive their signals from uncompensated ion chambers. The output current of the chamber is applied to the input of an amplifier operating as a current-to-voltage converter. The output of the amplifier drives a front panel meter and a bistable trip unit with the trip set below the licensed reactor power. A mercury-wetted rehy contact of the bistable trip unit is connected in the magnet supply line and directly controls the initiation of protective action.
The amplifier output with zero input signal can be checked by switching the front panel switch to the zero position where the input is short circulted. A full scale calibration check is also provided to measure the current sensitivity of the instrument. A trip test control circuit can add a current to the incoming signal current to increase the amplifier output. The potentiometer portion of this control
allows the operator to adjust the magnitude of the added signal and to test the trip point of the bistable trip connected to the amplifier output. External Scram Channel. Two normally-closed-contact type input devices are connected to the external scram inputs. Dese are a timer after pulse scram and an earthquake scram. Activation of the external scram channel is annunciated by an indicator light located on the left drawer assembly. The visual Indicator will remain illuminated until it is manually reset by depressing the indicator switch. However, the indicator will not be extinguished if the trip condition still exists. Fuel Temperature Channel. The fuel temperature channel measures the temperature of the fuel by means of a thermocouple in special lastrumented fuel element. The output of the thermocouple is connected to an amplifier and the amplifier output j may be read on a meter. f Physical Arrangement of Protection Channels. One channel of each protection I subsystem and some control equipm(nt are grouped together and placed on each of two drawer assemblies. These two assemblies are located on opposite sides of the 1 console to obtain as much physical separation as possible within the limitation of the standard console. C. Reactor Operation i The following paragraphs briefly describe the cperation of the system and the configuration of the control and protection systems in the four operating modes: MANUAL, AtJTOhfATIC, PULSE LOW, and PULSE HIGH. l Manual Mode The manual mode can be used to control the steady state reactor power level from the source level to the maximum licensed power level. The wide range log f channel displays the reactor power level from below the source level to above full ) power. The output of this log channel is displayed by the ' green pen' of the dual pen strip chart recorder, i l
a Automatic Mode The automatic mode is identical to the manual mode except that the regulating rod position is controlled by a feedback control system to regulate the reactor power :evel as detected by the linear multirange channel. Aotomatic mode operation can be initiated at any power level detectable by the linear multirange channel. The setpoint for the regulator is determined by operator adjustment of the percent demand control located to the right of the strip chart recorder. The shim and transient rods may be controlled manually while the regulating rod is contro!!ed in the automatic mode. The protective subsystems in the automatic mode are identical to those in the manual mode. Pulse Low and High Modes The pulse modes are used to generate high peak fluxes or power levels for short periods of time. In these modes an interlock requires the reactor power level to be below a specified level in order to initiate a pulse. Prior to pulsing, the transient rod remains inserted and its cylinder is raised to the position which will produce the desired reactivity change when air is applied to the transient rod piston in the cylinder. After the pulse a timer initiates a scram sequence, dropping all control rods back into the core. The Pulse Low and Pulse High modes differ only in the range of peak reactor power displayed on the linear chart recorder. In the Pulse Low mode, full scale on the chart corresponds to 200 Mw. In Pulse High mode, full scale corresponds to is 1000 Mw. During the pulse, the peak power is displayed by the
' red pen' of the recorder, and the temperature at the center of the thermocouple-instrumented in one fuel element is displayed by the "green pen.'
~
After the transient rod fire button is operated, the power pulse is limiOd by the prompt temperature coefficient anc. di control rods are inserted by the timer after the pulse. No further operator control operations are required to scram the reactor. The recorder pens provide data to document information about pulse ' operation. D. Other Control Room Equipment In addition to the standard TRIGA control console, the control room of the University of Arizona research reactor contains wCmounted monitors of area . radiation, reactor pool water sctivity, and reactor pool water conductivity. A radiation monitor readout and chart recorder on the west wall of the control room provides reading of a radiation detector mounted on the emergency exhaust duct. i 1 E i 1 1 i i I J i 1 . j l I
.=,-
1 I
VII. SAFETY EVALUATION i A. Introduction This evaluation summarizes calculations, measurements, and references w' # .h show that the reactor poses no health or safety risks to the public either as a result . of normal operation or as a result of any credible accident. r The potential situations, either occurring in normal operation or as a nesult of an accident, which will be considered are: ,
- 1. Fuel element failure due to loss of reactor pool w3ter.
- 2. Radiation due to loss of reactor pool water
- 3. Release of fission products from a fuel element
- 4. Fuel element temperature lacrease as a result of large reactivity insertions r
- 5. Production of argon-41 in pool water and irradiation faci'ities
- 6. Production of nitroger,-16 .
- 7. Other radioisotopes in reactor -W 42ter B. Basis for Safety Analysis Calculations The following safety analysis calculations are based on standard techniques and data and on measurements made at the facility itself. Of necessity, some assumptions must be made about power history, release mechanisms, and dispersion f
u of radioactivity which cannot be known in advance. Where such assumptions are L necessary, a highly conservative approach has been used, so that the results of the t calculation will represent an upper limit of damage or risk, and the calculations will In most cases represent a risk which is much greater than could be obtained in any t real situation. The basis of the more signifiennt parameters, factors, methods, and : assumptions used in the safety analysis calculations are si/en below. . 1 F i i <
=
. So .
Maximum Fuel Temperature In a review 1 of the analysis of credible.ccidents for TRIGA reactors the fuel-moderator temperature is identified as the crucial safety parameter. The limiting temparatures are given as ll50*C for ZrH3 ,, and 1000*C for ZrH3 ,, TRIGA fuel. For this reason, for the Univusity of Arizona TRIGA, fuel temperatures must remain below 1000*C under any mode of operation. The fuel temperature limit ot' 1000*C will be the safety limit for the facility. Fission Product Retention Measurements at General Atomic 8 have demonstrated that a very small fraction of fission products are released from the ZrH fuel material during and after reactor operation. This release fraction depends upon the fuel temperature, and has the value of 1.5 x 10-8 for fuel temperatures below 350*C. Operation of the University of Arizona TRIGA in the steady state mode maintains fuel temperatures well below 350*C. For this reason, the release fraction value of 1.5 x 10-t is
- appropriate to 1,se for the calculations reported here, j Postulated Power History The license for the University of Arizona TRIGA Reactor is intended to t
permit legal operation of the reactor under all conditions permitted by the Technical Specifications. This will include operation at power levels not exceeding 110 kilowatts thermal. The Technical Specification limitation on steady-state reactor power will give a requirement for safety channels to scram the reactor with settings at a minimum of 110 kilowatts. It is clearly not intended that an occurrence where i
*Hawley, S. C., and Kathren, R. L., NUREG/CR-2587, Credible Accident Analyses for a TRIGA and TRIGA-Fueled Reactors,1982,page 15.
8M. T. Simnad The U-ZrH, Allog its Properties and Use in TRIGA Fuel, GA { Report E-Il7-833, February 198D. 1 4
0 *
,1 -
proper operation of the safety channels in shutting down the reactor at a power level at or below 110 kilowntu would be a violation of the license conditions. For the purposes of conservatism in calculation, the safety analysis will assume operation at a maximum power level of 110 kilowatts. Because the reactor is an educational facility, the historical operation pattem for the University of Arizona TRIGA involves mostly operation at low power for education purposes, with operation at the power of 100 kilowatts for one to three hours as needed a few times per month in order to perform neutron activation analysis for student research projects. The average annual power generation since the facility wv put in operation is 6262 kilowatt hours per year, equivalent to 62 hours of operation per year at 100 kilowatts power. The average during the five-1 year period 1983-1988 of 6410 kilowatt hours per year is only a slight increase above the lifetime average. For this safety analysis calculation, operation at a power !cvel of 110 kilowatts for 40 hours per week,52 weeks per year for 30 years will be postulated,. ( This would give an annual power generation of 228,800 kilowatt hours, about 36 I times the historical operation pattern. For the calculation of fission product releases, failure will be assumed to occur at the end of the eighth hour of operation ou the fifth consecutive day of operation. For the calculation of gamma-ray emission rates after loss of pool water, it will be assumed the fuel will be uncovered one hour after shutdown on the fifth day of operation. Power distribution among fuel elements The current core loading of the Ur.iversity of Arizona TRIGA contains 85 fuel elements and two fuel follower control reds so the power generation is distributed among 87 eternents. Measurements of the power generation distrioution in the reactor, accomplished by measuring temperature increases in varlovi fuel positions with the instrumented fuel element, have deraorstrated that the maximum
4 . l power generation rate is in a B-ring fuel element, and is 1.446 times that of the + average fuel element in the core. For (;omputation of fission product release, it will be assumed the release is from a B-ring fuel element and that this elernent will contain the fraction 1.446/87 of the total fission product inventory of the reactor
! l' Core. ' ' Dispersion of Fission Product Gases in Reactor Room .
If a cladding failure occurs in a fuel element during reactor operation, the ! fission products will bubble through 15 or more feet of water before reaching the pool surface. Measurements at the facility have shown that large air bubbles, with i volume of 50 to 200 cubic centimeters, released in the open pool at the depth of the l 4 reactor core will take about 12 seconds for the first air to rise to the surface cf the water. Small bubbles which are finely dispersed in the water will take up to 10 i seconds longer. It is expected that the bubbles squirting out of a crack in a fuel element would be finely dispersed, and also that their rate of rise would be slowed [ i 1 i somewhat by passing through the top grid plate. No information is available either h about the bubble size from a fuel element release or the time for the bubbles to ! l traverse the grid plate, so it is conservatively assumed that it tskes 12 seconds for I l fission product gases to rise to the surface of the water. l l Dispersal of fission product gases after leaving the reacto. pool will depend j \ ! j on air circulation within the room, the temperature of the pool water, and the size c of the bubbles. It will also depend strongly upon whether the doors covering the i i pool at floor level are open or closed. Since these factors depend on circumstances ! j which cannot be predicted, it will be assumed that a!! of the fission product gases lr I are uniformly distributed in the room air 12 seconds after fuel element failure. , 4 Beta particle and Gamma ray Dose Rates to Personnel and the Public , i ! It will be assumed that all of the fission product gases released to the reactor f room air will remain suspended, reduced only through air removal from the room by ; 1 . 1
the exhaust system and by radioactive decay. Doses to personnel and the public will result from radiation-emitting gases in the air. For this calculation, it will be assumed that the dose rate will be the same as the dose rate at the center of a hemisphere with the same volume and same fission product concentration as the postulated exposure situation. This is a conservative assumption because it r*sumes that the fission products are closer to the persons than the actual geometry, and does not take credit for shielding dn to buildings or other objects. For the calculation of dose rates inside the reactor room, the hemisphere will have the same volume as the reactor room (6500 cubic feet). For calculation of dose rates outside the building, the hemisphere will have the same volume as a parallelepiped with cross section equal to the cross sectional area of the building and with length equal to the wind speed times the time fission product gases have been exhausted. This time is assumed to start 12 seconds after fission product release under water. Wind speed is taken to be I meter per second. C. Fuel Temperature Rise from Loss of Coolant The maximum fuel tecnperature for the UARR at the licensed steady state power level (100 kW) h.u been measured to be 109'C and the maximum fuel temperature during the maximum credible inadvertent reactivity inwrtion (33.13) has been estimated to be 435'C. The fuel temperatures stated above are both considersbly lower than the fuel temperatures spe:lfied for similar operating conditions that were presented for two Instances of fuel element cladding failure analysis considered in the review of Credible Accident Analysis for TRIGA reactors.' In both instances, the analyses assumed a much higher power history than
*!!awley, S. C., and Kathren, R. L., NUREG/CR-2387 Credible Accident Analyses for a TRIGA and TRIGA-Fueled Reactors,1982, page 15.
1
is possiele with the UARR, and immediate loss of coolant. The review report concluded that damage to stainless steel fuel cladding due to loss of coolant while at the operating conditions and temperatures specified, did not pose a credible accident for a reactor with TRIGA fuel. In view of the lower power history of the UARR and the very low probability of an instantaneous loss of coolant, the fuel temperatures resulting from a loss of coolant at the UARR will not present a hazard to the general public from fission product releases due to temperature-induced stainless steel fuel element cladding failure. D. Radiation due to Loss of Reactor Pool Water Loss of pool water to the extent that the radiation shielding of the core is compromised will produce enhanced radiation dose rates directly above the pool and in ad.lacent areas. Calculations were performed to determine the dose rates from direct radiation and scattered radiation which would be present in such a case. After a loss-of-pool-water accident it would be necessary to enter the reactor room safely to begin refilling the pool and to make temporary repairs to any leak. The time interval from the last reactor operation to the loss of pool water is an important parameter in the calculation of dose rates. It is clear that no credible leak could drain the pool instantly. Even wita a leak rate of 50 gallons per minute, more than one hour would elapse before the core were uncovered. Thus the times for which calculations were performed were chosen as I hour,8 hours, and 24 hours after shutdown. The Dorst Wheeler model* and the postulated power history were used to calculate the rate of gamma ray energy release per unit time after shutdown. It wts assumed that all of tl.e gamma rays were of 1 MeV energy. This is a conservative
'Lamarsh, John R. Nuclear Reactor Theory, Addison-Wesley,1966, pp. 94-95.
l
i assumptbn, since the majority of fission product samma rays are of a lower energy ! and thus are more readily shielded. The calculation included geometric (1/r8) attenuation, and self-attenuation of gamma rays leaving the U-ZrH fuel material 1 using the Taylor
- formulation dose builduo factors. In order to determine the dose rates with the pool partially filled with water, attenuation of 1-MeV gamma ra)1 f
passing through various thicknesses of water, with Taylor dose buildup factors, was i included as an option in the calculation. l A fraction of the radiation from the reactor core striking the ceiling of the f j reactor room will be reflected back and produce a radiation field throughout the l ) i i room and adjacent areas. Although the fraction reflected could possibly be . 1 calculated, the particular eouinment available in the reactor room permitted measurement of the reflected radiation w th considerable mo,; accuracy than might I be achieved with a calculation. There are six 7-foot deep,12-inch diameter shielded storage pits in the , [ reactor room floor, which are used for storage of sources, irradiated parts, fission ' 4 chambers, and graphite-loaded reflector elements for the TRIGA. Installed at the t
- i bottom of one of these pits is a 1.5 Curie Cobalt-60 source which is used for low- ,
I I dose-rate irradiation of semiconductor devices. Measurements of the direct radiation !
+
i from this Cobalt-60 source with the shle! ding plus removed and the radiation ] f I reflected from the reactor room ceiling were used to predict the relation between the j direct radiation from the reactor core and its reflection from the ceiling. - l ) The direct radiation from the Cobalt-60 source was measured with a f t calibrated ion chamber, 'Ihe sesttered radiation w.v measured with a one inch
; Sodium lodide scintillat:on detector with a Canberra Omega-1 multichannel analyzer l L
'Tsouanidis N., Measurement and Detection of Radiation, McGraw-Hill,1983, p. !
154, i t )
operating in the pulse-height analysis mode. Near the source where the scattered i radiation was most intense, it was possible to measure both the dose rate with the calibrated ion chamber and as the count rate from the scintillation detector at the same location. This measurement was used to calculate a calibration factor between counts per second of scattered radiation and dose rate with the ion chamber. At : greater distances from the source, the scattered radiation was too low in intensity to be measured with the ion chamber, but it could be measured with the scintillation l r detector so dose rates could be computed using this calibration factor. i 3 The direct beam of radiation from the Cobait-60 source can be visualized as a cone whose solid angle is determined by the depth and diameter of the storage pit. i The source of scattered radiation is in the shape of a circle where this cone of direct i
. radiation strikes the ceiling of the reactor room. The source of scattered radiation l.
from the reactor with shielding water removed would also be a circle where the cone , of direct radiation from the reactcr core strikes the reactor room ceiling. Thus the ratio of scattered to direct radiation for the reactor is equal to the ratio of scattered to direct radiation measured for the Cobalt-60 souret times the ratio of the direct radiation scattering source areas on the ceiling. This ratio was calculated to have a ! value of 3.37 from the geometry of the two sources. The fraction of Cobalt-60 gamma radiation transmitted threugh the ceiling into the unoccupied storage room above the reactor was also measured, and this fraction is expected to be about the same for gamma radiation from the reactor. l Table 7.1 lists the dose rates due to loss of pool water one hour after the last e tractor operation with the postulated power history. It is clear that these dose rates f I would require the adjaceat rooms and hall to be evacuated. However, after actual ! I measurement of the doses, a reentry team could safely reenter to add water to ! i. restore some of the shielding and thea initiate a temporary repair of the leak. l t L.f
. ) l The calculated dose rates at all of the locations are reduced after the addition of pool water. With two feet of shielding water above the reactor core, the dose rates are reduced by a factor of 6.1; with four feet of water they are reduced by a factor of 14.7; and with six feet of water the dose rates are reduced by a factor of , 605. This indicates that the risk of radiation exposure is small, when even one-half [ of the pool water has drained. A leak of sufficient size to drain the pool would i develo:: slowly over time, and low water would be observed days or weeks before ! enough water was lost to compromise the shielding. Thus the probability of pool l l water loss immediately after operation is very small. : l l Table 7.1 Calculated Gamma-Ray Dose Rates with no Water I Shielding the Reactor at One Hour After Operation.' l At floor level above pool 83 rem /hr { i In room 216 above reactor room 6.2 rem /hr ; In reactor room 2 meters from pool 1.3 rem /hr i 1 In hall 4 meters from pool 0.26 rem /hr r
- These dose rates are further reduced by a facter of 2.4 after eight hours and by a factor of 3.75 after 24 hours.
E. Release of Fission Products from a Fuel Element During reactor operation, fission products will be produced within the fuel l [ elements at a rate proportional to power generation. Most of these fission products [ are solids at normal operating temperatures, c.i. cept the noble gases (krypton and < xenon) and the halogens (bromine and lodine). The latter, if not chemically combined, will vaporize at the highest temperatures achievable in an operating fuel element. In a TRIGA fuel element, most of these gaseous or vaporizable fission products remain trapped within the ZrH fuel material, but a small fraction will esespe and accumulate in the spaces between the fuel material and cladding. This
escaping fraction has been measured to be 1.5 x 10-8 for TRIGA fuel operating at low temperatures. If the cladding of the TRIGA fuel element fails, some of the rare gases and halogens which have accumulated in the spaces will escape and will bubble through the pool water to the surface. The amount of accumulated gas which escapes depends upon the location and size of the cladding failure and the gas pressure within the fuel element relative to its environment. , A computer program was written to calculate the production, release, and dispersal of noble gases and "bgens from a TRIGA fuel element. Production of all bromine and krypten radioisotopes with masses from 83 to 92 and all iodine and xenon radioiso: opes with masses from 129 to 141 was computed. Fission mass chain yields and isotope halflives were taken from the Chart of the Nuclides.8 As a conservative assumption, both the halogen and rare gas componenu were considered to be produced with a yield fraction equal to the sum of the individual fission yield fractions foe a!! isotopes with the same atomic mass number. The fraction of accumulated gas which escapes the fuel element is assumed to be 100 percent. It is also conservatively as:umed that there is no retention of rare gases in the water and that 90 percent of halogens remain in the pool water while 10 percent reach the pool surface and will be dispersed in the room. This is a highly ennservativa assumption, since in the Three Mile Island power reactor acciderit, over 99.99 percent of iodine did not become dispersed in air, but instead remained in solid or in water-soluable form.8 1 Chart of the Nuclides, General Electric Company, Knolls Atomic Power Laboratory, Schenectady, N. Y.,1977. 8 Report of the President's Commission on the Accident at Three Mile Island, John G. Kemeny, Chairman, Washington D. C., (October, l')79), page 31.
i l Fission product populations in the reactor core were calculated for the postulated power history. The fission products passing through the pool water were ! k assumed to be uniformly dispersed in the reactor room 12 seconds after release. L L Table 7.2 gives the calculated dose rates in millirem per hour due to fission product concentrations in air in the reactor room at various times after fuel element failure I with different exhaust fan rates. This table indicates that for all exhaust rates ! considered, the fission product concentrations will not present a significant hazard for persons who might make a short foray into the reactor room to mitigate damage i l to the facility or to assist persons to leave the reactor room. , l j Table 7.2 Dose Rates in Reactor Room in tisem/hr i due to F'ssion Product Concentratioet
! sfter B ring Fuel Element Failure.
4 e 1 Exhaust Fan Speed i Time After - Failure 0 cfm 100 cfm 500 cfm 1250 crm 15 seconds 1.66 1.66 1.66 1.65 1 30 seconds 1.37 1.36 1.33 1.29 60 seconds 1.08 1.06 1.01 0.92 L 4 300 seconds 0.55 0.51 0.38 0.22 i i > 600 seconds 0.38 0.33 0.18 0.06 l l 3600 seconds 0.19 0.08 0.00 0.00 ( I
- i 4
The dose rates due to fission products exhau.*ted from the reactor room were 2 calculated in the area outside the building assuming a wind velocity of I ( meter /second and assuming no filtration of the exhaust. The volume of dispersal
- was calculated as the vertical cross tectional area of the building multiplied by the ;
j wind velocity and multiplied by the time since first dispersed. The concentration of t i l 4 d I
60 - fission products was calculated to be the quantity of fission products exhausted to the air up to that time diviCw! by the volume of dispersal. The dose rate due to each radioisotope was calculated, and these rates were summed over all radioisotopes. Dose rates in the restricted area are greatest with a hw exhaust fan rate, while in the unrestricted area they are greatest with a high exhaust fan rate. Table 7.3 summarizes the dose rates in the unrestricted area for an exhaust fan rate nf 1250 cfm. For lower exhaust rates, the radiation dose rates are smaller. From these dose rate es!culations it is concluded that for the postulated fission produce release, the total radiation dose during the first hour will be less than 0.5 millirem in the reactor room and less thar. 0.005 millirem in the unrestricted area. Dus any of these fan rates, with or without filtration, would provide sufficient protection of both the restricted and unrestricted areas in the event of a fuel element failure. Table 7.3 Dose Rates in mrem /hr due to Fission Product Concentrations in Unrestricted Areas with Exhaust Fan Rate of 1250 cfm. Time After Failure Dose Rato (millirem / hour) 15 seconds .013 30 seconds .010 60 seconds .007 300 seconds .003 600 seconds .002 ; 3600 seconas .000 - F. Inadvertent Transient , The UARR is currently licensed for a maximum of $2.50 reactivity insertion during pulse mode operatioi. Tne UARR standard core has an excess reactivity of about $2.70, which is just sufficient to bring the reactor to critical with the transient - w K
rod fully inserted. If the pneumatic transfer system and the fir were removed from the F-ring and replaced with two new stainless steel clad fuel elements, this would result in an increase of $1.28 excess reactivity for a maximum total of 53.98. A willful act would be necessary to achieve this excess reactivity. As the removal of the two irndistion facilities and replacement with fuel requires administrative action, the increase of $1.28 excess reactivity cannot be considered as a contribution to a credible inadvertent transient. Ilowever, there are times that the UARR standard core is modified by the removal of the fir. Removal of the fir without filling the core position with fuel adds 50.43 excess reactivity. The maximum withd. wal of the tnnsient rod drive system is limited by positioning switches and by a meet.:'lical block such that the mailmum worth of the transient rod in any mode of operation is limited to 52.50. Because of the restriction of motion imposed on the transient rod drive system, the ..dess reactivity remaining on the transient rod below the mechanical block (approximately 50.93) cannot be considered during a credible inadvertent transient. Therefore. the maximum credible inadvertent tnnsient is limited to $3.13 ($2.70 + $0.43) from the cold, critical condition, or 52.35 (53.13 - 50.78 (power defect at 100 kW)) from the maximum licensed steady state power level. An inadvertent reactivity insertion of $3.13 (extrapolated from UARR pulsing data, see Table 3.2) would produce a maximum fuel temperature of approximately 435'C from the cold, clean, critical condition. An inadvertent reactivity insertion of 52.35 from the maximum licensed steady state power level would produce a maximum fuel temperature of approximately 400*C. In the event of either inadsettent reactivity insertion (the maximum exe.us reactivity available for steady state and pulse mode operation), the maximum fuel temperature will not exceed the 1000*C feel temperature Safety Limit.
e, Basec on the above data and operating experiences from similar TRIGA reactors, the maximum available inadvertent reactivity insertion in either mode of operation, while not allowable under the existing UARR IIcense, would not produce a situation that wculd be hazardous to the UARR operations staff or to the public. G. Production of Argon-41 in the Irndiation Facilities and Pool Water There are three regions of the reactor where production of argon-41 is of ; significance--in the pool water, the pneumatic tnnsfer system, and the rotary specimen rack. Two other facilities, the neutron thermalizer and the neutron radiography beam tube are sealed, and argon-41 cannot enter the air from them. The Fast Irradistion facility, Fir, is not sealed but it is a blind tube with only one entrance, so air can not norma!!y leave it in significant amounts while the argon-41 remains radioactive. In addition, thermal neutron absorption in argon 40 is limited in the Fir because it is shielded to stop thermal neutrons. Production rates of argon-41 in the University of Arizona TRIGA were experimentally determined by activation of the natural argon in air and reactor pool ; water. Two 1.85-mi!!iliter vials containing air at atmospheric pressure (approximately 706 mm li3 in Tucson) were irradiated in the rotary specimen rack (lazy susan) and pneumatic transfer system (rabbit) for 300 seconds with the reactor at 100 kilowatts. The resu; ting activity was measured in a Gell samma ray spectrometer system. The absolute efficiency was calibrated with NBS Mixed-Radionuclide Point-Source Standard SRM 4275-D-62. The point source standard and the two vials were counted at the same shelf of the gamma-ray spectrometer, and the spectrum of each was analyzed with the gamma-ray spectrum ar.alysis l
program, CINA,' giving absolute activity of argon-41 in microcuries at the time of counting. After correcung for Irradiation time, the time between irradiation and f counting, and for the air volume, the activation rates, R, were determined to be: for the lazy susan R = 0.514 microcuries/cc hr at 100 kilowatts, and for the rabbit R = 1.367 microcuries/cc-hr at 100 kilowatts. The propagated error in these activation rates, including counting statistics - for the standard and samples and the uncertainty in the standard, was about 2.0 ! j percent for the lazy susan and !.6 percent for the rabbit. 1 i Using the detailed construction dimensions of the facilities, the air volume of [ the laay susan was estimated to be 23 liters, and the air volume of the rabbit t terminus was estimated to be 0.22 liters. Thus the production rate of argon-41 in l [
! ' these facilities is !
l i in the lazy susan 0.514 x 23,000 = 14,400 microcuries/ hour and i L l In the rabbit 1.367 x 220 - 300 roterocuries/ hour at 100 kilowatts. l To determine the pieduction rate of argon 41 in reactor pool water, one-liter l bottles of water were collected at intervals during prolonged reactor operation at 100 1
- kilowatts for a power calibration exercise, and also l
for several hours after shutdown. i The absolute effkiency of the GeLi spectrometer system was determined for 1 l gamma ray emitters uniformly distributed in water in a one-liter bottle. This was done by irradiating a 0.11 gram pellet of NaOH in the reactor, and determining the j absolute activity of sodium-24 using the NBS SRM 4275 B point source standard. j The NaOH pellet was then dissolved and mixed in one liter of ordinary water in on.e l
- ' Nelson, George W., 'CINA. A Program for Complete Instrumented Neutron i Activation Analysis with a PC t)pe Minicomputer,' Journal of Radioanalytical and
! Nuclear Chemistry, Vol.114, No. 2,1987, pp. 231-236. I l I
}
l 64 - of the bottles, and the efficiency for a radioactive source dissolved in water in the bottle geometry was determined from a count of the 1364.5 kev gamma ray of Sodium-24 on the GeLi spectrometer system. The production rate of argon-41 in the reactor pool was determined to be 0.201 (+/- 3%) microcuries/ liter-hr with the reactor operatics at 100 kilowatts. At the time of this determination the pool water temperature was initially 7 degrees Celsius, and during 157 minutes of operation, it increased to 16.7 degrees. Water l samples at the surface of the pool and at a depth of I meter below the surface were coltected and capped, then surveyed, dried and packaged in plastic bags for transfer ! to an adjacent counting laboratory. From 5 to 10 minutes elapsed from the time of I collection to the start of counting. 'The pool water was stirred by a propeller f attached to a 1/2-hp drill motor during the heat up. At the end of 157 minutes, the l t reactor was shut down and the stirrer was turned off. Water samples were collected at the two locations for the following 170 minutes, with the stirrer turned off. l The water volume of the pool, corrected for the volume of the core and i 4 ref1ector, thermalizer, beam tube, and cooling coil pipes is about 16,220 liters, so the ] argon-41 production rate in water was determined to be 0.201 x 16,220 = 3 60 microcuries/ hour in pool water at 100 kilowstts. i While the growth of argon-41 activity in the water was used to calculate the production rate of argon 41, the rate of decrease of arson 41 activity at the surface ]i I of the water after operation gives a measure of the rate of transfer of arson from i f the pool into the Joom air, if N is the concentration of argon 41 in a sealed container (no possibility of leakage), the differential equation governing its time i dependence is i j di..AN de f where A is the radioactive decay constant of argon 41, i The solution to this equation is l l 4
-- ,--n-----,n~. . - - - - - - . _ . -.~--~~e,,-- vn
N = N. exp(-A t) On the ether hand, if there is transfer of argon out of the container, the argon-41 activity will decreate more rapidly. In this case the differential equation is df=-AN-CN where C gives the rate of transfer of argon-41 out of the container. The solution to this equation is N = N. exp(-(A + C) t) where, as expected, the rate of decrease would be greater than the decay rate with no transfer of argon-41 out of the container. The samoles collected after reactor shutdown were used to determine the coefficient of an exponential decrease in activity with time, in order to estimate the rate of transfer from the pool, C. The effect shou'd be most noticeable at the surface of the pcol, since only the top layer of water will exchange argon with the air.
, During 170 minutes after shutdown, ten water samples were collected from the surface of the pool (or within a few centimeters of the surface). The argon-41 concentration of the samples decreased during this time froc 0.278 to 0.099 microcuries per liter. The standard deviation of the relative activity of each of the water samples, based on counting statistles, ranged from 0.9 percent to 1.8 percent.
A least-squarts fit to an exponential function using the Los Alamos 'Deming* curve-fitting code' gave an exponentia' decrease in time with a time coefficient given by
.00627 +/ .00019 per minute = (1 + C)
This is to be compared with the expected value of the coefficient with no transfer _ = - -
*Rinard, P. M., and Goldman, A., A Curve-Fitting Package for Personal Computers, Report LA-Il082-MS, Los Alamos National Laborr. tory (Nosember 1987).
66 - of argon to the air, based on the 1.83-hour halflife of argon-41 of A = .00631 per minute. This indientes that, within the exp:rimental error C = 0 and there is no measurable transfer of argon-41 from the pool surface into the air. Release of argon-41 from the irradiation facilities is considerably less than the amount produced. Removal of samples from the lazy susan is accomplished with a solenoid-activated tool which is lowered into the access tube on a cable. Little air , is removed from the lazy susan itself when samples are raised up this tube except the volume of air residing within the irradiation capsule and sample itself. Removal i of cap ules is usually done the day after tne irradiation if the irradiation is done for one ho tr or more and at 100 kilowatts power. The argon-41 released from a susan irradiation capsule after a 3 hour irradiation at 100 kilowatts which is opene:113 hours aftei irradiation is less than 0.1 microcuries, which is not a significant quality. A different situation exists for the rabbit facility. Since the air in the rubbit sptem circulavs continuously through the irradiation terminus in the reactor core at any time the circulation blower is turned on, argon-41 will be distributed in this air and may be released to room air during blower operation or :.fter it is shut off. For the purpose of calculat ing argon-41 release, it was assumed that 100 percent of argon-41 produced in this facility will be released any time the blower is turned on. , ne historical experience has been that most rabbit irradiations are done at ! power levels of I to 10 kilowatts, and this facility is little used when the reactor is operating at 100 kilontts, ne reason for this is that the primary use of the rabbit system is for activation analysis of short halflife isotopes, and in the sample sires encount red, these isotopes would become too active for counting if they were irrsdu.ed at 100 kilowatts. A summary of all times the rabbit facility wts used during a 12-month period encompassing the maximum use of rabbit irradiations showed that the total power summed over all reactor runs in which the blower was
l
=
turned on at least once, was 150$ kilowatt hours, nis would correspond to 15 hours at 100 kilowatts (although in fact it consists of many more hours at lower power levels). The argon-41 produced in the rabbit system during this time would be about 4500 microcuries. In keeping with the conservative mode of analysis, this quantity was multiplied by 36, the ratio between the postulated operation history to the actual history. A calculation was perforraed to determine the dose from the release of 36 x 4500 microcuries = 162 millieuries of argon-41 in one year. The dc4e to persons in the reactor room was calculated assuming the room was an equal-volume hemisphere with radius 445 cm with the argon-41 uniformly distributed throughout it. Using the flux-to-dose conversion factor and mass attenuation factor for 1.293 MeV gamma rays, it was calculated that the whole body (gamma) dose for 162 mi!!! curies per year of argon-41 per year distributed in the air with the exhaust fan turned off is 45 millirem / year. With the fan operating si 500 cfm the calculated gamma dose is 1.3 millirem / year, and with a fan rate of 1250 cfm it is 0.5 millirem / year. The dose from beta particles, which will penetrate the skin to a depth of about 0.4 mm, would be 271 millirem / year without the fan in operation, and 7.9 and 3.3 millirem per year at fan rates of 500 and 1250 cfm. No significant risk is attributed to the dose rates with the fan in operation at either ' speed. A Technical Specification requirement will preclude operation of the reactor without the fan in operation, except at low power levels and for a short time. Dose rates outside the reactor facility due to 300 micorcuries per hour of - argon-41 exhausted with a wind speed of I meter per second were computed with the equivalent volume hemisphere model. Both the dose rate due to gamma rays and the dose rate due to beta particles were 0.00002 millirem / hour. This is less than one hundredth of the dose rate due to natural background radiation. Operation of the rabbit system for 540 hours per year at 100 kilowitts would gise a maximum dose L
outside the restricted area of 0.02 millirem / year. Nei*her the dose rate nor maximum annual dose presents a significant risk to the public. H. Production of Nitrogen-16 Significant quantities of nitrogen-16 (7.4 second half-life) are formed in the reactor by the (n.p) reaction of fast reactor neutrons on oxygen in water cireulating through the core. The threshold for this reaction is about 10,hteV. A thorough analysis of the pool surface dose rate and the reactor room dose rate du to the concentration of nitrogen-16 was perfor.ned by General Atomic in the Safety Analysis Report for the initial operating license of the UARR. This analysis was performed for 1000 kW operation and for a nitrogen-16 rise time based on the velocity of the water as it passes through the reactor core. hieasurements performed at the UARR have determined the rise time for nitrogen-16 from the top of the core to the surface of the pool to be 190 seconds.* Taking into account the decreased power level (100 kW) and the increased rise time (190 seconds), similar calculations predict the dose rate at the surface of the pool to be 1.6 x 10-* millitem/ hour and the dose rate in the reictor room air due to the nitrogen-16 concentration to be 2.6 x 10-8 millirem / hour. The predicted dose rates are well below those values which would be hazardous to the UARR operations staff or the public. I. Other Radioisotopes in Reactor Pool Water No radioisotopes could be detected in reactor pool water 24 hours after the reactor had been operated for 3 hours at 100 kilowatts. However, during the time cf reactor operation, five radioisotopes in addition to argon 41 could be detected with the Gell gamma-ray spectrometry system.
"Convective Heat Transfe; by N1' hiapping in the TRIGA hfark I Reactor,"
Hellend, R. T., hiasters Thesis, University of Arizona, (1971).
The radioisotopes listed in Table 7.4 were detected for a water sample collected 1 mete
- below the pool surface after 127 minutes of operation at 100 kilowatts. The quantities are given in microcuries per liter of pool water, corrected for the decay from the time of sample collection to the time of counting. The concentration of each of these isotopes in . vater is well below the 10CFR20 Appendix D limits and does not present a risk. Because of the short halflives of these isotopes, reconcentration in a biological or chemical system is impossible.
The tritium concentration in r-actor pool water has been measured by the University of Arizona Radiation Control Office to be 0.015 microcuries per liter. This is consistent with reported tritium concentrations measured for similar low-power research reactors.* This concentration is a factor of 200 lower than the MPC allowed for release of tritium in water to unrestricted areas (10CFR20 Appendix D, Table II, Column 2), and thus the tritium in reactor pool water does not constitute a safety hazard.
*Plappert, R. and McLveen, J. W., ' Tritium Survey of Pool-Type Research Reactors," !!calth Physics, Vol. 31 No. 2, (August 1976), pp.169-170.
Table 7.4 Isotopes Measured in Reactor Pool Water at 100 Kilowatts Power.
~
Isotope Activity Halflife uCl/ liter Aluminum-28 0.028 2.24 min Argon-41 0.306 1.83 hours Chlorine-38 0.026 37.2 min Magnesium-27 0.025 9.45 min Manganese-56 0.013 2.56 hours Sodium-24 0.004 15.0 hours l VIII. CONDUCT OF OPERATIONS The University of Arizona Research Reactor (UARR) has been lastalled for the purpose of (a) laboratory instruction b) research in Nuclear Engineering, and (c) providing a source of nuclear radiation and radioisotopes. A. Organization and Responsibility The facility is operated by the Nuclear and Energy Engineering Department. The Reactor I.aboratory Director, who must be a licensed fenior Operator, is responsible for the safe and efficient operation of the facility. He reports to the Head ot' the Department who in turn reports to the Dean of the College of Engineering and Mines. Figure 8.1 is an organizational chart 'or the University of Arizona Research Reactor. The operating staff consists of a Reactor Sup,rvisor and support technicians. The Reactor Supervisor is required to hold a Senior Reactor Operator license for the facility.
.. o i
COMGE RADIATION CONTROL l an M 'DE*R
- DT,AN I
i t i HEAD of NUCLEAP RADIATION l and ENERGY ENGINC.ENG CONTROL OFFICE t a DEPARTMDit
. I 4 .
, F i REACTOR [ COMMITTEE l i L I l REACTOR ,/ REACTOR HEALTH PHYSICIST LABORATORY / SUPERVISOR j 1 , DIRECTOR / a 1 i I
~
REACTOR OPERATIONS , I i > L Tigure 8.1 UARR Organisational Chart l t ! l I
~ <
.t.
ENVIRONMENTAL IMPACT APPRAISAL I. !bTRODUCTION This environmental impset appraisal is prepared in accordance with 10 CFR 51.45 as a part of the license renewal application for the University of Arizons Research Reactor. The reactor is a 100 kW eduestional re setor on the esmpus of the University of Arizona, Tucson, Arizona. A full description of the reactor facility > snd its construction and use are described in the Safety Analysis Report which is also a part of the license renews! sppliestion. II. PROPOSED ACTIONS It is proposed to continue operation of the University of Arizons Research Resctor at the same level of activity which has characterized the successful educational uses of the facility fcr nearly 30 years. No increase in power level is requested, and no additions or modifiestions to the facility are included with the request for renewal of the operating license. III. IMPACT OF Tile PROPOSED ACTIONS ON Tile ENVIRONMENT Contir,ustion of operation of the facility for educational purposes is expected to have a hight)- beneficial impset on the environment. These benefits are directly l secrued to the nsWn and the world as follows: a) provulns invalusble education to nuclest engineers 8.ntimately involved in the 1esign, construction, operation, and regulation of nuclear power resctors and other nuclear facilities, b) performing research with the facility in support of programs in hydrology, biology, geology, and other earth, life, and physical sciences l which provide knowledge essential to understanding, sensitivity, and informed decision making in environmental and resource utilization questions, and
2 c) providing eduestions! and resesich opportunities for students in the biological, physical, and life sciences who, as research selentists in the 1990's and into the 21st century will be involved in identifying, communicating, and solving environmental problems. The basis f,r the above benefits are given in the following sectic't. 3.1 Need for Education, Well-educated nuclest engineers are essential t) safe operation of nuclear power plants. Proper procedures and administration, based on proper understanding of the fund:.ments!s involved in their facilities, could have prevented the accidents at Three Mile Island, at Chernobyl, and at other reactor facilities. The laboratory courses in nuclest engineering using the University of Arizona Research Reactor provide students with an in-depth understanding of the processes in a nuclest reactor ana .dsc bridges the gap between textbook or theoretiest learning and a real system. This is a valuable contribution to their eduntion and to their ability to satisfy society's 1:eeds for the services of their profession. Educated nu, lear engineers must be a part of a team which maintains safe, economical operation and prevents reactor accidents. The environmental effects of a l reactor accident are relatively minor when siewed in comparison to the larger effect l of loss of the nuclear option for power generation because of adverse public reaction to a reactor accident. 3.2 Need for Nuclest Power Nuclear power is of major environmental significance because it produces thermal energy without release of carbon dioxide. The scientific evidence i demonstrates a strong correlation between carbon dioxide concentration in the I atmosphere and mean earth temperature. Increase of mean earth temperature can be expected to cause negative worldwide environmental changet resulting in hardship and suffering which will continue at an accelerating rate for centuries. At this time, l l _ _ - - - - - - -
3 only solar, nuclear fission, hydroelectric, and some renewable sources can produce , significant amounts of energy without adding to the carbon dioxide concentration of the atmosphere. All of these cources will be needed to satisfy the energy needs of the present world population. In the light of what is now known about the environmental effecu of power l production, the major environmental effect caused by the Three Mile Island accident is not the radioactivity released from the reactor, but the gases released by fossil fuel power plants which are built where nuclear power pbnts would have been built had the accident not caused loss of public and utility acceptance. The acid rain from these fossil fuel power plants will continue to damage environmentally fragile areas for thirty or more years in the future. The carbon dioxide from these plants will remain in the atmosphere for centuries. 3.3 Education as Purpose and Goal of the Facility Education, research, and public service are the stated goals of the University ] of Arizona and also of the Nuclear Reactor Laboratory. Of these, the greatest environmental impact of the University of Arizona Reactor is through its use for i education. Nuclear engineers must be educated to have a full understanding of nuclear reactors and the ramifications of nuclear engineering decisions. Educating , I nuclear engineers without a University reactor is like educating concert musicians without a University orchestra or educatlag journalists without a University newspaper. It can be done, but in each example the employer must remedy deficiencies in the education. Not all employers of nuclear engineers can do this, although the environmental risk of employing poorly-educated nuclear engineers is great, i l )
4 IV. ADVERSE ENVIRONMENTAL RISKS WHICH CANNOT DE AVOIDED Certain environmental risks cannot be avoided. Other situations, which pose environmental risks for other facilities are not present in the particular situation of this low-power educational facility. These are summarized below. 4.1 Nuclear Fuel Cycle The fuel cycle of the University of Ariwna Research Reactor does not present adverse environmental risk during the projected 30-year licensing period. The present fuel, consisting of 87 previously-used TRIOA fuel elements with 4.8 grams average burnup and six fuel elements and two control rod followers with little burnup are sufficient to operate the reactor for more than the license period. There is thus no need for obtaining new fuel or shipping or disposal of spent reactor fuel during the license period. 4.2 Radios,tive Waste Radioactive waste from the facility consists of paper tissues, plastic gloves, labmat, spent samples and standards from neutron activation analysis, and resins
. from the demineralizer system. For the five-year period from mid-1983 through mid-19?8, the average volume of radioactive waste was approximately 7 cubic feet annually and the activity about 60 microcuries annually. Compared to the quantity of radioactive waste collected by the Radiation Control Office from laboratories of other University and Arizona Health Sciences Center users, this is less than 1% of the total both in volume and in activity. This quantity of radioactive waste is not a significant additional environmental burden.
4.3 Release of Radioactive Gases A risk from release of argon-41 from the irradiation facilities has been calculated, based on a postulated power history which assumes more power j generation each year than has been generated for the entire 30-year past history of I the facility. These calculations are summarized in the Safety Analysis Report. This I
5 report also shows that .se dose rate outside the facility would be on the order of
.0004 millirem per hour if all of the argon-41 generated in the pneumatic transfer system were released continually. This dose rate is sma!!er than variations in natural background radiation for various sites in the area because of differences in altitude or soil types, and thus is not of environmental significance.
4.4 Transpcrtation The adverse effects for persons driving to and parking at the University to use the reactor for education and research are no different than the adverse effects of driving and parking to visit the University for less environmentally beneficial purposes. However, the presence of the reactor on the University campus also reduces the amount of driving or other transportation which would be required if a reactor at another location were to be used for these research and education purposes. 4.5 Heat Generation The heat generated by the reactor and exhausted by the refrigeration system, is about 7000 kwhr per year. This is insignificant in comparison with the heat exhausted by the refrigeration system providing sir conditioning to the Engineering Building, of about 2,000,000 kwhr per year. 4.6 Radiation Exposure of Personnel Radiation exposures of reactor users and the operating srtff have generally been too sm:,ll to be detected. A summary of radiation exposures recorded on film badges and TLD finger dosimeters for the past 10 fiscal years (through July 31, 1933) is provided in table 4.1. Each 1xposure which is measured is reviewed with the staff person involved as Imt of the ALARA program. 4.7 Environmental Radiation Exposure Thermoluminescent dosimeters (TLD's) are placed on the Engineering building and on other buildings as part of an environmental monitoring program to
6-identify whether increased doses might result from reactor operations. These dosimeters are replaced monthly and are analyzed by a certified organization. For 1987, the average exposure for the three dosimeters located on the Engineering Building was 10 millitem per year greater than control dosimeters stored in a lead pig, while for eight dosimeters placed on other buildings at the University the i average exposure was 15.8 millirem per year greater than the control dosimeters. ' ne same pattern cf slightly lower doses on the Engineering Building has been , observed each year since the initistion of the TI.D monitoring program. There is no noticeable correlation between the power genersted by the reactor in a month and the dose on the TLD's for that month. It is thus concluded that changes in local radiation dose due to the reactor are either non-existent or so small that these , changes are obscured by normst variations due to building materists, height, or local ! 1 l shielding from natural radiation. ! 4.8 Radioactivity in Reactor Pool Water There are no drains in the reactor room or control room. Radioactivity of ] reactor water is below the level of detection except for tritium and short halflife l isotopes which have been measured during prolonged reactor operation of several ! . t j hours. The activity of the short halflife isotopes is summarized in the Safety I l Analysis Report. The tritium concentration in reactor pool water is about 0.015 I microcuries per liter. This is low in comparison to the 3.G microcuries per liter of f tritium which is the MPC for water dischstged to an unrestricted ares (10 CFR 20 , Appendix B Table 11). The other isotopes measured are in low concentration, and their short halflives make biological or chemical - stion i...e sible. Although I release or loss of reactor pool water is prevented by 2.ctihy procedura and will be ! avoided, it is clest that this water presents no environmental hazard bcesuse cf radioactivity. I
7 V. ADDITIONAL ENVIRONMENTAL BENEFITS An additional benefit of the continued operation of the University of Arizona TRIGA Reactor is its value in reducing the amount of radioactive waste generated by some other laboratories in the University. Many biological ar.d l physical processes studied in student and faculty research programs use radioactive i tracers to derive transport data. In many such tracer studies, only commercially-l available isotopes can be used because of restrictions on the chemical form needed, r I Commercially-supplied isotopes will be selected to have long halflife c: higher initial activity in order to retain adequate activity after shipment to the University from the supplier. Because of longer halflife, experimental equipment using these isotopes will contain radioactivity which necessitates either cleaning or disposal of the equipment. Both the cleaning solutions and contaminated equipment contribute to the volume and activity of radioactive waste. In addition, commercially purchased isotopes, like other supplies, are often ordered in larger quantities than actually needed for a project. This is because, while the base price of an isotope order is large, the incremental price for a larger activity of the isotope in the same order is small. Thus stocks of extra unused long 141flife isotopes build up, which must [ t ultimately be disposed of as radiosctive waste. In some studies, however, short halflife isotopes produced by the Univenity reactor can substitute for longer-lived commercial isotopes. When this is possible, l l l waiting for a sufficient time will remove all radioactivity, and no radioactive waste l is produced either in the form of cleaning solutions or contaminated equipment. A typical example cf the waste reduction possible at the University is a completed l research project in Entomology which involved studies of lon fluxes in nerve signal transmissions in cockroach nerves. Without a reactor at the Ur'.ersity, these studies t l i i would have used sodium-22 with a halflife of 2.26 years. Instead, sodium-24 with a j 1 halflife of 15.0 hours was produced by the reactor and used is the tracer. This i
.g.
svoided addition of sodium-22 contaminated radioactive waste to the University's responsibility without adding sodium-24 contaminated waste. III. ALTERNATIVES TO TIIE CONTINUED OPERATION OF Tite REACTOR There is no comparable alternstive facility. The net result of not continuing to use the reactor for eduestion and research is a decrease in the quality of education of nuclest engineers and restriction of the eduestion which secompanies research projects for students in other disciplines. Some use cf reactors at other facilities may substitute for a portion of the eduestional benefit which would be lost. However, the personal risk of si: or automobile travel, the time consumed, and the cost will restrict the use of substitute reactors to a minimum. There are no suitable or more economical alternatives available which can satisfy the educational needs of nuclear engineers at the University of Arizors as described in the previous sections of this appraisst. V11. TIIE RELATIONSilIP DETWEEN LOCAL SHORT-TERht USES AND LONO-TERht BENEFITS The local short-term uses of the reactor laboratory involve eduestion ot' nuclear engineers, education of research se,,,,tists, and developing student and public understanding of nuclear reactors. The long-term benefits come from the contributions which these engineers and scientists make to the nation and the world in solving tha considerable problems of design, construction, and operation of nuclest and other power systems and the equally considerable problems of regulstias them. One cannot say in advance how many of the engineers and scientists who will be educated using the Unhersity of , 1 Arizona Reactor Lsboratory will contribute significantly to these problems in the I future, liowever, it can be stated unequivocally that for the pset 30 years graduates of this nuclest engineering program, which has included laboratory work with the reactor, have Nen and are now very great contributors to research, design, s: re
9 operation, and reguistion of nuclear power in the United States and in many other countries. The contributions of future gr.iduates can be expected to be equally significant. t 7.1 Irreversible Commitments of Resources Continued operation of the facility after renewal of the license is a commitment to continue its use but this is not an irrevenible ecmmitment. Changes in programs, including decommissioning, will be equally possible at any time in the i future. Vill. ANALYSIS ~ The University of Arizona TRIGA Reactor is an educational facility which 1 has positive benefits to the environment through quality education of nuclear i engineers and research scientists in the earth, life, and physicsl sciences. It has no signifiesnt adverse environmental impsets. Radiation exposures to persons outside the facility due to operation of the reactor are not significant even when related only to the variation in naturally-occurring radiation in the same geographical ares. 7 The reactor facility is already in operation, so no new capital funds are , l needed. It is an act of good stewardship of public funds to approve the license t i amendment so the public may continue to benefit from the past investment through l t the education of its future nuclear engineen and scientists. l [ The University reactor provides quality education, and qualified graduates of ! t !
- the program are sought by many employers. Former graduates are making ,
significant contribut.ons in design, construction, administration and regulstion. ! Future graduates are expected to make equally significant contributions. I a i IX. LONO-TERht EFFECTS ON THE ENVIRONhtEhT i When the reactor is decommissioned the ares will be returned to general use, j ! The small additional burnup of U-235 due to 30 more years of operation, which is i l I expected :o be about 0.1 grams per fuel element, will not significantly change the l c l l 1 1
4 future usefulness of the reactor fuel for other facilities. The fuel will be transferred either to another facility which can continue to use it in a reactor, or to a DOE l f storage facility. The choice will be determined by DOE It the time. Activated f materials will be packaged ano shipped for disposal through commercial disposal agencies. The long-term environmental effects of renewing !% operating licence for ! the faellity are expected to be insignificant, i l 1 ; j 1 i i t 4 i ' l I l I' 1 1 ! a l I 1 I I i l I I i t
Table 4.1 Annual Whole Body Exposure for Faculty Staff, and Students with Radiation Dosimetry at al.e Ur.*versity of Arizcas Department of Nuclear and Energy Engineering Number of Persons Receiving Penetrating Radiation Exposures Whole Body Exposures (mrem) Finger Ring Esposures (mrem) Fiscal Year 0-25 26-50 51-100 101-500 501-1100 0-25 26-50 51-100 101-50^ 501-1000 1978-79 2 l' la - - 2 - - - l' 1979-80 1 - - - - 2 3 1+1' - - 1980-81 - - - - - 1 I - - - 1981-82 - - - - - - - - - - C
?
1982-83 - - - - - - - - - - 1983-84 1 - - 1* - - - - 2* - 1984-85 - - - - - - 1* - - - 1985-86 - - - - - - 1+1' - l' - 1986-87 - - - - - 1 - - - - 1987-88 2 - - - - - - - - -
- These persons received exposures from radiation sources not attributable to the TRIGA reactor.
r 1
.. g SPEC 1/.L NUCLEAR MATERIALS REQUIREMENTS AND NEUTRON SOURCE REQUIPEMENTS
e Special Nuclear htsterials and Neutron Source Requirements for the University of Arizona Research Reactor The Special Nuclest histerial to operate the reactor is already licensed under Arizona License 10-24. The UARR operates using standa d 8.5 wt% fuel enriched to less than 20% in the isotope uranium-235. The SNht is of low strategic significance as defined in 10 CFR 73. (y).(2), and is exempt from certain physical protection requirements pursuant to 10 CFR 73.6.(a). We specifically request that this license permit The University of Arizona to receive, possess and use up to 3.5 kilogrsms of contained uranium-235 in TRIGA fuelin connection with operation of the reactor. Current inventory of Special Nuclear htsterials, present in the Reactor Laboratory or und:r the control of the Reactor Laboratory staff is as follows: 96 irradiated TRIGA fuel elements and fuel followers containing less than 3:00 grams of U-235, average enrichment less than 18 percent 4 uranium-lined fission chambers containing 6 grams of U-235 enriched to approximately 90 percent i experimental demountable TRIGA fuel element containing 34 grams of U-235, enriched to 19.9 percent item Description Staterial Content i Pu-Be Neutron 15.35 gram Pu source 4 Pu-Be Neutron 63.70 grsm Pu sources
! Pu-Be Neutron 16 gram Pu source (custody of .
Physics Department, University of Arizona) 1 Am-De Neutron 1.374 gram Am-241 source 1396 Savaanah 2493.59 kg normal River slugs uranium 1 Californis-25 5 micrograms source}}