ML20043D539
| ML20043D539 | |
| Person / Time | |
|---|---|
| Site: | 05000113 |
| Issue date: | 05/31/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-1390, NUDOCS 9006080200 | |
| Download: ML20043D539 (74) | |
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3 NUREG-1390 s
7, Safety Evaluation Report related to the renewal of the oaerating license for the TRIGA training and research reactor at the University of Arizona Docket No. 50-113 U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation May 1990
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AVAILABILITY NOTICE F r blications Availability of Reference Materials Citeo in NRC Pu il ble from one of the tollowing Most documents cited in NRC publications will be ava a W, Lower Level, Washington, DC The NRC Public Document Room, 2120 L Street, N sources:
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NUREG-1390 Safety Evaluation Report related to the renewal of the operating license for the TRIGA training and research reactor at the University of Arizona Docket No. 50-113 U.S. Nuclear Regulatory Commission Omce of Nuclear Reactor Regulation May 1990 fo* ase.q,,
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ABSTRACT This Safety Evaluation Report for the application filed by the University of Arizona for the renewal of Operating License R-52 to continue operating its research reactor at an increased operating power level has been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission.
The facility is located on the University of Arizona campus in Tucson, Arizona.
The staff concludes that the reactor can continue to be operated by the University of Arizona without endangering the health and safety of the public.
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3 t
TABLE OF CONTENTS
.P,.a21 ABSTRACT........
iii 1
INTRODUCTION 1-1 1.1 Summary and Conclusions of Principal Safety Considerations.......................
1-1 1.2 Reactor Description 1-2 1.3 Reactor Location......................
1-3 1.4 Shared Facilities and Equipment 1-3 1.5 Reactor History 1-3 1.6 Comparison With Similar Facilities.............
1-3 1.7 Nuclear Waste Policy Act of 1982..............
1-4 2
SITE CHARACTERISTICS 2-1 2.1 Reactor Site........................
2-1
- 2. 2 Demography.........................
2-1 2.3 Nearby Industrial Transportation, and Military F ac i l i ti e s.........................
2-1 2.4 Meteorology 2-1
- 2. 5 Hydrology 2-3 2.6 Geology and Seismology...................
2-3 2.7 Conclusion.........................
2-3 3
DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS..........
3-1 3.1 Reactor Facility Description................
3-1 3.2 Wind and Water Damage 3-1 3.3 Seismically Induced Reactor Damage.............
3-1 3.4 Mechanical Systems and Components 3-1 3.5 Conclusion.........................
3-3 4
REACTOR.............................
4-1 4.1 Reactor Core........................
4-1 4.1.1 Fuel Elements....................
4-2 4.1.2 Core Support Strn:ture and Reflector 4-5 1
4.1.3 Control Rods 4-5 4.1.4 Neutron Source 4-7 4.2 Reactor Tank and Biological Shield.............
4-7 4.3 Reactor Instrumentation 4-8 4.4 Dynamic Design Evaluation
~4-8 4.4.1 Excess Reactivity and Shutdown Margin........
4-8 NUREG-1390 v
9 4
L TABLE OF CONTENTS (Continued)
P.,aJL' 4.4.2 Normal Operating Conditions.............
4-9 4.4.3 Assessment 4-9 4.5 Functional Design of Reactivity Control Systems 4-10 4.5.1 Standard Rod Drive Assembly.............
4-10 4.5.2 Transient Rod Drive Assembly 4-10 4.5.3 Scram-Logic Circuitry and Interlocks 4-11 4.5.4 Assessment 4-11 4.6 Operational Procedures...................
4-12 4.7 Conclusion.........................
4-12 5
REACTOR COOLING AND ASSOCIATED SYSTEMS 5-1 5.1 Reactor Cooling System...................
5-1 5.2 Reactor Purification System 5-1 5.3 Primary Coolant Makeup System 5-3 5.4 Assessment.........................
5-3
- 5. 5 Conclusion.........................
5-3 6
ENGINEERED SAFETY FEATURES 6-1 6.1 Stack Vent Exhaust System 6-1 6.2 Conclusion.........................
6-1 7
CONTROL AND INSTRUMENTATION SYSTEMS...............
7-1 7.1 Reactor Control System...................
7-1 7.1.1 Control Console...................
7-1 7.1.2 Operating Modes.........,.........
7-3 7.1.2.1 Manual Mode 7-3 7.1. 2. 2 Automatic Mode................
7-3 7.1. 2. 3 Pulse Low and Pulse High Modes.......
7-3
- 7. 2 Instrumentation System...................
7-3 i
7.2.1 Nuclear Instrumentation...............
7-4 7.2.1.1 Reactor Scram System............
7-5 7.2.2 Nonnuclear Process Instrumentation 7-5 7.3 Conclusion.........................
7-5 8
ELECTRICAL POWER SYSTEM,......
8-1 8.1 Normal Power........................
8-1 NUREG-1390 vi i
b Q
P TABLE OF CONTENTS (Continued)
.P.agtg 8.2 Emergency Power 8-1 4
8.3 Conclusion.........................
8-1 9
AUXILIARY SYSTEMS........................
9-1 9.1 Ventilation System.....................
9-1 9.2 Fire Protection System...................
91 9.3 Communication System....................
9-1 9.4 Air Conditioning System 9-2 4.5 Fuel Handling and Storage 9=2
- 8. 6 Conclusion.........................
9-2 10 EXPERIMENTAL PROGRAMS......................
10-1 10.1 Experimental Facilities..................
10-1 10.1.1 Pneumatic Sample Transfer System 10-1 10.1.2 Rotary Specimen Rack 10-1 10.1.3 Fast Irradiation Facility, Neutron Radiography Tube, and Graphite Thermalizer Block 10-1 10.1.4 Central Thimble..................
10-2 10.1.5 Demountable Fuel Element 10-2 10.2 Experiment Review.....................
10-3 10.3 Experiment Reactivity Limitations.............
10-3 10.4 Conclusion 10-3 11 RADI0 ACTIVE WASTE MANAGEMENT 11-1 11.1 As Low As Is Reasonably Achievable (ALARA)........
11-1 11.1.1 ALARA Policy Statement 11-1 11.1.2 ALARA Commitment 11-1 11.2 Waste Generation and Handling Procedures 11-1 11.2.1 Solid Waste....................
11-1 l
11.2.2 Liquid Waste 11-2 11.2.3 Airborne Waste 11-2 11.3 Conclusion 11-2 12 RADIATION PROTECTION PROGRAM 12-1 12.1 ALARA Commitment 12-1 12.2 Health Physics Program 12-1 12.2.1 Health Physics Staf fing..............
12-1 12.2.2 Procedures 12-1 12.2.3 Instrumentation..................
12-2 12.2.4 Training 12-2 NUREG-1390 vii
TABLE OF CONTENTS (Continued)
.P, age 12.3 Radiation Sources.....................
12-2 12.3.1 Reactor.....................,
12-2 12.3.2 Extraneous Sources 12-2 12.4 Routine Monitoring 12-2 12.4.1 Fixed Radiation Monitoring System.........
12-2 12.4.2 Experimental Support 12-3 12.5 Occupational Radiation Exposures 12-3 12.5.1 Personnel Monitoring Program 12-3 12.5.2 Personnel Exposure 12-3 12.6 Effluent Monitoring....................
12-4 12.6.1 Airborne Effluents 12-4 12.6.2 Liquid Effluents 12-4 12.6.3 Environmental Monitoring 12-4 12.6.4 Potential Dose Assessment.............
12-4 12.7 Conclusion 12-5 13 CONDUCT OF OPERATIONS......................
13-1 13.1 Overall Organization 13-1 13,2 Training 13-1 13.3 Operational Review and Audits...............
13-1 13.4 Emergency Planning 13-1 13.5 Physical Security Plan 13-1 13.f Conclusion 13-2 14 ACCIDENT ANALYSIS........................
14-1 14.1 Fuel-Handling Accident 14-1 14.1.1 Scenario 14-2 14.1.2 Assessment 14-3 14.2 Rapid Insertion of Reactivity (Nuclear Excursion).....
14-3 14.2.1 Scenario 14-4 14.2.2 Assessment 14-4 14.3 Loss-of-Coolant Accident 14-4 14.3.1 Scenario 14-5 14.3.2 Assessment 14-6 NUREG-1390 viii
t 4
TABLEOFCONTENTS(Continued) l Paa!
14.4 Mi spl aced Experiments...................
14-6 14.5 Mechanical Rearrangement of the Fuel 14-6 14.6 Effects of Fuel Aging...................
14-7 14.7 Conclusion 1
14-8 15 TECHNICAL SPECIFICATIONS 15-1 l
16 FINANC'1.4L QUALIFICATIONS 16-1 17 OTHER LICENSE CONSIDERATIONS 17-1 17.1 Prior Reactor Utilization.................
17-1
.17.2 Conclusion 17-2 18-CONCLUSIONS............................
18-1
~19 REFERENCES 19-1 i
FIGURES
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2.1 Map of university of Arizona, Tucson 2-2 3.1 Elevation of Reactor and Pit 3-2 4.1 University of Arizona Nuclear Reactor Laboratory 4-3
4-4 4.3 University of Arizona TRIGA Fuel Element 6 5.1 Schematic View of University of Arizona TRIGA Reactor Water Purification System.......................
5-2 7.1 Simplified Diagram of the Reactor Control and Instrumentation Systems.............................
7-2 13.1 Administration Chart 13-2 TABLES 4.1 Principal Design Parameters...................
4-2 7.1 Minimum Reactor Safety System Channels 7-5 12.1 Number of Individuals in Exposure Interval per Calendar Year 12-3 14.1 Doses Resulting From Postulated Fuel-Handling Accident 14-3 14.2 Radiation Doses From Uncovered Core at the UATR Following the Maximum Loss-of-Coolant Accident 14-5 l
.1,
'NUREG-1390 ix
1 INTRODUCTION By letter (with supporting documentation) dated October 17, 1988, and supple-mented on July 17 and September 15, 1989, and January 30, 1990, the University of Arizona (UA/ licensee) submitted to the U.S. Nuclear Regulatory Commission (NRC/ staff) a timely application for a 30-year renewal of Class 104 Operating License R-52 (NRC Docket No. 50-113) for its TRIGA research reactor.
The licensee also requested an increase in the operating power level from the existing 100 kilowatts thermal [kW(t)) to 110 kW(t) to permit testing of the reactor or scram settings without exceeding its licensed power level.
The licensee intends to normally operate the reactor at or below 100 kW(t).
The licensee is permitted to operate the reactor within the conditions authorized in past license amendments in accordance with the Commission's regulations in Title 10 of the Code of r deral Regulations, Section 2.109 (10 CFR 2.109), until e
NRC action on the renewal request is completed.
The staff technical safety review, with respect to issuing a renewal operating license to the UA facility, has been based go the information contained in the renewal application and supporting supplements plus responses to requests for additional information.
The renewal application includes financial information, Safety Analysis Report, Technical Specifications, Emergency Plan, Environmental Report, Physical Security Plan, and a Reactor Operator Requalification Program.
This material is available for review at the Commission's Public Document Room at 2120 L Street N.W., Lower Levei, Washington, D.C., except for the approved Physical Security Plan, which is protected from public disclosure untier 10 CFR 2.790(d)(1).
The purpose of this Safety Evaluation Report (SER) is to summarize the results of the safety review of the UA TRIGA reactor (UATR) and to delineate the scope of the technical details considered in evaluating the radiological safety aspects of continued operation.
Thir, SER will serve as the basis for renewal of the license for operation of the UA facility at thermal power levels up to and in-cluding 110 kW(t).
The facility was reviewed against the requirements of 10 CFR Parts 20, 30, 50, 51, 55, 70, and 73; applicable regulatory guides (principally Division 2, Research and Test Reactors); and appropriate accepted industry stan-dards [American National Standards Institute /American Nuclear Society (ANSI /ANS) 15 series).
Because there are no specific accident-related regulations for research reactors, the staff has compared calculated dose values with related standards in 10 CFR Part 20, the standards for protection against radiation, both for employees and the public.
This SER was prepared by Theodore S. Michaels, Project Manager, Division of Reac-tor Projects-III/IV/V and Special Projects, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission.
Major contributors to the review were the Project Manager and R. Carpenter, R. Carter, and C. Cooper of the Idaho National Engineering Laboratory under contract to the NRC.
1.1 Summary and Conclusions of Principal Safety Considerations The staff's evaluation considered the information submitted by the licensee, past operating history recorded in annual reports submitted to the Commission l
NUREG-1390 1-1
5 by the licensee, and reports by the NRC Region V office.
In addition, as part of its licensing review of several TRIGA reactors, the staff obtained laboratory studies and analyses of several accidents postulated for the TRIGA reactor.
The staff's conclusions, based on evaluation and resolution of the principal issues reviewed for the UATR, are as follows:
(1) The design, testing, and performance of the reactor structure and systems and components important to safety during normal operation are inherently safe, and safe operation can reasonably be expected to continue.
(2) The expected consequences of a broad spectrum of postulated credible accidents have been considered, emphasizing those that could lead to a loss of integrity of fuel-element cladding.
The staff performed conserv-ative analyses of the most serious credible accidents and determined that the calculated potential radiation doses outside the reactor room would not exceed 10 CFR Part 20 guidance for unrestricted areas.
(3) The licensee's management organization, conduct of training and research activities, and security measures are adequate to ensure safe operation of the facility and protection of its special nuclear material.
(4) The systems provided for the control of radiological effluents can be operated to ensure that releases of radioactive wastes from the facility are within the limits of the Commission's regulations and are as low as is reasonably achievable (Al. ARA).
(5) The licensee's Technical Specifications, which provide limits controlling operationofthefacilityllbeoperatedsafelyandreliably.
are such that there is a high degree of assur-ance that the facility wi (6) The financial data provided by the licensee are such that the staff has determined that the licensee has sufficient revenues to cover operating costs and eventually to decommission the reactor facility.
(7) The licensee's program for providing for the physical arotection of the of 10 CFR Part 73 pecial nuclear material complies wit) the requirements facility and its s (8) The licensee's procedures for training reactor operators and the plan for operator requalification are acceptable.
These procedures give reasonable assurance that the reactor facility will be operated competently.
(9) The licensee has submitted an Emergency Plan that is in compliance with the existing applicable regulations.
This item is discussed further in Section 13.4 of this report.
1.2 Reactor Description The UATR is a heterogeneous, open tank-type TRIGA reactor.
The core is cooled by natural convection of light water, moderated by zirconium hydride and light water, and reflected by a combination of light water and graphite.
The core is located near the bottom of a 7-ft-diameter steel tank that has a wall thickness NUREG-1390 1-2
1 of k in.
Approximately 8 in of poured concrete surrounds the outside of the tank, except for a window 4 f t wide by I ft 10 in, high, which was lef t in the c:ncrete to allow for the insertion of a thermal column at a later date, and a 3-in.-diameter circular opening, which was intended to accept a van de Graaff generator beam tube.
The tank rests on a 1-ft-thick concrete slab.
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 approximately 4 in.
thick.
The reactor core cansists of stainless-steel-clad uranium-zirconium hydride (U-ZrH ) fuel elements that are assembled in concentric rings and supported by x
a 0.750-in.-thick aluminum grid plate.
The reactor is licensed to operate at thermal power levels up to and including 110 kW(t), using uranium fuel enriched to less than 20 percent in the uranium-235 (U-235) isotope.
1.3 Reactor Location The reactor is located on the first floor of the north wing of the Engineering Building in the nuclear reactor laboratory on the campus of the University of Arizoa, city of Tucson, Pima County, Arizona.
1.4 Shared Facilities and Equipment j
The reactor facility shares its utilities--electricity, water, natural gas, sanitary sewage, and the like--with the remainder of the Engineering Building.
The reactor room has a dedicated air exhaust system, but its heating and air-
)
conditioning system is integral with the rest of the building, j
1.5 Reactor History l
The reactor went into operation in December 1958 at 10 kW(t).
The original core loading consisted of 61 aluminum-clad fuel elements.
Subsequently, the number of elements was increased, as was the power (100 kW), and the fuel was replaced by stainless-steel-clad fuel in February 1973; 87 partially used stainless-steel-clad fuel elements were obtained through an Atomic Energy Commission grant.
The stainless-steel-clad fuel elements permit operation in the pulsed mode; the reac-tor is licensed to operate in this mode with a maximum reactivity insertion of 2.50$.
In 1971 the control console, bridge, control rods, and control rod driver and all the aluminum-clad fuel elements were transferred to the University of Utah.
These components were replaced at the UATR with components having features of more recent TRIGA reactors.
Only the reflector, the reactor pool itself, and l
the refrigeration and water purification systems are original equipment.
l 1.6 Comparison With Similar Facilities The reactor fuel elements are similar to those in most of the 55 TRIGA-type reactors in operation throughout the world, 24 of which are in the United States.
Of the 24 U.S. reactors, 21 are licensed by the NRC.
The instruments and controls are typical of the original TRIGA reactors and similar in principle to both the more recent TRIGA reactors and most of the other non power reactors licensed by the NRC.
NUREG-1390 1-3 A
1.7 Nuclear Waste Policy Act of 1982 Section 302(b)(1)(B) of the Nuclear Waste Policy Act of 1982 provides that the NRC may require, as a precondition to issuing or renewing an operating license for a research or test reactor, that the applicant shall have entered into an agreement with the U.S. Department of Energy (DOE) for the disposal of high-level radioactive wastes and spent nuclear fuel.
DOE (R. L. Morgan) has informed the NRC (H. Denton) by letter dated May 3,1983, that it has determined that univer-sities and other government agencies operating non power reactors have entered into contracts with DOE that stipulate that DOE retain title to the fuel and that it is obligated to take the spent fuel and/or high-level waste for storage or reprocessing.
Because the University of Arizona has entered into such a con-tract with DOE, the applicable requirements of the Waste Policy Act of 1982 have been satisfied for the UATR.
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i NUREG-1390 14 n
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2 SITE CHARACTERISTICS 2.1 Reactor Site i
The UATR is located in the Engineering Building on the campus of the University of Arizona (UA).
The 325-acre campus is located in the city of Tucson, Pima i
County, Arizona, in an area zoned for residential use and small business.
The city of Tucson is in southeast Arizona approximately 65 miles from the Mexican border.
It is 2410 ft above sea level and is situated in a high desert valley surrounded by the Santa Catalina Mountains to the north, the Rincon Moun-tains to the east, the Santa Rita Mountains to the south, and the Tucson Moun-tains to the west.
The valley gradually rises toward its center, and the university is located near this hil h point.
The valley is drained by three rivers or washes:
the Santa Cruz.iver Rillito River, and Pantana Wash.
The location of the reactor in the Engineering Building within the UA campus is j
shown in Figure 2.1.
2.2 Demography
- The population of Tucson's metropolitan area in 1988 was 648 492 and is pro-jectedtogrowto 943,000 by the year 2000.
ThenearestresIdenceisYumaHall, a student dormitory, located 300 ft west of the nuclear reactor laboratory.
The nearest private residences that are not under UA control are approximately 1300 ft west of the reactor laboratory.
The nearest private residences in the direction of the prevailing winds (southwest to northwest) are also approximately 1300 ft from the laboratory.
2.3 Nearby Industrial, Transportation, and Military Facilities There is no heavy industry in the vicinity of the UA campus.
The nearest airport, Tucson International Airport, is 6.5 mi south of the campus.
Inter-state 10, a major highway, passes within 1.75 mi from the west of the reactor laboratory. The Southern Pacific Railway and Amtrak pass within 1.1 mi south of the reactor laboratory.
The Davis Monthan Air Force Base is the only major military facility in the vicinity of Tucson and is located 5 mi southeast of the campus.
In view of the safe operating history of the past 30 years and the location of nearby industrial, transportation, and military facilities, the staff concludes l
that these facilities pose no significant risk to the safe operation of the UATR.
2.4 Meteorology The climate of Tucson is classified as a west coast desert climate.
Tucson's coldest month is January when the average low is 36.4 F and the hottest month 6
is July when temperatures reach ah average high of 99.1 F.
Tuscon is approxi-mately 325 mi from the Pacific coast, and therefore hurricane and tropical depression energy directed toward the city is depleted by the time the disturb-ance reaches the city.
Typical wind direction is from the southeast in the morning, shifting to a general west-northwest direction in the afternoon.
NUREG-1390 2-1 c
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Figure 2.1 Map of University of Arizona, Tucson i
NUREG-1390 2-2 c.
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Wind speed is usually between 5 to 10 mi/hr.
The average annual rainfall in Tucson is 11.2 in.
Tornados in Tucson are rare; sightings occur once in 2 years and one may touch down once every 10 years.
On the basis of the meteorological data presented in the licensee's Safety i
Analysis Report, the staff concludes that the meteorological conditions at the reactor site do not pose a significant risk of damage to the reactor nor render i
the site otherwise unacceptable for the facility.
2.5 Hydrology Tucson is located near the center of a valley almost completely surrounded by mountains.
The valley gradually rises toward its center, and the university is located near this high point.
The licensee provided data, in the submittal dated July 17, 1989, that show that the base of the reactor is above the 100-year flood level of near or distant stream beds.
The physical contours of the site and the drainage path near the building will not permit accumulation of water around the building.
There are no credible paths for reactor pool water to get into the campus water system or sanitary sewer system.
The water table at the university is typically 200 ft below the surface.
2.6 Geology and Seismology Tucson is in the Basin and Range Province.
The physiography of the area is typical of the arovince with alternating broad valleys and mountains.
Tucson is located in tie center of the Santa Cruz River Valley.
This valley appears to be a grabben or down-dro) ped block between valley bounding normal faults.
The valley is fiMed with a)out 2000 ft of upper Tertiary and Quaternary sedi-mentary deposits composed of alternating sands, silts, clays and gravels, which are underlaid by Tertiary volcanic rocks.
ThesurfacedeposItsaregravelsof 1
composition simila" to the bedrocks of the nearby mountains.
Tucson is in an area of relatively low seismicity.
Nomajorearthquakeshave been reported in the area.
A search of the U.S. Geological Survey's earthquake data file indicates that the nearest reported earthquake was a magnitude 4.5, which occurred on March 9, 1972, and had an epicenter more than 44 mi from.
Tucson.
Tucson is reported to have experienced Modified Mercalli intensities of up to IV from distant earthquakes.
S. M. Dubois and others (NUREG/CR-2577) place the site in their Seismic Zone 3, which they define as a zone of sparse seismicity.
2.7 Conclusion The staff has evaluated the UATR site for man-made as well as natural hazards and concludes that there are no significant hazards associated with this site that would render it unfit for continued operations.
NUREG-1390 2-3 A
n n
I
-1 3 DESIGN OF STR"CTURES, SYSTEMS, AND COMPONENTS
=
m 3.1 Reactor Facility Description which is located on the first Y
The UATR is in the nuclear reactor laboratoryllding. The Engineering Building floor of the north wing of the Engineering Bu is of brick and reinforced-concrete construction, including most floors and ceil-E ings.
The reactor is located near the bottom of a circular pit approximately 14 ft below ground level. The pit contains a steel tank resting on a 1-ft-thick E
concrete slab.
Approximately 8 in. of poured concrete surrounds the outside
=
of the tank, except for a window 4 ft wide by 1 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-in.-diameter circular opening, which was intended to accept a van de Graaff generator beam tube.
The inside of the steel tank is covered on the g-B" sides by a layer of Gunite approximately 2 in, thick and on the bottom by a layer 4 in. thick.
The entire inner surface of the Gunite is coated with Amercoat (an epoxy-base paint).
Figure 3.1 shows the reactor in the circular g
pit.
3.2 Wind and Water Damage p
E The Tucson area experiences very few extreme wind conditions such as tornados E
or inland hurricanes.
Furthermore, as described above, the reactor room is con-structed of reinforced concrete and brick.
The flood levels of near and distant stream beds are well below the reactor elevation, and the physical contours of gy the site and drainage path near the building will not permit accun,ulation of a
water around the building.
Therefore, wind or water damage to the UA facility is very unlikely.
E 3.3 Seismically Induced Reactor Damage h
Seismiology of the re$cally inactive area. Tremors that have been measured have ion is discussed in Section 2.6 of this repo~rt.
The UATR is located in a seism g
been no greater than the feeble-shock (intensity IV) range (felt by many, but no E
noticeable damage).
Because the reactor is below ground level inside a tank surrounded by steel-ancased concrete and because the floor and ceiling of the reactor and control rooms are constructed of reinforced concrete, the staff E
concludes that damage to the reactor's safety-related cumponents and systems
_g from any seismic event is unlikely.
[
3.4 Mechanit. Q Sysiems and Components The mechanical systems important to safety are the neutron-absorbing control rods suspended from the superstructure. The motors, gear boxes, switches, and wiring are all above the level of the tank water and readily accessible for visual inspections, testing, and maintenance. The staff has addressed the effects of aging on the continued performance of these components in Section 14 of this SER.
NUREG-1390 3-1 n
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3.5 Conclusion-On the basis of the above considerations, the staff concludes that the UATR was designed and built to adequately withstand all credible and likely wind, water, and seismic damage associated witn the site.
The design and performance of the safety systems have been verified by 30 years of operation.
Accordingly, the staff concludes that the reactor systems and components are adequate to provide--
reasonable assurance that continued operation will not cause significant radio-logical risk to the health and safety of the public.
.i.
I 4
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I NUREG-1390 3-3 y1
4 REACTOR The University of Arizona TRIGA reactor (UATR) is a General Atomic Mark 1 TRIGA reactor, currently licensed to operate at a maximum steady-state power level of 100 kW, or in the pulse mode with a maximum reactivity insertion of 2.50$.
The maximum allowed excess reactivity of the core is 3.25$.
In the application for renewal of the_ license, all operating parameters remain as originally licensed, except the maximum reactor power level will be increased to 110 kW.
The reason for the requested increase from 100 kW to 110 kW is to acrate4Wete the checking of the 110-percent power scram set points without exctec
, the licensed power requirements.
Nominal full power operation will continv< st 100 kW. This SER will include the requested 110-kW maximum licensed power as part of the evaluations.
The UATR uses solid uranium-zirconium hydride fuel containing 8.5 weight percent (wt%) uranium enriched to less than 20 percent U-235.
All of the UATR fuel ele-ments are clad with stainless steel.
The reactor core is immersed in an open tank of light water that serves as the neutron moderator, coolant, and partial shield.
The reactor power is regulated by inserting or withdrawing neutron-absorbing control rods.
Pulse-mode operation of the reactor is initiated by pneumatic ejection of a transient rod.
The UATR achieved initial criticality in December 1958 in accordance with License R-52.
The UATR is used principally for laboratory course instruction,-
thesis research, and university irradiation services.
Currently, the UATR is operated at an average of about 0.4 megawatt-day per year. The principal design parameters for the UATR are listed in Table 4.1.
The reactor facility is located at the University of Arizona, Tucson, Arizona.
The reactor is housed in Room 124 of the nuclear reactor laboratory, which is located on the first floor of the north wing of the Engineering Building (Build-ing 20).
In addition to Room 124 Room 122 (the reactor cor. trol room) and Room 124A (an equipment storage room) are security areas (restricted areas in accordance with 10 CFR Part 20) and are defined in the Emergency Plan as the emergency planning zone (EPZ).
Figure 4.1 is a plan of the north wing showing the EPZ.
4.1 Reactor Core The core is a right-circular cylinder, normally consisting of a lattice of cylin-drical fuel-moderator elements, control rods, and sample irradiation facilities, all of which are immersed in a pool of water.
Any unfilled core positions may contain water or aluminum-clad graphite " dummy" elements or additional experi-mental facilities.
Figure 4.2 shows a schematic view of the reactor core.
The reactor core assembly forms a 43.3-in.-diameter by 22.8-in.-deep right cylinder and contains 85 fuel elements (~3.0 kg U-235).
The fuel elements, two control rods, and one transient rod are positioned by the upper and lower aluminum grid plates. Water occupies about one-third of the core volume.
Calculations show that the resultant fuel-to-moderator ratio provides, very nearly, the minimum critical mass.
l NUREG-1390 4-1
I Table 4.1 Principal design parameters 1
Parameter Description Reactor type TRIGA Mark I Maximum licensed power level 110 kW Maximum pulse 2.50$ (0.017ak/k)
{
Fuel element design l
Fuel-moderator material U-ZrH1.68 Uranium content 8.5 wt%
Uranium enrichment
<20% U-235 Shape Cylindrical Length of fuel 15 in. overall Diameter of fuel 1.43-in. outer diameter Cladding material 304 stainless steel Cladding thickness 0.020 in.
Number of fuel elements 85 Weight U-235/ fuel element 35 g Excess reactivity, maximum 3.25$ (0.023ak/k)
Number of control rods 3
Transient (air-followed) 2.50$ (0.017Ak/k)
Shim (fuel-followed) 3.10$ (0.022ak/k)
Regulating (fuel-followed) 3.94$ (0.028ak/k)
Total reactivity worth of rods 9.54$ (0.067ak/k)
Reactor cooling Natural convection of pool water p effective 0.007 4.1.1 Fuel Elements i
The fuel ioading of the VATR core consists of 85 standard TRIGA fuel elements i
and 2 control rod fuel followers.
The fissile volume of each standard TRIGA fuel element 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 slightly less than 20 percent U-235.
The nominal weight of the l
U-235 in each UATR fuel element is 35 g.
The hydrogen-to-zirconium atomic ratio I
is approximately 1.68:1.
There is a 0.25-in.-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 fuel contains samarium oxide, a burnable poison.
Each i
fuel element is clad with 0.020-in.-thick 304 stainless steel tubing.
Sections of graphite 3.4 in. long are located above and below the fuel to serve as top
{
and bottom reflectors for the core.
Stainless steel end fixtures are provided i
l NUREG-1390 4-2 l
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- ROBOTICS LABORATORY
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ORATORY SL = DOORS WIT!! SECURITY IDCKS ROOM 118 SC E 1* = 10'-0" Figure 4.1 University of Arizona nuclear reactor laboratory
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' NUREG-1390 4-3
1 PNEUMATIC TRANSTER SYSTEM J l 1
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Figure 4.2 Schematic view of University of Arizona TRIGA reactor NUREG-1390 4-4 i
on both ends of the fuel element.
The overall fuel-element length is 28.37 in.
A *chematic view of a typical TRIGA stainless-steel-clad fuel element is shown in r3gure 4.3.
The fuel-follower sections of the two control rods are the same as the standard TRIGA fuel element, except the U-235 loading is slightly less (32 g) and the L
upper section of the graphite reflector has been removed so that the fueled section can be fit immediately below the control rod poison section.
4.1.2 Core Support Structure and Reflector The fuel elements are supported and spaced by means of top and botton grid plates of 6061 aluminum. The bottom grid plate is 0.75 in, thick, with one set of holes to allow convective flow of coolant upward through the reactor core and another set of holes to receive the end fixtures of the fuel elements in order
-to position the fuel in the core lattice. The top grid plate is also 0.75 in, thick, with holes that have an inner diameter of 1.5 in, for the fuel elements and the control rods.
The top grid plate does not support the weight of the fuel elements.
The holes serve only to determine the lateral position of the fuel elements and to permit withdrawal 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 spaces machined in the top end fixture of each fuel element.
Cooling water may also flow through holes in the top grid plate between the fuel elements and in a narrow gap between the top grid plate and the reflector.
The reactor core is located inside a cylindrical graphite reflector that is 12 in, thick and 22 in, high and has an inner diameter of 18 in.
The graphite reflector is completely clad in aluminum to prevent water intrusion.
A well to accommodate the experiment space known as the rotary specimen rack is provided-in the reflector so that the rotary specimen rack and the reflector are each individual watertight assemblies.
The reflector material above and below the core is composed of 3.4-in.-long pieces of graphite in the top and bottom ends of the fuel elements.
The reflector assembly rests on the reflector platform and provides the support for the two grid plates.
4.1.3 Control Rods Three control rods are used to control reactivity and regulate the power level in the UATR:
a transient rod in position C-10, with travel limited by a mechanical stop to provide a reactivity addition on pulsing that is less than 2.50$; a shim rod in position D-10, with reactivity worth of 3.10$; and a regula-ting rod in position C-4, with 'eactivity worth of 3.94$.
Additionally, pro-visions have been incorporated into the grid structure to add a control rod in core position 0-1, should thia Secome necessary.
The active neutron absorber 1 i ine control rods is sintered boron carbide (B 0).
4 The poison section of the shin snd regulating rods is 15 in. long and has an outer diameter of 1.435 in. The poison section for the transient rod is 15 in.
long and has an outer diameter of 1.25 in.
The fuel-follower rods are clad in 304 stainless steel tubing, and the transient rod is clad in aluminum.
The transient rod has a vertical travel of 9.72 in, and runs in a water-flooded aluminum guide tube; the control rods have a vertical travel of 15 in. and do NUREG-1390 4-5
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STAINLE$s gTgEL TOP END FITTING l
m GRAPHITE 3.4 IN.
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80TTOM ENO FITTING Figure 4.3 University of Arizona TRIGA fuel element l
HUREG-1390 4-6
not run in guide tubes.
During normal' steady-state o)eration, the transient rod functions as a control rod.
It is normally fully witadrawn to its mechanical stop at 9.72 in, and criticality is adjusted by the two control rods.
During transient mode operations, the transient rod is fully Inserted criticality is adjustedbythetwocontrolrods,andthetransientrodmechanismisadjusted so that the proper amount of transient Ap (change in reactivity) is added when the transient rod is ejected to its 9.72-in, mechanical stop.
The maximum withdrawal rate for the control rods is 0.404 in./s, which corre-sponds to a maximum reactivity addition of 0.17$/s for the regulating rod in position C-4.
Ganged operation of the tvo control rods is not possible.
The
. maximum poison withdrawal rate for the t,ansient rod when it is used as a con-trol rod (when using the drive) is 0.38 in./s, which corresponds to a reactivity addition rate of 0.14$/s.
'4.1.4 Neutron Source The startu) source used in the UATR is a 4.7-Ci americium-beryllium neutron source dou)1y encapsulated with 0.045-in, type 304 stainless steel.
The source is attached to a threaded stud at the end of an aluminum rod approximately 12 in. long.
The top of this rod has a 5-in, aluminum disk topped with an end fitting. identical to that of a standard TRIGA fuel element.
The standard end fitting allows the use of the TRIGA fuel-handling tool for incore positioning of the source when necessary.
During the startup of the UATR, the-source is normally positioned outside the graphite reflector in the southwest quadrant of the core, which is across the core from the startup channel.
From this position (Figure 4.2), a sufficient neutron count rate is detected on the startup channel to satisfy the minimum count rate specified in the facility's Technical Specifications.
4.2 Reactor' Tank and Biological Shield The reactor is located near the bottom of a steel tank resting on a 1-ft-thick concrete slab that is, in turn, located in a cylindrical pit in the reactor room floor.
The inside diameter of the steel tank is 7 ft, and the wall thickness is 0.25 in.
A;mroximately 8 in. of poured concrete surrounds the outside of the tank, except for a window 4 ft wide by 1 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-in.-diameter circular opening, which was intended to accept a van de Graaff generator beam tube.
The steel tank served as the inner form for pouring the concrete, and the outer form was a corrugated steel cylinder, which was left in place after the concrete was poured.
The inside of the steel tank is covered on the sides by a layer of Gunite approx-imately 2 in, thick and on the bottom by a layer of Gunite approximately 4 in, thick.
A depression 1 ft by 1 ft by 2 in deep has been provided in the Gunite at the bottom of the pool to facilitate complete draining of the tank by a port-able pump and hose, should this become necessary.
The entire inner surface of the Gunite is coated with Amercoat (an epoxy-base paint).
The pool is 20 ft deep and contains approximately 5000 gal of purified water.
Shielding above the top of the reactor core is provided by a minimum of 14 ft NUREG-1390 4-7 1
of water.
Shielding between the reactor core and the van de Graaff vault in the basement consists of approximately 1 ft of graphite, 1.5 ft of water, 2 ft of concrete, and 12.5 ft of soil.
4.3 Reactor Instrumentation The reactor instrumentation includes five neutron detectors in four independent neutron detection channels; water radioactivity, temperature, and conductivity monitors; an area radiation monitor; and a control console.
The neutron channels use one fission chamber and four ion chambers, mounted on the perimeter of the reflector, which provide steady-state power indication from 10-3 kW to 110 kW and pulse power up to 106 kW.
The reactor instrumentation is described in detail in Section 7.
4.4 Dynamic Design Evaluation The UATR is operated by manipulating control rods in response to changes in parameters such as temperature and neutron flux (power) as measured by the instrument channels.
Interlocks prevent excessive reactivity additions, and a scram system initiates a rapid shutdown (reactor scram) if a preset power limit has been reached.
In addition, the unique characteristics of the U-ZrH fuel-x moderator material provide a large, prompt, negative temperature coefficient that reduces the reactivity in the event of a significant increase in fuel tem-perature.
This provides additional operating stability and safety during any power transient.
The negative temperature coefficient results principally from the neutron-spectrum-hardening properties of ZrH at elevated temperatures, x
which increase the leakage of neutrons from the fuel-bearing material into the water-moderator material, where they are absorbed preferentially.
This reac-tivity decrease is a prompt effect because the fuel and ZrH are mixed homoge-x neously; thus, the ZrH temperature rises essentially simultaneously with fuel x
temperature (reactor power).
An additional contribution to the prompt, negative temperature coef ficient is the Doppler broadening of U-238 resonances at high temperatures, which increases nonproductive neutron capture in these resonances (Simnad, 1980; Simnad et al., 1976)
The inherent reactivity feedback property of the U-ZrH fuel has been the basis for the successful TRIGA reactor design with pulse capability as a normal mode of operation for many years.
The large, prompt, negative temperature coefficient rapidly and automatically compensates for step insertions of excess reactivity.
In the pulse mode, it terminates the resulting excursion without depending on electronic or mechanical safety systems or operator action.
In the steady-state mode, it serves as a backup safety feature, mitigating the effects of accidental reactivity insertions (Simnad, 1980; Simnad et al., 1976).
4.4.1 Excess Reactivity and Shutdown Margin The Technical Specifications for the UATR limit the maximum excess reactivity to 3.25$ in the cold xenon-free condition.
The Technical Specifications also require a minimum shutdown margin of 0.50$ with the highest worth control rod fully withdrawn, all non-secured experiments in their most positive reactivity state, and the reactor in the cold xenon-free condition.
NUREG-1390 4-8 i
The rea::tivity worth of a movable experiment in the reactor is_ limited by the Technical Jpecifications to less than 1.00$, that of any single experiment to less than 1.00$, and the maximum value of_all experiments to less than 5.00$.
The control rod worths for a typical core configuration are 2.50$ for the tran-sient rod,.l.10$ for the shim rod, and 3.94$ for the regulating rod, for a total rod e th o1 9.54$.
Assuming the UATR is configured with its maximum allowed exc m reactivity of 3.25$, the shutdown margin is -2.35$ (= 3.25 - 2.50 - 3.10),
which sdequatt.ly satisfies the Technical Specification shutdown margin require-ments.
With ail rods fully inserted (the normal shutdown conditirn), the reactor is subcritical at 6.29$.
4.4.2 Normal Operating Conditions The UATR Technical Specifications impose a peak steady-state fuel element tem-perature safety limit of 1000 C.
To protect this safety limit, the UATR Tech-nical Specifications impose a limiting safety system setting (LSSS) of 110 kW for'the power-level scram in the steady-state mode, which corresponds to a peak fuel temperature of 120 C.
The safety limit for the UATR high-hydride (ZrH1.68) stainless-steel-clad fuel elements is based on preventing excessive stress buildup in the cladding because of hydrogen pressure resulting from disassociation of the ZrH.
Based on theo-x retical and experimental evidence (Simnad, 1980; Simnad et al., 1976), the 1000 C fuel temperature limit represents a conservative value to provide confidence that the fuel elements will maintain their integrity and cladding will not be damaged.
Limitations are imposed on reactor power level and pulse reactivity insertion to ensure that the safety limit will not be exceeded.
At the maximum licensed nonpulsing power level of 110 kW, the maximum fuel temperature is 120 C.
During the maximum allowed 2.50$ pulse, local fuel temperatures will be less than 350 C.
Scrams are provided to-shut down the reactor whenever the pulsing power level reaches 110 percent of the expected peak pulse power with a maximum LSSS of 1100 MW, which corresponds to a peak fuel temperature in the core of less than 400*C.
Both the steady-state LSSS scram level of 110 kW and the pulse mode LSSS scram level of 1100 MW ensure the 1000 C fuel temperature safety limit is not exceeded anywhere in the core.
4.4.3 Assessment The staf f concludes that the inherent large, prompt, negative temperature coef-ficient of reactivity of the U-ZrH fuel moderator provides a basis for safe x
operation of the VATR in the nonpulsing mode and is the essential characteristic supporting the capability of operating the reactor in a pulse mode.
Furthermore, the Technical Specifications require that the core excess reactivity and experiment reactivity worths be limited so the reactor always can be brought to a subcritical condition, even if the highest worth control rod was totally removed from the reactor.
The current core configuration meets all of these limitations.
-The safety limits at the UATR are based on theoretical and experimental investigations and are consistent with those used at other similar reactors.
Adherence to these limits provides confidence that the integrity of the fuel l
NUREG-1390 4-9
I elements will be maintained. -Operating data at maximum licensed nonpulsing L
power and at maximum pulse reactivity insertion show that the maximum fuel ele-j ment-temperatures remain well below the prescribed safety limit.
TRIGA reactors H
similar to the UATR have demonstrated safe and reliable operation at nonpulsing 1
power levels up to 1.5 MW and pulse reactivity insertions up to 5.00$ (Simnad, 1980; Simnad et al., 1976).
On the basis of the above considerations, the staff concludes that, under normal operating. conditions, there is reasonable assurance that the UATR can be operated safely at a nominal power level of 100 kW with a maximum licensed power level of 110 kW and a maximum pulse limit of 2.50$, as 1
prescribed by the Technical Specifications.
4.5 Functional Design of Reactivity Control Systems j
i The power level in the UATR is controlled by two standard control rods (one shim and one regulating rod) and one transient rod.
The two standard control rods and the transient rod, which is operated as a control rod in the steady state-i power mode, contain sintered boron carbide as the neutron poison.
The-positions of the three control rods are shown in Figure 4.2.
Rods are moved using rack-I and pinion electromechanical drives for each standard control rod and a_ pneumatic electromechanical drive for the transient rod.
Each control rod drive system is energized from the control console through its own independent electrical cables and circuits, which tends to minimize the probability of multiple mal-4 functions of the drives. - On receipt of a scram signal, all three control rods-fall by gravity into the core, thereby shutting down the reactor.
4.5.1 Standard Rod Drive Assembly The control rod drive assemblies for the two standard control rods are mounted I
on a center channel bridge over the pool.
Each assembly consists of a reversible-single phase electric motor coupled to a rack-and pinion drive system. A draw tube connected to the rack supports an electromagnet that, in turn, engages an iron armature attached to the upper end of a long connecting rod.
The control rod proper is attached to the lower.end of the connecting rod.
During normal operation, the electromagnet is energized,-and the motorized system inserts or withdraws the. shim rod at a maximum rate of 19 in./ min, corresponding to an aver-i age reactivity insertion rate of about 0.065$/s, and the regulating rod at a, maximum rate of 24 in./ min, corresponding to an average reactivity insertion rate of 0.105$/s. The transient rod when used as a control rod during steady =
state conditions has an average reactivity insertion rate of 0.095$/s.
If power to the electromagnets is interrupted for any reason, the connecting rods are released and the two control rods fall by gravity into the core, rapidly shut-ting down (scramming) the reactor.
Three indicators, which are part of the rod control indicator-switch assemblies, illuminate the "UP" light when the rod is fully withdrawn, the "DOWN" light when the rod is fully inserted, and the " CONT" light when the armature is in contact with the magnets.
Continuous indication of rod position is displayed on the center vertical section of the control console.
4.5.2 Transient Rod Drive Assembly The transient rod drive is mounted on a frame that is bolted to the bridge and is operated by a pneumatic drive system consisting of a single-acting pneumatic NUREG-1390 4-10 i
I l
4
cylinder whose piston is attached to the transient rod by a connecting rod.
For pulse operation, compressed air is admitted to the botta of the cylinder through a solenoid valve, driving the piston upward in the cylinder and its connected transient rod out of the core.
At the end of its stroke, the piston strikes the anvil of a shock absorber and decelerates at a controlled rate.
Adjustments of the cylinder positions in relation to the piston head control the stroke length of the piston and hence the extent of the withdrawal of the transient rod from the core and the corresponding amount of reactivity inserted during a pulse.
The adjustment is performed electrically at the rod drive housing.
If power to the transient rod drive is deliberately or inadvertently interrupted, the solenoid valve is-deenergized, the air is vented from the cylindcr, and the transient rod drops by gravity into the core.
4.5.3 Scram-Logic Circuitry and Interlocks The UATR scram-logic circuitry and interlocks ensure that several reactor core and operational conditions are satisfied so that reactor operation may occur or continue.
The scram-logic safety system receives signals from core and other system instrumentation that initiate a scram by interrupting the electrical power to the control rod magnets and the transient rod solenoid air-operated valve, allowing the two control rods and the transient rod to drop into the core by gravity.
Scrams may be initiated by any of the following:
high reactor power in steady-state mode (110 percent) high peak power in transient mode (110 percent) pool water level less than 14 ft above the reactor core e
manual scram preset timer on the transient rod o
- earthquake sensor magnet power. supply failure external manual scrams power failure e
safety channel switched to " calibrate" or "zero" position e
Two operating system interlocks are provided.
The first is a rod withdrawal prohibit to prevent rod withdrawal until a minimum signal is obtained on the wide-range log power channel.
This interlock prevents the withdrawal-of a control rod if a minimum neutron source level count rate is not present. The second interlock is a pulse mode permissive that requires the reactor power to be less than a specified maximum value in order to activate the transient rod in the pulse mode.
Additional details concerning the safety-logic circuitry and interlocks are provided in Section 7, 4.5.4 Assessment The UATR is equipped with safety and control systems, control rods, rod drives, scram-logic circuitry, and interlocks that have performed reliably and satis-factorily in the UATR for many years.
NUREG-1390 4-11
1 The control systems allow for an orderly approah to criticality and for safe shutdown of the reactor during normal and abnorrra' conditions.
There is suffi-cient redundancy in the control rods to ensure safe reactor shutdown, even if the most reactive rod fails to insert on receiving a scram signal.
Interlocks prevent pulse mode operation at initial power levels that could result in exces-sive fuel temperatures.
A manual scram button allows the operator to initiate a scram independently for any condition requiring a prompt shutdown.
In addi-tion to the active electromechanical control and safety ~ systems, the large, prompt, negative temperature coefficient of reactivity inherent in the U-ZrHx fuel moderator.provides an ultimate backup safety feature.
Additionally, because i
the UATR fuel is slightly less than 20 percent enriched, almost 80 percent of I
the fuel is composed of U-238.
U-238 exhibits strong absorption resonances in the epitharmal neutron energy range (Doppler effect), which increases the proba-bility of neutron capture during slowing down, which, in turn, reduces the avail-able thermal neutrons that predominately cause fission. This inherent shutdown l
feature enhances the prompt, negative temperature coefficient of the U-ZrH TRIGA x
4 fuel On the basis of the above discussion, the staff concludes that.the reactivity control systems of the UATR are designed adequately ard will function to pro-vide reasonable assurance of safety for the reactor system as well as for the individual fuel elements.
Additionally, inherent shutdown characteristics ensure the reactor will remain safe even in the extremely unlikely event the engineered reactivity control systems fail.
4.6 Operational Procedures The University of Arizona has implemented administrative controls that require review, audit, and written procedu'es for all reactor safety-related activities.
The Reactor Committee reviews all aspects of current reactor operation to ensure that the reactor facility is opera *.ed and used within the terms of.the facility license consistent with the safety of the public as well as the operating per-sonnel. The responsibilities of this committee include review of-operating pro-cedures, experiments, and proposed changes to the facility or its Technical Specifications.
Written procedures reviewed by the Reactor Committee are established for safety-related activities, including reactor startup, operation, and shutdown; preven--
tive or corrective maintenance; and periodic inspection, testing, and calibration of reactor equipment and instrumentation.
The UATR is operated by trained NRC-licensed personnel in accordance with these procedures.
4.7 Conclusion The staff review of the VATR included a study of its design and installation and control and safety instrumentation.
These features are similar to those typical of the TRIGA-type research reactors operating in many countries of the world, more than 20 of which are licensed by the NRC.
There are currently about 10 1RIGA reactors operating at 1 MW or greater with no apparent safety-related problems.
On the basis of its review of the UATR and because of the operating experience of these other facilities, the staff concludes that there is reason-able assurance that the UATR is capable of safe operation as limited by its Technical Specifications.
NUREG-1390 4-12
.L
5 REACTOR COOLING AND ASSOCIATED SYSTEMS The reactor fuel is cooled by the natural circulation of about 5000 gal of deionized water.
Heat is removed from the water by the incore cooling coils of a Freon refrigeration system.
The quality of the reactor coolant is maintained by a purifiestion system.
A schematic of the water purification system is given
.in Figure 5.1, and the coolant system instrumentation is described in Section 7.
5.1 Reactor Coolina System s
The primary cooling system consists of a 7.5-ton vapor compression refrigeration system utilizing Freon-22 as the refrigerant that is capable of removing about 25 kW of heat from the' reactor pool.
The compressor and air-cooled condenser for this refrigeration system are located on a concrete slab outside the building.
The Freon is circulated through an aluminum evaporator (cooling coils) located in the pool water above the core.
The evaporator is made of two concentric coils of aluminum tubing in an array approximately 9 in. thick and 4 ft deep with an inner diameter of 5.5 ft.
This shape provides clearance for the removal of reactor core components if necessary.
The neutron and gamma fluens at the location of the evaporator are too low to cause either appreciable cctivation or radiation-induced decomposition of the Freon.
L The flow of the cooling water is downward along the outer radial region of the l
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 circulation.
5.2 Reactor Purification System The purification system. consists of a 15 gal / min pump, a Kuno-type fiber cartridge filter, a mixed-bed-type demineralizer, a flowmeter, and a surface skimmer connected by piping and valving.
A probe for measuring the conductivity of the water is located in the reactor pool.
The conductivity is. maintained below 5 pmho/cm.
Reactor pool water inadvertently lost thr) ugh the purification system cannot enter the university's sanitary sewer system because there are no inlets to the sanitary sewer system that are accessible frcm the reactor room.
L A limited amount of pool water, however, could enter the s',orm sewer system through leakage from the demineralizer system.
This leakage would be limited to a maximum of about 280 gal by the underwater siphon break holes in the pool inlet and outlet pipes.
Assuming the maximum amount of pool water that would credibly be released at the highest possible tritium level, the calculated max-imum radioactivity level in the storm sewer system is 0.015 pCi/L, which is a factor of 200 below that allowed by 10 CFR Part 20 for the discharge of water to an unrestricted area.
The purification system is shown schematically in Figure 5.1.
NUREG-1390 5-1
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5.3 Primary Coolant Makeup System Makeup water to replace pool water lost as a result of evaporation is supplied by containers of distilled water that is poured into the pool.
It is not pos-sible to introduce pool water into the campus water supply because there are no water lines in the reactor room.
Under emergency conditions, water from the university water supply could be added rapidly and directly to the pool using fire hoses.
5.4 Assessment Because the maximum licensed authorized power level of the UATR is 110 kW(t) and the capacity of the coolant system is 25 kW(t), the University of Arizona staff considers the Freon refrigeration system to be an operational aid rather than a device to allow unrestricted operations at the nominal power level of 100 kW.
It allows for prolonged operation at low power levels and a speedy recovery (cooling) of the system following operations at high power levels.
At 100 kW, the rise in the UATR pool water temperature when the coolant system is operating at its maximum capacity is about 4C* per hour.
Administrative procedures limit the maximum reactor pool water temperature to 45C, which results in a reactivity decrease from ambient pool conditions of about 0.10$.
5.5 Conclusion The staff concludes the following:
(1) The size, design, condition, and maintenance level of the UATR cooling system are adequate for ensuring cooling of the reactor under the routine operating conditions specified in the UATR operating license at power levels less than 25 kW.
At power levels equal to or greater than 25 kW, the UATR cooling system is adequate to aid in slowing the rise in pool water temperature, thus allowing somewhat longer high power runs, and to aid in accelerated cooling of the system following reactor shutdown.
(2) The size, design, condition, and maintenance level of the UATR purification system are adequate for ensuring that the quality of the coolant water remains sufficiently high to preclude any possible water corrosion damage to the core or core support structure under all operating conditions speci-fled in the operating license.
The purification system at UATR has the same design features as those used in many other operating TRIGA reactor facilities.
There is no new or unproven technology involved in this system.
(3) The UATR cooling system is designed so that an inadvertent release of reactor pool water to the campus water supply system or campus sanitary sewer system is not credible.
An inadvertent release of up to 280 gal of pool water to the storm sewer system is credible; however, the associated reactivity release is negligible.
On the basis of the above observations, the staff concludes that the cooling and purification systems at the UATR are adequate for safe operations.
NUREG-1390 5-3
6 ENGINEERED SAFETY FEATURES Engineered safety features are those features or systems that mitigate the potential consequences of accidents. The one engineered safety feature asso-ciated with the UATR facility is the stack vent exhaust system.
This system is designed to limit the uncontrolled release of airborne radioactive materials under normal reactor operating conditions as well as accident conditions.
6.1-Stack Vent Exhaust System Under normal operating conditions. inlet air to the reactor room is supplied by the building ventilation system; this air is exhausted to the outside through a window-mounted exhaust fan that is located about 12.5 ft above the outside ground level.
Should-an airborne contamination alarm be generated by the con-3 tinuous air monitor, the normal exhaust fan will stop and an emergency 580-ft /
min exhaust fan will start.
In this mode, air will be drawn through a high-efficiency particulate air (HEPA) filter and exhausted through a vent on the building roof that is located about 50 ft above the-level of the ground sur-rounding the Engineering Building.
Using this stack vent exhaust should reduce-airborne contamination by a factor of 10-5 Even though this engineered safety system is not necessary to mitigate the calculated consequences of the iodine-131 release from the maximum, hypothetical accident to levels below those allowed in 10 CFR Part 20 (see Section 14), its installation at the UATR demonstrates-the large safety margin present at this facility.
6.2 Conclusion The stack vent exhaust system at the UATR is designed to accommodate all air exhausted from the reactor room with the building ventilation system on (i.e.,
forced. air into the room) and the window fan off.
In this mode, all air leaving the reactor room will be forced through the HEPA filter and none will backflow into other rooms in the Engineering Building or leave the reactor room through the window fan opening.
To demonstrate this, the licensee conducted air current tests during a site visit in May 1989.
These tests demonstrated that when the 3
HEPA fan was run at high speed (580 ft / min), all air in the reactor room was exhausted up the stack.
The staff concludes that when the HEPA fan is run at high speed, the UATR stack vent system is adequately designed to perform its intended function.
l NUREG-1390 6-1
7 CONTROL AND INSTRUMENTATION SYSTEMS The reactor control and instrumentation systems are designed to provide safe, reliable operation of the UATR in four different modes:
manual, automatic, pulse low, and pulse high.
The control console, which is located in the control room, displays the reactor parameters, including power level, fuel element tem-perature, pool water temperature, and rod position.
A logic diagram of the control and instrumentation systems is shown in Figure 7.1.
7.1 Reactor Control System The reactor control system consists of those components that control the operation of the reactor control rods as well as associated equipment appro-priate to the reactor operating mode selected.
The reactor control rods and rod drive mechanism are described in Section 4.
7.1.1 Control Console The reactor control console and desk contain the control, iiidicating, and recording instrumentation required for operating the reactor.
These controls and instrumentation are placed to allcw the reactor operator convenient visual and manual access.
An area on the desk surface allows the logging of data during operation.
The left-hand vertical section of the console contains meters that display a percent power channel (left safety channel), the wide range log power channel, and the reactor period.
Visual annunciators on the panel below the meters indicate the status of the protection system components mounted in this section of the console.
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 that display continuous position for the two control rods and the transient rod.
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 right of the position indicators.
The lower center section mounted on the sloped desk portion of the console contains the manual rod control switches in the center. To the left of the rod control switches are the transient rod control switch and the key switch, and immediately below it, the power-on indicator switch.
To the right of the rod control switches is the multirange linear piccammeter range selector switch.
The right vertical section of the console contains meters that 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 components mounted in this section of the console.
NUREG-1390 7-1
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7.1.2 Operating Modes Four modes of operation are associated with the UATR:
manual, automatic, pulse low, and pulse high.
7.1.2.1 Manual Mode sThe reactor is always started in the manual mode; this mode can be used to control the steady-state reactor power level from the source level to the maxi-
' mum licensed power level.
The wide-range log 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.
7.1.2.2 Automatic Mode The automatic mode is identical to the manual mode except the regulating rod-position is controlled by a feedback control system to regulate the reactor power level as detected by the linear multirange channel.
Automatic mode oper-ation can be initiated at any power level detectable by the linear multirange channel and is usually selected at a reactor power level of about 1 W.
The set
. point 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 trensient rods may be controlled manually, and the regulating rod is controlled in the automatic mode.
7.1.2.3 Pulse Low and Pulse 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 before a pulse is initiated.
Before a pulse, the transient rod poison remains inserted in the core and its cylinder is raised to the position that will produce the desired reactivity change when air is applied to the transient' rod piston in the cylinder.
Reactor criticality is adjusted by use of the shim rod and regulating rod only.
Af ter the pulse, a timer initiates a scram sequence, during which all control rods are dropped 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 the pulse high mode, full scale
' corresponds to 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 1
instrumented fuel element is displayed by the " green pen."
After the transient rod fire button is operated, the power pulse is terminated by the prompt temperature coefficient.
All the control rods are then inserted by the timer to maintain the reactor in a shutdown condition.
No further operator control actions are required to shut down the reactor.
7.2 Instrumentation System The instrumentation system is composed of nuclear instrumentation and nonnuclear process instrumentation that provide the operator with the information necessary for the proper operation of the facility.
NUREG-1390 7-3
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7.2.1 Nuclear Instrumentation i
Four independent systems (startup, linear power, safety, and pulse) are used to l
process the input signals from five nuclear detectors.
Four of these detectors
--one fission chamber for startup, one compensated ion chamber for the linear power system, and two uncompensated ion chambers for the safety system-provide wide-range power indication (10 8 W to 110 kW), which covers the source range up to 100 percent of licensed power with appropriate overlaps among the channels.
The fif th detector is an uncompensated ion chambar that provides pulse power up to 106 kW.
The startup channel consists af a wide-range log power channel that measures power over the complete neutron flux range below source level'to above the full steady-state power level.
It receives its input signal from the fis-sion chamber.
The output signal is displayed both by the " green. pen" of the dual pen chart recorder and on the wide-range log power meter.
A pulse log-count-rate technique is used for the lower 5 decades of the range and a log Campbell technique for the upper 5 decades to produce a single output signal for over 10 decades.
Neither technique is strongly affected by the gamma-ray background of the TRIGA reactor.
Test and calibration circuits provide appro-priate signals into the input of the preamplifier for checking six calibration levels.
The log channel contains (1) a period (rate of change of power) circuit with front panel meter display and (2)-two bistable trip units, which are used l
for control system interlock functions.
One bistable trip acts as a rod with-drawal prohibit if the source level count rate as detected by the channel is below 2 counts /s.
The other bistable trip prohibits pulse operation if the reactor power level is above a specified minimum level.
The multirange linear power channel utilizes a multirange linear picoammeter, which is a de measuring instrument specifically designed for use with nuclear reactors.
The channel receives an input signal 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 that produces an output
' voltage proportional to the input current.
Full-scale output of 10 V may be l
obtained with input currents between 10 10 ampere and 10 3 ampere in 15 ranges.
Test and calibration signals are built into the channel and may be selected from the range selector switch.
The multirange linear picoammeter is located
.beneath the rod control panel in a shielded aluminum box.
The linear amplifier, range switch, ranging resistors, and output buffer amplifier are constructed as an integral unit and may be removed for maintenance-if required.
The output of the multirange linear channel is displayed by the " red pen" of the dual pen strip chart recorder.
This output is also input to the flux regulator as a feedback signal in the automatic mode of operation.
The right and left power level channels that'make up the safety system derive their signals from uncompensated ion chambers.
The output current of the cham-ber 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 bi-stable trip unit with the trip set below the licensed reactor power.
A mercury-wetted relay 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 is to the zero position where the input is short circuited.
A full-scale calibra-tion check is also provided to measure the sensitivity of the instrument to NUREG-1390 7-4
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current.
A. trip _ test control circuit can add 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.
7.2.1.1 Reactor Scram System A protective action at the UATR will interrupt the control rod magnet current and the power to the transient rod air-operated solenoid valve and result in the immediate release of all rods when activated by a safety channel._ The minimm required safety system channels and their set points are listed in Table 7.1.
Table 7.1 Minimum reactor safety system channels Safety channel Function Set point Manual scram Scram Manual Pool water level Scram 14-ft Linear power level Scram-110% of full scale Percent power level Scram 110% of full scale Peak pulse power Scram 110% of full scale-High voltage Scram Loss Magnet current Scram Loss Earthquake sensor Scram As desired Pulsing timer Scram As desired External safety switches Scram As required 7.2.2 Honnuclear Process Instrumentation The nonnuclear process instrumentation measures reactor coolant wate' tempera-ture and water conductivity.
The water temperature monitor utilizes a thermister located in the pool about 10 ft below the surface of the water.
The readout is via a_ meter on the right side of the console. Operating procedures preclude operation at a reactor pool temperature that is greater than the Technical Specification limit of 45 C.
The water conductivity monitor consists of one conductivity probe about 10 ft below the pool surface and a detector with a meter readout in the corner cabinet.
7.3 Conclusion The control and instrumentation systems at the UATR, which are similar to those in other operating TRIGA reactors, are well designed and provide for reliability and flexibility.
All power and instrumentation wiring is well identified and protected from physical damage, where possible, by conduits. The nuclear power monitoring circuits are redundant and diverse.
In particular, nuclear power measurements are overlapped in the ranges of the log-N, linear power, and percent power level channels.
The control system is designed to shut down the reactor automatically if electrical power is lost or interrupted, i
HUREG-1390 7-5
4
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-From the preceding analysis and the' formal administrative controls' required'for the operation of-:the UATR, the staff concludes that both the control and instru--
. mentation systems at the UATR comply with the requirements and performance objectives of the Technical Specifications and are acceptable to' adequately
- ensure the continued safe operation.of the reactor..
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S 8' ELECTRICAL POWER SYSTEM The electrical power system at the UATR is a standard electrical supply system designed and constructed to specifications similar to those at other research reactor facilities.
8.1 Normal Power Electrical power for the UATR facility is provided by the university's electri-cal power distribution system, which is supplied by the Tucson Electric Power Company. Transformers located outside the reactor laboratory building step down the' campus distribution from 4160 V to 120/220 V.
8.2 Emergency Power Emergency electrical. power is supplied by trickle-charged batteries that sup-ply power to the security-intrusion system and the emergency lighting system.=
Trickle-charged batteries also are used as backup to power the fire alarm sen-sors and the communication link from the fire alarms to the police station.
This ensures that each of these systems will continue to operate normally in.
the 9 vent of a power failure.
8.3. Conclusion The staff concludes that the electrical power system at the UATR is a standard electrical supply system typical of research reactor facilities and is adequate for the duration of-the proposed license period.
The staff also concludes that emergency power in addition to that currently available is unnecessary.
NUREG-1390 8-1
1 9 AUXILIARY SYSTEMS 9.1 Ventilation System-Fresh air for the three rooms in the Engineering Building that constitute the nuclear reactor laboratory (i.e., Room 124, the reactor room; Room 122, the con-trol room; and Room 124A, the equipment storage room) is supplied by the Engi-neering Building forced convection air handlers of the heating, ventilation, and air conditioning (HVAC) system.
Each of these rooms has a ceiling-mounted vent to allow fresh air in.
rovided with a window-mounted exhaust fan with a minimum The reactor room is p/ min.
This is the only room in the nuclear reactor labo-flow rate of IN0 ft3 ratory with forced exhaust to the outside.
During reactor operation, the win-dow. fan is in operation.
The licensee has demonstrated with artificial smoke that air ~ is pulled out of the reactor room when this f an is operating and is prevented from entering the Engineering Building through the HVAC ducts or through spaces around doors.
When an airborne contamination alarm is generated by the continuous air monitor, the normal exhaust fan stops and an emergency exhaust' fan starts automatically.
In this mode, air is drawn through a high-efficiency particulate air filter and exhausteu through a vent on the building roof that is located about 50 ft above the level of the ground surrounding the Engineering Building (see Section 6).
9.2 Fire Protection System The Engineering Building and its fire protection system conform to national and Arizona fire protection codes.
The fire main, pressurized by the university's water system and backed up by the city of Tucson water system, supplies the sprinkler systems and hose stations in the building exclusive of the nuclear reactor laboratory.
Portable fire extinguishers in the Engineering Building hallways are of the pressurized-water and carbon dioxide types.
Two carbon dioxide portable fire extinguishers are mounted in the reactor laboratory.
Fire protection system monitoring panels are mounted on the north wall inside the main entrance and outside on the south side of the building.
These monitoring panels.are powered by a noninterruptible power supply and are linked to the University of Arizona police dispatcher console, which is staffed at all times.
As required, the University of Arizona police dispatcher notifies the Tucson Fire Department,-which is staffed at all times with professional fire fighters.
During test exercises, these fire fighters respond to the site within a few minutes of an alarm.
University police personnel are assigned to work with the
- Tucson Fire Department on fires involving potential radiation or in areas with radioactive materials.
The requirements for radiation safety under such cir-cumstances are coordinated by the University Emergency Director.
9.3 Communication System A commercial telephone is available to the control room operator for communica-r tion between the reactor laboratory and the outside.
In addition, adjacent laboratories have commercial phones, and the intrusion alarm or the fire alarm will summon police and emergency' personnel.
NUREG-1390 9-1 1
n 9.4 : Air Conditioning System
-The Engineering Building, including the reactor laboratory, is air conditioned by forced convection air handlers.
The chilled water and heating steam, which 1
are circulated through the air handlers, are provided by a refrigeration and steam plant.
- 9. 5 Fuel Handling and Storage A 30 position fuel-element storage rack on the floor of the reactor pool, which is equipped with epoxy-covered cadmium strips between the fuel locations, and 1
two 13 position holsters on the side of the pool provide storage for fuel during
. fuel inspections and approach to critical experiments.
The K,ff assuming all available locations are loaded with a TRIGA fuel element is estimated to be less than 0.90.
9.6 Conclusion The staff concludes t at the auxiliary systems at the VATR facility are designed '
and maintained appropriately and are adequate for their intended purposes.
The UATR reactor area ventilation system and equipment are adequate to control the release of airborne radioactive effluents during normal operations in compliance with regulations and to limit credible potential releases of airborne radioac-4 tivity in the event of abnormal conditions.
Fire protection at the reactor facility is adequate to provide protection against the types of fire hazards associated with the operation of a.research' laboratory, and the fuel handling and storage capability is consistent with the UATR requirements and is designed in accordance with accepted criticality safety standards and good engineering practice.
All of these auxiliary systems have performed adequately and reliably over the 30 years of operation of the UATR.
l NUREG-1390 9-2
1 :
I i
10 EXPERIMENTAL PROGRAMS I
The UATR is used as a teaching facility by the Department of Nuclear and Energy Engineering.
It also is used to provide irradiation services, primarily acti-vation analysis, for the faculty and students in many other departments of the University of Arizona.
Less than 1 percent of the operation consists of serv-ices provided to industrial firms. Thus, the reactor is primarily operated for laboratory course instruction, thesis research, and university irradiation serv-ices.
Specimens can be irradiated in the pneumatic transfer tube, rotary spec-imen rack central thimble, demountable fuel elements, or in the fast irradia-tionfacility,neutronradiographytube,andgraphitethermalizerblock.
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10.1 Experimental Facilities 10.1.1 Pneumatic Sample Transfer System The pneumatic sample transfer system (rabbit system) is used for neutron irradiation of single samples.
It consists of two tubes connecting a sender-receiver station in the F-ring of the reactor core with a terminal station on
'the reactor bridge. 1he sample is packaged in an irradiation capsule that moves in one tube the other tube is used to complete the air passage circuit.
A blowerprovIdesthepressuredifferenceformovingthecapsule.
l 10.1.2 Rotary Specimen Rack The rotury specimen rack (lazy Susan) is used for neutron irradiation of many samples simultaneously.
In this facility, 40 evenly spaced aluminum cups, which serve as holders for irradiation capsules, are located in the radial graphite l
reflector in an aluminum ring that can be rotated around the core.
The ring can be either driven by an electric motor or rotated manually from the top of the reactor pit.
Any cup can be aligned with the isotope-removal tube (lazy Susan access tube) for inserting and removing specimens.
An indexing device is pro-vided to enable positioning of the cups.
The-rotary specimen rack is completely enclosed in a welded aluminum box.
The aluminum ring is located at approximately the level of the top grid plate.
The l
specimen cups extend from the ring down to about 4 in, below the top of the ac-tive lattice.
In the radial direction, the centers of the cups are located in the graphite reflector about 13 in, from the center of the core.
Because the box enclosing the rotary specimen rack constitutes a considerable void in the gra)hite reflector, it has been designed to ensure that it will remain water-tigit.
Flooding of this void would increase the reactivity of the core about 0.00$.
10.1.3 Fast Irradiation FacWty, Neutron Radiography Tube, and Graphite Thermalizer Block In addition to the thermal neutron irradiation facilities described above, irradiation facilities are available that were designed and constructed at the University of Arizona and are particular to the UATR.
NUREG-1390 10-1
The fast neutron irradiation facility (fir) is a normally voided test space that is inserted in a standard fuel-element position.
It is a tube whose ter-i minus is lined with boron, cadmium, and gold to absorb thermal neutrons while permitting fast neutrons to pass into the sample.
Samples are lowered into the tube in an aluminum can on a nylon cord. The Flr pertrits studies of f ast neu-tron damage to materials without introducing the radioactivity caused by absorp-tion of thermal neutrons.
Similar tubes that are not lined with thermal-neutron-absorbing materials are provided to permit the incore irradiation of single samples and to contain neutron detectors for experiments measuring reactor parameters.
The neutron radiography tube is a 12-in.-diameter beam tube that permits the streaming of a near parallel, low-flux beam of thermal neutrons from the reflec-tor area of the reactor.
This beam is used to expose neutron radiographic pic-tures. When not in use, a scattering block and two shield plugs prohibit neu-tron or gamma radiation from reaching the surface of the reactor pool.
The graphite thermalizer block is encased in a watertight cover of 0.125-in.
6061 aluminum and is permanently positioned west of the reactor core.
The inner
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and outer curved surfaces of the block match the radii of the graphite core reflector and reactor pool, respectively, so that the block occupies the avail-able space between the core reflector and the pool wall.
The block is the same height as the core reflector and has three aluminum irradiation thimbles with inside diameters of 1.38 in, set into its top surface so that the bottoms of the thimbles are approximately 5 in, below the vertical centerline of the reac-tor core.
The centerlines of the thimbles are in a radial line due west of the reactor core approximately 4, 7, and 11 in, from the edge of the graphite core reflector.
To access 0 +.her of the thimbles, an access tube with 0-ring seals is slipped into the t w sle from above the pool surface.
The tube is then pumped dry, allowing access of the sample to the thermalizer block.
10.1.4 Central Thimble The central thimble is provided to permit irradiation of experiments at the center of the core in the region of maximum neutron flux.
It can also be used to provide a highly collimated beam of neutrons and gamma rays when it is emp-tied of water.
The thimble is an aluminum tube with an insid6 diameter of
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1.33 in.
It extends from the top of the tank through the two grid pir4tes and r
terminates in a plug at a point approximately 7.5 in. betuw the lower grid plate.
The tube is normally filled with water.
However, the water can be replaced with air or a gas such as helium, so it can serve as a beam tube.
Replacing the water with gas decreases the UATR reactivity about 0.08$.
10.1.5 Demountable Fuel Element The demountable fuel element is a special facility designed for flux mapping and the determination of reactor parameters within a TRIGA fuel element.
It consists of 1-in.-long,1.435-in.-diameter sections of ZrH containing approximately 12 g x
of uranium, enriched to 19.99 percent in U-235.
The demountable fuel element
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has graphite end reflectors and is clad with the same stainless steel material f
as standard TRIGA fuel.
The top of the demountable fuel element may be unscrewed in order to install or remove foils between the fuel sectiuns.
l NUREG-1390 10-2
4 10.2 Experiment Review Before any new experiment mev be conducted using the reactor or the associated experimental facilities, it is reviewed by the Reactor Committee.
The committee is composed of at least five members, including a health physicist and members competent in the field of reactor operations, radiation science, or reactor engineering.
The membership of the Reactor Committee is designed to provide a spectrum of expertise to review the experiments and their potential hazards.
The review and approval process for experiments also allows personnel experienced in reactor operations to consider and suggest restrictions and/or alternative operational conditions, such as using a different experimental facility or dif-ferent power level and irradiation times, that might lead to decreased personnel exposure and potential release of radioactive materials to the environment.
Restrictions generally include requirements that the experiment remain sealed, no materials be released to the pool if the experiment should be breached, and the experiment and its components be monitored for the presence of radiation and loose radioactive material as they are removed f rom the pool.
Specific restric-tions may also apply depending on the nature of the experiment.
The review process also includes considering the ef fect of the experiment on the natural convective flow of water through the core of the reactor.
Convective cooling flow at any power level must not be disturbed in a way that would lead to overheating of any part of the core.
The size and the position of the pro-posed experiments will determine the effect on the flow of water.
10.3 Experiment Reactivity Limitations The Technical Specifications include a 1.00$ limit of reactivity for each movable experiment and a maximum of 5.00$ total reactivity insertion for all combined experiments in the reactor at any one time.
However, the maximum allowed excess reactivity of the UATR under any experimental conditions shall be no greater than 3.25$.
In Section 14.2, the staff evaluates the consequences of inadvertent reactivity insertions equal to the above Technical Specification limitation on excess.
The staff concludes that the reactivity limitation in the Technical Specifications would ensure that increases in fuel and cladding temperatures will remain well below the critical limits in the unlikely event that all of this reactivity is added to the UATR in a stepwise fashion.
10.4 Conclusion The staff concludes that the design of the experimental facilities at the UATR combined with the detailed review and administrative procedures applied to all research activities is adequate to ensure that experiments are unlikely to (1) fail, (2) release significant radioactivity to the environment directly, and (3) cause damage to the reactor systems or its fuel.
Therefore, the staff believes that reasonable provisions have been made so that the experimental pro-grams and facilities do not pose a significant risk of damage to the reactor or of an uncontrolled release of radioactive materials that would pose a signifi-cant radiological risk to the facility staff or the public.
1 HUREG-1390 10-3 b
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11 RADI0 ACTIVE WASTE MANAGEMENT The major radioactive waste generated by operation of the UATR is activated gases, principally argon-41 and nitrogen-16.
Small volumes of liquid and solid radioactive waste, mostly spent ion-exchange resins, are also generated, pri-marily in connection with the experimental uses of the reactor.
11.1 As low As Is Reasonably Achievable (ALARA) 11.1.1 ALARA Policy Statement The following University of Arizona ALARA policy statement was signed by Michael A. Cusanovich, Ph.D., Vice President for Research, and was submitted to the NRC on July 17, 1989.
The principle of keeping radiation doses from all University of Arizona licensed operations as low as [is) reasonably achievable (ALARA) forms the basis for the university's radiation protection program.
The principle of ALARA will be applied wherever possible in order to minimize occupational exposures to radiation and releases of radioactive material.
11.1.2 ALARA Commitment All radioactive materials are handled and released and all exposure to ionizing radiation is controlled using the ALARA principle.
The objectives of the ALARA program are to minimize the exposure of individuals to ionizing radiation, to minimize the release of radioactive materials, and to minimize the release of radioactive materials to the uncontrolled environment.
Training, planning, practice sessions, shielding, distance, special tools, monitoring, and design of experiments are used to achieve the goals of the ALARA program.
11.2 Waste Generation and Handling Procedures 11.2.1 Solid Waste The disposal of spent fuel is not expected to occur during the term of this license renewal.
However, should the reactor be unexpectedly decommissioned during the license period, the University of Arizona has an agreement with the U.S. Department of Energy for the ultimate disposal of the spent fuel and/or any high-level waste.
The largest volume expected of solid radioactive waste is slightly contaminated paper and plastic material.
Most of the activity in the solid radioactive waste is found in activated samples, components, and equip-ment.
Dried spent ion-exchange resin (2.5 ft3 every 1 or 2 years) is also treated as solid radioactive waste (see Section 11.2.2).
The solid waste is collected by the University of Arizona health physics staff in specially marked packages.
The waste is held temporarily before it is packaged and shipped to an approved disposal site in accordance with applicable regulations.
NUREG-1390 11-1 i
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11.2.2 Liquid Waste The largest volume of liquid waste water is collected when the purification system filter and spent ion-exchange resins are changed.
This liquid is poured back into the pool.
Any other liquid waste, such as that which may be associ-ated with some solid waste, is dried and disposed of as long-lived, low-level radioactive solid material.
11.2.3 Airborne Waste Airborne discharges from the re'ictor facility include argon-41 (Ar-41) from the activation of air dissolved in the pool water and Ar-41 from the activation of air contained in the pneumatic sample transfer system (rabbit system).
Small amounts of nitrogen-16 (N-16) are also generated from the activation of oxygen in the pool water.
The UATR staff assessed the Ar-41 release resulting from normal and abnormal operating conditions both inside the reactor laboratory (restricted area) and outside the reactor laboratory (part of the campus-wide unrestricted area).
The results for the unrestricted area are presented in this section.
The results for the restricted area are presenteo in Section 12.6.1.
The staff has reviewed the UATR staf f's assumptions and calculations, which are presented below, and agrees with its findings.
There are only two sources of significance of Ar-41 at the UATR, the pool water and the pneumatic sample transfer system (rabbit system).
All the other devices that contain air are sealed and do not contribute to airborne Ar-41 in the reactor laboratory.
At 110 kW, the rabbit system generates a calculated 9.1 x 10-2 pCi/s and the pool releases a calculated 1.2 x 10-2 pCi/s, for a com-bined release of 10.3 x 10-2 pCi/s of Ar-41.
Assuming this is all pumped out-side through the winoow fan in the reactor room results in a concentration in the unrestricted area of about 3 x 10'R pCi/cm3, This is more than a factor of 100 below the concentration allowed in 10 CFR Part 20 for continuous occu-pancy, that is. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> par day for 365 days per year.
The total dose rate for continuous occupancy for a year is calculated to be less than 1 mR/yr.
Although radioactive N-16 is formed in the core of the UATR, the transport time from the core to the surface of the pool has been conservatively estimated to be more than 10 times the 7.4-s half-life of N-16.
The amount of radioactive N-16 released to the environment, therefore, is considered to be negligible.
11.3 Conciusion
.The staff concludes that the waste management activities at the UATR facility have been conducted and are expected to be conducted in a manner consistent with 10 CFR Part 20 and the ALARA principle.
The staff conducted its review using the methods of ANSI /ANS 15.11, " Radiological Control at Research Reactor Facilities," along with other guidance.
Because Ar-41 is the only significant radionuclide released by the reactor to the environment during normal operations, the staff reviewed the history, current practices, and future expectations of reactor operations with respect to this radionuclide.
Because of the extremely small amounts of Ar-41 released from the UATR over the past 30 years at the maximum authorized power level of NUREG-1390 11-2
100 kW, the staff concludes that the amount of Ar-41 released in unrestricted areas outside the UATR reactor laboratory will be significantly below the limits specified in 10 CFR Part 20 when averaged over a year.
NUREG-1390 11-3 n:
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l 12 RADIATION PROTECTION PROGRAM The University of Arizona has a structured radiation safety program and a health physics staff to implement the program.
The reactor facility has the equipment to detect, measure, and control area and personnel radiation expo-sures.
Use of radioactive material and radiation sources is controlled care-fully, and releases of radioactive material to the environment are kept to a minimum.
12.1 ALARA Commitment The University of Arizona radiation protection program is designed to be con-sistent with the policy of ALARA (as low as is reasonably achievable).
The University of Arizona, through the nuclear reactor laboratory (NRL) and the Radiation Control Office (RCO), has developed a training program that incorpo-rates procedures and equipment to implement this policy.
Personnel and environ-mental monitoring programs ensure that radiation exposures to the NRL personnel and the general student population and public are kept ALARA.
12.2 Health Physics Program 12.2.1 Health Physics Staffing The University of Arizona RCO provides health physics services to the NRL.
Professional and technical health physics staff members oversee the radiation protection program and serve as resource persons for all the users of radiation and radioactive materials on campus.
Reactor personnel perform the normal radi-ation safety function at the reactor facility.
One of the health physicists from the RCO oversees the radiation protection program at the reactor facility and participates in the review and approval of experiments through his position on the Reactor Committee. The reactor operations staff performs many health physict.-tyre activities, with assistance and consultation from the RCO.
The the RCO is notified of any emergency within the NRL, and, when Direct-1 or his designated alternate assists the Emergency Director in reques e evaluatix, e te emergency conditions and implementing emergency control i
measures.
12.2.2 Procedures Detailed written procedures have been prepared that address the radiation safety support that is provided for the routine operation of the UATR facility.
These procedures identify the interactions between the operational and experimental personnel and also specify numerous administrative limits and action points, as well as appropriate responses and corrective actions if these limits or action points are reached or exceeded.
The Reactor Committee, which includes a health physicist, reviews reonests for authorizatiin to use radioactive materials.
NUREG-1390 12-1 I
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12.2.3 Instrumentation A variety of detecting and measuring instruments for monitoring potentially hazardous ionizing radiation at the University of Arizona campus are available at the UATR facility.
The procedures and techniques for calibrating instruments ensure that any credible type of radiation and any significant intensities can be detected promptly and measured correctly.
12.2.4 Training All students, staff, faculty, and independent users of the UATR facility are required to participate in a health physics orientation program.
The orienta-tion lecture includes information on university, State, and Federal regulations pertaining to the use of radioactive material as well as instruction in radia-tion protection principles and practices.
Reactor operator training and requalification include lectures on radiation control tnd safety.
12.3 Radiation Sources 12.3.1 Reactor Sources of radiation directly related to reactor operations include the reactor core, the ion-exchange columns, the cooling water cleanup system, and radioactive gases (primarily argon-41).
The fission products are contained within the stainless steel cladding of the TRIGA fuel elements.
Radiation exposures from the reactor core are reduced to acceptable levels by water and concrete shielding.
The ion-exchange resins are changed routinely about once every 2 years before high levels of radioactive materials have accumulated, thus limiting personnel exposure.
Personnel exposure to the radiation from chemically inert argon-41 is limited by dilution and prompt removal of this gas from the reactor area and its discharge to the atmosphere, where it is further diluted in the unrestricted area.
12.3.2 Extraneous Sources Sources of radiation that may be considered as incidental to normal reactor operation but associated with reactor use include radioactive isotopes produced for research, activated components of experiments, and activated samples or specimens.
Personnel exposure to radiation from intentionally produced radioactive material a; well as from the required manipulation of activated experimental components is controlled by extensively developed and reviewed operating procedures that incorporate the normal protective measures of time, distance, and shielding.
12.4 Routine Monitoring 12.4.1 Fixed Radiation Monitoring System Four fixed radiation monitors are located in the UATR nuclear reactor laboratory.
Three, a continuous air monitor (CAM) and two remote air monitors (RAMS), are NUREG-1390 12-2
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located in the reactor room, and the fourth, a Geiger-Mueller (G-M) monitor, is located in the exhaust stack.
The CAM is located in the northwest corner of the reactor room and automatically shuts down the window vent and starts the high-efficiency particulate air (HEPA) (stack) vent.
It has an audible alarm and also a flashing light. The readout is in the reactor room, but it may also be observed through the control room window. The two RAMS are located at the south and west windows of the reactor room with audible alarm and readout in the con-trol room.
The G-M stack monitor is located downwind from the HEPA filter in the stack exhaust system with readout and audible alarm in the control room.
In addition to the four fixed radiation monitors, portable survey meters are available for routine use in the laboratories and for checking the radioactivity of samples removed from the core of the reactor.
Thin-window G-M tube dete-tors are used in the laboratory to monitor equipment and people for traces of rad?o-active contamination.
12.4.2 Experimental Support The health physics staff of the RCO participates in planning experiments through its membership in the Reactor Committee by reviewing all proposed procedures for methods of minimizing personnel exposures and limiting the generation of radio-active waste. Approved procedures specify the type and degree of radiation safety support required by each activity.
12.5 Occupational Radiation Exposures 12.5.1 Personnel Monitoring Program The UATR staff and other university users of the UATR facility are issued film badges by the health physics personnel of the RCO to monitor their radiation exposure.
Visitors are normally issued pocket dosimeters.
12.5.2 Personnel Exposure The annual exposure history of the UATR personnel for the last 5 years is given in Table 12.1.
The results indicate that the management of potential radiation exposure at the VATR is acceptable and well within 10 CFR Part 20 guidelines.
Table 12.1 Number of individuals in exposure interval per calendar year Number of individuals in each range Whole-body exposure range 1984 1985 1986 1987 1988 No measurable exposure 95 71 57 62 53 Less than 0.1 rem 1
0 0
0 2
More than 0.1 rem 0
0 0
0 0
NUREG-1390 12-3
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12.6 Effluent Monitorina 12.6.1 Airborne Effluents As discussed in Section 11, airborne radioactive effluents from the reactor facility consist principally of low concentrations of argon-41 (Ar-41) in the reactor room from activation of air dissolved in the pool water and from acti-vation of air trapped in the pneumatic sample transfer (rabbit) assembly.
The small amount of Ar-41 released into the reactor room from the pool is diluted by the 1.84 x 108 cm3 volume of air in the reactor room.
This air, in j
turn, is normally exhausted to the outside at the rate of 5.9 x 105 cm3/s, for a turnover rate of less than 6 minutes.
As shown in Section ll, the yearly averaged dose rate outside the UATR facility for the Ar-41 released from the pool that results from expected reactor operations is significantly less than c'
the values allowed in unrestricted areas by 10 CFR Part 20.
If the reactor room 1
ventilation system were to malfunction, trapping all the Ar-41 released at 110 kW in the reactor room, and if this condithn were allowed to continue long enough
(~9 hours) to establish an equilibrium u ncentration of Ar-41, the reactor room would contain about 1000 pCi or about 5.d x 10-8 pCi/cm2 Exposure to this con-centration of Ar-41 for 2080 hours0.0241 days <br />0.578 hours <br />0.00344 weeks <br />7.9144e-4 months <br /> per year (8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per day, 5 days per week, 52 weeks per year), which is 20 times the total projected operating time for the reactor and more than 100 times the projected operating time of the rabbit l
assembly, results in a yearly exposure of about 215 mR, which is well below the 5000 mR for restricted areas allowed by 10 CFR Part 20.
12.6.2 Liquid Effluents The reactor generates no radioactive liquid waste during routine operations.
Small quantities of liquid waste that may be generated by the changeout of the i
ion-exchange resin from the water purification system are poured back into the pool water.
12.6.3 Environmental Monitoring Thermoluminescent dosimeters (TLDs) are placed in the unrestricted area in the l
Engineering Building and in other buildings as part of an environmental moni-toring program to identify whether increased doses might result from reactor operations.
These dosimeters are replaced monthly and are analyzed by a certi-fied organization.
For 1987, the average exposure for the three dosimeters located in the Engineering Building was 10 mrem /yr greater than that for con-trol dosimeters stored in a lead pig; for eight dosimeters placed in other buildings at the university, the average exposure was 15.8 mrem /yr greater than that for the control dosimeters.
The same pattern of slightly lower doses in the Engineering Building has been observed each year since the initiation of the TLD monitoring program.
There is no noticeable correlation between the power generated by the reactor in a month and the dose on the TLDs for that month.
The staff concludes that changes in local radiation dose due to reactor
]
operations are negligible.
12.6.4 Potential Dose Assessment i
Natural background radiation levels in the University of Arizona area result in an exposure of about 90 mR/yr to each individual residing there.
At least an NUREG-1390 12-4
l additional 9 percent, about 8 mR/yr, will be received by those living in brick or masonry structures.
Any x-ray txamination for diagnostic purposes will add to the natural background radiation, increasing the total accumulative annual exposure.
Conservative calculations based on the effluents of the facility and the results of the environmental monitoring program indicate that reactor operations do not contribute significantly to the annual exposure in unrestricted areas.
12.7 Conclusion The staff concludes that the radiation protection program at the UATR currently receives appropriate support from the University of Arizona.
The staff further concludes that (1) the staf f and equipment to implement the program are ade-quate, (2) the reactor health physics staff has adequate authority and lines of communication, (3) the procedures are integrated correctly into the research plans, and (4) surveys verify that operations and procedures follow the ALARA principle.
Additionally, the staff concludes that the UATR radiation protection program is adequately based on the results of personnel monitoring and the environmental monitoring programs.
Furthermore, there is reasonable assurance that personnel and procedures will continue to protect the health and safety of the public during routine operations, as has been adequately demonstrated during the past 30 years.
NUREG-1390 12-5
13 CONDUCT OF OPERATIONS 13.1 Overall Organization Responsibility for the safe operation of the reactor facility is vested within the chain of command shown in Figure 13.1.
The Reactor Laboratory Director is
' delegated responsibility for overall day-to-day operation of the facility.
13.2 Trsd ning n
Most of the i.vaining of reactor operators is done by in-house perscnnel. The licensee's ope.ator requalification program was revised on September 15, 1989, in conjunction with this license renewal application, and the staff concludes that it meets the applicable regulations [10 CFR 50.54(1-1) and 10 CFR Part 55) and is consistent with the guidance of ANSI /ANS 15.4.
13.3 Operational Review and Audits The Reactor Committee provides an independent review and audit of facility ac-tivities.
The Technical Specifications outline the qualifications that members must possess.
The committee must review and approve plans for modifications to the reactor, new and certain classes of experiments, and proposed changes to the license or procedures.
The committee also is responsible for directing audits of reactor facility operations and management and for reporting the results thereof to the university's administration.
13.4 Emergency Planning In 10 CFR 50.54(q) and (r), the NRC requires that a licensee authorized to pos-sess and/or operate a research reactor follow and maintain in effect an emergency plan that meets the requirements of Appendix E to 10 CFR Part 50.
The licensee submitted a revised Emergency Plan dated January 30, 1990, inconjunctionwith this license renewal application. The staff concludes that tbc revisions do not decrease the effectiveness of the Emergency Plan, and that the Emergency Plan complies with the regulations.
13.5 Physical Security Plan The licensee has established and maintains a program to protect the reactor and its fuel and to ensure its security. The NRC staff has reviewed the revised Physical Security Plan in conjunction with this license renewal application and concludes that it meets the requirements of 10 CFR 73.67(f) for special nuclear material of low strategic significance.
The UATR inventory of special nuclear material for reactor operation falls within that category.
Both the Physical Security Plan dated December 8, 1989, and the staff's evalua-tion are withheld from public disclosure under 10 CFR 2.790(d)(1).
The amend-ment renewing Facility Operating License R-52 incorporates the Physical Security Plan as a condition of the license.
NUREG-1390 13-1
J 1
13.6 Conclusion On the basis of the above discussion, the staff concludes that the licensee's experience, management structure, and procedures are sufficient to provide reasonable assurance that the UATR will continue to be managed in a way that will cause no significant radiological risk to the health and safety of the public.
0 an WIN RADIATION CONTROL DEAN DIRECTOR i
7 1
HEAD of NUCLEAR and ENERGY ENGINEERING ------ 9 RADIATION DEPARTMENT CONTROL OFFICE l
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7 l
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REACTOR REACTOR REACTOR L~
HEALTH U
Y
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D SUPERVISOR COMMITTEE PHYSICIST REACTOR r
OPERATIONS e
7 Figure 13.1 Administration chart NUREG-1390 13-2
14 ACCIDEFT ANALYSIS The staff ias evaluated the documentation submitted by the University of Arizona and its analyses of potential site-specific accident events.
These analyses included the various types of possible accidents and the potential consequences to the public resulting from the operation of the UATR.
The following potential accidents or effects were considered to be sufficiently credible to warrant independent evaluation and analysis by the staff:
fuel-handling accident e
rapid insertion of reactivity (nuclear excu"sion) loss-of-coolant accident misplaced experiments mechanical arrangement of the fuel effects of fuel aging Of the potential events, the fuel-handlirg accident with the loss of cladding integrity of one irradiated fuel element in air in the reactor room would have the potential of releasing the highest level of radioactivity to the environment both inside and outside the VATR facility.
Thus, the fuel-handling accident will be designated as the maximum hypothetical accident (MHA).
The results of the analyses of the other credible accidents with less severe consequences than the MHA are also addressed in the following sections.
14.1 Fuel-Handling Accident This potential accident, designated the MHA, includes various incidents involv-I ing at least one or more irradiated fuel elements in which the fuel cladding might be breached or ruptured.
To remain conservative, the staff did not try to develop a detailed failure scenario, but assumed the limiting scenario.to be the complete cladding failure of one fuel element outside the reactor pool that instantly results in the release of all the volatile fission products that have accumulated in the free volume (gap) between the fuel and the cladding.
Several series of experiments by the fuel vendor [ General Atomics (GA)] have provided data on the species and fractions of fission products released from U-ZrH under various conditions (Baldwin et al, 1980; Foushee, 1968; Foushee x
l and Peters, 1971; Simnad, 1980; Simnad et al., 1976),
The noble gases were the principal species found to be released. When the fuel specimens were irradiated at temperatures below 350 C, the fraction released could be summarized as a constant equal to 1.5 x 10-5, independent of the temperature or operating his-tory.
GA accepts as reasonable the interpretation of these low-temperature results and concludes that the 1.5 x 10-5 release fraction reasonably could be applied to TRIGA reactors operating up to at least 800 kW.
This release fraction is, thus, applicable to the 110-kW UATR.
On the basis of the above discussion, the staff assumed a fission product release fraction of 1.5 x 10-5 of the cotal noble gas and halogen inventories for the fuel-handling accident.
On the basis of the GA analysis, this fraction NUREG-1390 14-1
is a conservative estimate of the potential release following prolonged operation at 110 kW with a maximum local fuel temperature of 120*C, or a maximum pulse of l
2.44$ with a maximum fuel temperature of 309'C.
Because the GA analysis assumes l
infinite operating time, this approach gives a con $rvatively high release value.
Because the noble gases do not condense or combine chemically, it is assumed l
that any noble gases released from the cladding will diffuse in air until radio-l active decay has reduced the concentration to an insignificant value.
Con-versely, the iodines are chemically active but are not volatile at temperatures below about 180'C.
Some of these radionuclides will be trapped by m.sterials l
with which they come in contact, such as water and structures.
Evidence indi-cates that most of the iodines either will not become or will not remain airborne I
under many acci lent scenarios that are applicable to non-power reactors (NUREG-0771, NUREG/CR-2079, Regulatory Guide 3.34).
However, to be certain that the fuel-cladding-f ailure scenario leads to upper-limit dose estimates for all possible eventt, the staff assumed that 100 percent of the iodines in the gap became airborns.
This assumption leads to computed thyroid doses that most certainly are unrealistically high for most reasonable scenarios; for example, those in whien the cladding failure occurs under water or following a significant shutdown ti.ne.
14.1.1 Scenario The staff analyzed an accident scenario for which it assumed that the cladding failure of one fuel element occurred in air and calculated subsequent doses to an individual in the reactor room and in the unrestricted area just outside the nuclear reactor laboratory.
For the cnalysis, the staff assumed that the clad-ding failure occurred in a B-ring fuel elemtat following an extended run at the maximum licensed power level of 110 kW, so that all fission products had reached their saturated activity levels.
This is a conservative assumption considering the typical operating history at the l ATR.
Normally, a significant amount of time elapses between reactor shutdown and the removal of any fuel from the reac-tor; however, the staff assumed that all fission product radionuclides were still at saturated activity levels '< hen they were released from the cladding.
All the noble gases and halogens in the fuel cladding gap were assumed to be released instantaneously from the fuel element and to be distributed uniformly in the reactor room.
No plateout was assumed.
Scenarios incorporating more realistic estimates of the above conservative assumptions would reduce the com-puted doses significantly.
However, using this scenario as a basis, the whole-body immersion dose (gamma-ray) and the potential thyroid dose f rom iodine inhalation were calculated for an individual in the reactor room (occupational) and in the unrestricted area immediately outside the nuclear reactor laboratory (public).
For the occupational exposure, the staff assumed that the total release fraction of the maximum expected inventory of radioactive materials from one B-ring fuel element was instantly released and inctantly mixed with the air in the 184-m3 reactor room.
It also assumed that the ventilation system was shut down at the time of the accident, the core contained 87 elements, and the failed element developed a power level about 1.5 times as high as that developed by an average element.
Because there is no credible way that the postulated MHA could occur without operating personnel being immediately alerted, the reactor room would be evacuated in an orderly manner within 1 minute.
For the outside exposure, the staff assumed that the unfiltered ventilation system was operating at its rated NUREG-1390 14-2
capacity of 5.9 x 105 cm2/s.
For the whole-body dose calculations, immersion in a finite cloud in both the reactor room and the unrestricted area was assumed.
14.1.2 Assessment The calculated dose for the above assumptions and locations are presented in Table 14.1.
As a result of the extremely conservative calculational and atmos-pheric assumptions, the calculated operational and public doses shown are higher than those that could occur realistically.
Table 14.1 Doses resulting from postulated fuel-handling accident Whole-body Thyroid immersion dose committed dese Exposure and location (mR)
(mR) 1-minute (occupational) 0.38 204 exposure in reactor room 1-hour (public) exposure 0.003 1.99 immediately outside the restricted area The staff, therefore, concludes that the consequences of the postulated fuel-handling accident would not pose a significant health risk to UATR reactor operations personnel, the university's student population, or the general population.
14.2 Rapid Insertion of Reactivity (Nuclear Excursion)
The U-ZrH fuel in the UATR exhibits a strong, prompt, negative temperature x
coefficient of reactivity, as discussed in Section 4.5.
This temperature coefficient acts to terminate a pulse or nuclear excursion by decreasing the reactivity as the temperature of the fuel increasas.
These results have been verified at many operating TRIGA reactors.
Although it may be theoretically possible to rapidly insert sufficient excess reactivity to create an excursion during which fuel damage would occur before the excursion could be terminated, the reactivity limits imposed by the Technical Specifications of the VATR are intended to preclude such an event.
In the VATR, full withdrawal of the transient rod results in a reactivity in-sertion greater than the authorized maximum pulse insertion.
For this reason, the maximum withdrawal of the transient rod drive system is limited by position-ing switches and by a mechanical block so that the maximum worth of the transient rod in any mode of operation is limited to 2.50$.
Because of this restriction imposed on the transient rod drive system, the excess reactivity remaining on the transient rod below the mechanical block (~0.93$) is not considered during a credible inadvertent transient.
The Technical Specifications for the UATR limit the maximum allowed pulse reactivity insertion and the maximum worth of
'the transient rod to 2.50$.
NUREG-1390 14-3
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14.2.1 Scenario The staff considered the scenario of the reactor operating at a power level between 0 and 110 kW, at which time all of the available licensed excess reactiv-ity, which is 3.25$ for the cold, clean core, is inserted because of some uniden-tified failure mechanism.
Because all of the 3.25$ is not available at higher power because of temperature feedback effects, the staff found that the higher the temperature at which the rapid insertion of all the available excess reac-tivity is initiated, the lower the final temperature of the fuel immediately after the transient.
Therefore, the staff assumed the worst case to be initia-tion of a 3.25$ transient with the core at ambient temperature and essentially zero initial power.
The potential significant consequences of the reactivity insertion accidents considered by the staff were melting of the fuel or cladding material and failure of the cladding as a result of high internal gas pressures and/or phase changes in the fuel matrix.
The primary cause of cladding failure l
at elevated temperatures in stainless-steel-clad TRIGA fuel elements would be i
excessive stress buildup in the cladding caused by hydrogen pressure resulting from disassociation of the ZrH ; however, calculations performed by GA and con-x firmed by many reactor pulses indicate that the integrity of immersed cladding i
is maintained at peak fuel temperatures as high as 1175'C (Coffer et al., 1966; Simnad, 1980; Simnad et al., 1976).
14.2.2 Assessment The staff reviewed the literature for large reactivity insertions into reactor cores similar to that of the VATR.
GA has performed many experiments with reactivity insertions as high as 5.00$ in an 85-element stainless steel TRIGA core.
GA measured, among other parameters, the temperature of the fuel in the hottest core position and examined fuel elements afterward (Coffer et al., 1906; Simnad et al., 1976).
There was no indication of undue stress in the cladding and no indication of either cladding failure or fuel melting.
The measured maximum temperature for the 5.00$ pulse was about 750'C, and the estimated peak transient temperature at any localized point in the fuel was 1175 C.
Because the radial temperature distribution in a fuel element immediately following a pulse is similar to the radial power distribution, the peak transient temperature immediately after the pulse is located at the periphery of the hottest fuel ele-ment.
It will fall rapidly (within seconds) as the heat flows toward the clad-ding and toward the fuel center.
It also was observed that for the 5.00$ pulse, the maximum measured pressure rise within an instrumented fuel element was far below the predicted equilibrium value at the peak temperature (Coffer et al.,
1966; Simnad et al., 1976; West, 1970).
On the basis of the above analysis, the staff concludes that the rapid insertion i
of the 3.25$ available excess reactivity into the VATR core will not result in fuel melting or cladding failure because of high internal gas pressure or high temperature.
Therefore, there is reasonable assurance that the fission products contained in the fuel will not be released to the environment as a result of the rapid insertion of reactivity.
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14.3 Loss-of-Coolant Accident The rapid loss of cooling water, along with its inherent shielding effects, immediately following reactor operation is a potential accident that would re-sult in increased fuel and cladding temperatures and increasec' radiation levels NUREG-1390 14-4 1
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in the react ? room.
Because water is required for the moderation of neutrons, the loss of coolant in the reactor would terminate the neutron chain reaction and, thus, would terminate the production of fission power.
However, the resid-ual radioactivity resulting from fission product decay would continue to deposit heat energy in the fuel and would constitute an unshielded radiation source in the bottom of the tank.
The complete loss of cooling water at the UATR is considered to be an extremely unlikely event.
All water lines enter the reactor tank from the top of the vessel; therefore, draining through an improperly aligned valve is not possible.
Water can only be pumped from the vessel to the suction break that still pro-vides for about 14 ft of water above the core.
The steel tank liner of the vessel itself and the associated concrete structure were designed and thoroughly tested to ensure tightness against leakage.
For the loss-of-coolant accident (LOCA), however, it is assumad that some failure mechanism, no matter how indued could lead to a leakage of water from the pool.
The level of water in the pool would then descend to a level below the core.
14.3.1 Scenario The licensee analyzed the case of an instantaneous and complete loss of cool-ing water at the UATF. as the limiting LOCA event.
It assumed the reactor had operated at full licensed power of 110 kW for a period sufficient to build up the maximum invento y of radioactive fission products just before the loss.
The loss of moderator terminated the neutron chain reaction but fission product decay continued to heat the fuel. The licensee assumed that only natural ther-mal convection of air up through the core would remove this decay heat.
Dose levels were calculated for four positions in the reactor area.
Time was measured from the cessation of 110-kW operation. The results of this study are presented in Table 14.2.
Table 14.2 Radiation doses from uncovered care at the UATR following the maximum loss-of-coolant accident Pose rate (R/hr) location 10 s I hr 24 hr At floor level above reactor 195 58 28 In Room 216 above reactor room
- 5. 8 1.7 0.8 In reactor room 6.6 ft from pool 1.5 0.9 0.4 In hall 13.1 f t from pool 0.4 0.2 0.1 It is clear that dose rates of this magnitude would require the Engineering Building to be evacuated.
Following this evacuation, there would be no further increase in the risk to the public because the radiation from the unshielded core would be collimeted upward by the tank and shield structure toward the uninhabited upper floors of the Engineering Building.
NUREG-1390 14-5 a
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.Using very conservative assumptions, the licensee calculated a maximum fuel temperature of 160*C, reached in about 25 minutes following core uncovery.
For fuel temperatures in this range, the changes in pressure inside the fuel clad-ding are caused by thermal expansion of any confined air and are small compared with the pressures at the temperatures discussed in Section 14.2.
14.3.2 Assessment The staff reviewed the licen.cee's analysis and concurs in the cssumptions and methods.
On the basis of the above considerations, the staff concludes that a LOCA at the UATR is a very unlikely event.
However, should a LOCA occur, it would not lead to fuel damage or consequent release of radioactivity to the environment or undue radiation exposure of the public.
14.4 Misplaced Experiments This type of potential accident is one in which an experiment sample or device inadvertently is located in an experimental facility where the irradiation con-ditions could exceed the design specifications.
In that case, the sampl9 might become overheated or develop pressures that could cause a failure of the experi-ment container.
As discussed in Sections 10 and 13, all new experiments at the UATR facility are reviewed before insertion, and all experiments in the region of the core are separated from the fuel cladding by at least one barrier, such as the pneumatic transfer tubes, the rotary specimen rack irradiation capsules, or the central thimble.
The staff concludes that the experimental facilities and the procedures for review of experiments at the UATR facility are adequate to provide reasonable assurance that failure of experiments is not likely, and even if failure occurred, the reactor fuel cladding would not be breached.
Furthermore, if an experiment should fail and release radioactivity within an experimental facility, there is reasonable assurance that the amount of radioactivity released to the env h n-ment would not be more than that of the proposed maximum hypothetical accident (MHA).
l 14.5 Mechanical Rearrangement of the Fuel This type of potential accident would involve the failure of some reactor sys-tem, such as the support structure, or could involve an externally originated event that would disperse the fuel and in so doing would cause breaches of the cladding of one or more fuel elements.
The staff has not developed an operational scenario for such accidents.
- However, it is conceivable that a heavy weight, such as a lead tranuer cask, could be dropped on the reactor core from above and could smash the core in such a way as to cause breaches of the claddin; of one or more fuel elements with the con-sequent release of radioactive materials.
Because of the dissolution of the hal-ogens in the pool water, the staff concludes that the quantity of fission prod-ucts released to the room air as a result of this accident would be lower than that released as a result of the fuel-handling accident (MHA) evaluated in Sec-tion 14.1, even if more than a single fuel element was damaged.
Therefore, the staff concludes that there is reasonable assurance that no credible inadvertent mechanical rearrangement of fuel would result in an accident with more severe consequences to the public than the MHA.
NUREG-1390 14-6
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14.6 Effects of Fuel Aging The staff has included a discussion on the phenomena of fuel aging in this section in order to address all credible effects that might contribute to the release of airborne radioactivity to unrestricted areas.
However, fuel aging should be considered normal with reactor operation and is, in fact, expected to occur gradually.
Reactions internal to the fuel cladding are discussed below.
There is evidence that the U-ZrH fuel tends to fragment with use, probably x
because of the stresses caused by high temperature gradients and the high heating rates during pulsing operations (Simnad 1980; Simnad et al., 1976; West, 1970).
Possible consequences of fragmentation include (1) a decrease in thermal conductivity across cracks leading to higher central fuel temperatures during operation and (2) an increase in the amount of fission products released into the cracks in the fuel.
With regard to the thermal conductivity, hot-cell examination of thermally stressed hydride fuel bodies has shown relatively widely spaced radial cracks that would cause minimal interference with radial heat flow (Simnad, 1980; West, 1970).
However, after pulsing TRIGA reactors have exhibited an increase in both steady-state fuel temperatures and power reactivity coefficients.
At power levels of 1 MW, the temperature has increased by about 23C* and power reactivity coefficients have increased by about 20 percent (AFRRI, 1960; General Atomic, 1965).
GA has attributed these changes to an increased gap between the fuel material and the cladding (caused by rapid fuel expansion during pulse heating) that reduces the heat transfer coefficient.
Experience has shown that the observed changes occur mostly durinp the first several pulses and have essen-tially resulted in saturation after 100 pulses.
Therefore, the UATR should not experience any further changes in the thermal conductivity caused by the maximum allowed 2.50$ pulse.
With regard to the fission products GA has identified two mechanisms for their release from TRIGA fuel (Foushee, 1968; Foushee and Peters, 1971; Simnad, 1980).
The first mechanism for fission product release is fission-fragment recoil into connected gaps within the fuel cladding.
This effect predominates up to about 400*C and is independent of fuel temperature.
GA has postulated that, in a closed system such as exists in a TRIGA fuel element, fragmentation of the fuel material within the cladding will not cause an increase in the fission product release fraction (Foushee, 1968).
The reason for this is that the total free volume available for fission products remains constant within the confines of the cladding.
Under these conditions, the formation of a new gap or widening of an existing gap must cause a corresponding narrowing of an existing gap at some other location.
Such a narrowing allows more fission fragments to traverse the gap and become embedded in the fuel or cladding material on the other side.
In a closed system, the average gap size and, therefore, the fission product release rate remain constant regardless of the degree to which fuel material is broken up.
The second mechanism for fission product release, which occurs above about 400'C, is diffusion, and the amount released depends on fuel temperature and i
NUREG-1390 14-7
._t...._._._
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a fuel surface-to volume ratio.
However, release fractions used for safety eval-uation are based on conservative calculations that assumed a degree of fuel I
fragmentation greater than expected in actual operation.
Additionally, because j
the maximum allowed pulse at the UATR (2.50$) results in a maximum fuel temper-ature of only 309'C, this mechanism would have a negligible effect on fission product release.
Because the two likely effects of aging of the U-ZrH fuel moderator will not x
have a significant effect on the operating temperature of the fuel or on the assumed release of gaseous fission products from the cladding, the staff con-cludes that there is reasonable assurance that fuel aging will not increase significantly the likelihood of fuel-cladding failure or the calculated conse-quences of an accidental release in the event of the loss of cladding integrity at the UATR.
14.7 Conclusion The staff has reevaluated the creditie accidents for the UATR on the basis of an increase in the maximum authorized power level from 100 kW to 110 kW and con-cludes that the postulated acciden, with the greatest potential effect on the environment is the loss of cladditg integrity of one irradiated fuel element in air in the reactor room. The analysis of this accident, however, shows that even if more than one UATR fuel element failed simultaneously, the expected dose equivalent in restricted and unrestricted areas still would be significantly below the guideline values of 10 CFR Part 20.
Of the other accidents analyzed, the only one that would be even slightly altered by the increased licensed power is the loss-of-coolant accident (LOCA).
As shuwn in Section 14.3, the probability of a LOCA is extremely remote at the UATR and even if the accident were to occur under the most pessimistic assumptions, the maximum core temperature would remain sificantly below that necessary to induce any fuel or cladding failure.
The staff, therefore, concludes that the design w the facility and the Technical Specifications provide reasonable assurance that the UATR can be operated with a low probability of accidents and that even the maximum hypothetical accident would pose no significant risk to the health and safety of the UATR reactor operations personnel, the university's student population, or the general public.
NUREG-1390 14-8 6
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15 TECHNICAL SPECIFICATIONS The staff has evaluated the licensee's Technical Specifications in this licensing action. These Technical Specifications define certain features, characteristics, and conditions governing the operation of the UATR facility and are explicitly included in the renewal license as Appendix A.
The staff has reviewed the format and contents of the Technical Specifications using ANSI /ANS 15.1-1982, " Standard for the Development of Technical Specifications for Research Reactors," as a guide.
On the basis of its review, the staff finds the Technical Specifications acceptable and concludes that norr=1 reactor operation within the limits of the Technical Specifications will not result in offsite radiation exposures in excess of 10 CFR Part 20 guidelines.
Furthermore, the limiting conditions for operation and surveillance requirements will limit the likelihood of malfunctions and mitigate the consequences to the public of off-normal or accident events.
NUREG-1390 15-1
16 FINANCIAL QUALIFICAT10NS The UATR is owned and operated by a State educational institution in support of its role in education and research.
On the basis of financial information supplied by the licensee in its submittal.of.0ctober 17, 1988, the staff con-cludes that funds will be made available, as necessary, to support continued operations and eventually to shut down the facility and carry out decommis-sioning activities. The staff reviewed the licensee's financial status and found it acceptable in accordance with the requirements of 10 CFR 50.33(f).
NUREG-1390 16-1
17 OTHER LICENSE CONSIDERATIONS 17.1 Prior Reactor Utilization In the previous sections of this SER, the staff concluded that the risk of radiation exposure to the public as a result of the normal operation of the reactor is insignificant and that only an off-normal or accident event could cause some measurable exposure.
The maximum hypothetical accident (MHA) was shown to result in potential radiation exposures that were within the applicatt e l
guideline values of 10 CFR Part 20.
The staf f concluded that the reactor was initially designed and constructed to operate safely.
During the review for license renewal, the staff considered whether prior operation would have caused significant degradation of the capa-bility of components and systems to continue to perform their safety functions.
Because fuel cladding is the component most responsible for preventing the release of fission products to the environment, possible mechanisms that could lead to detrimental changes in its integrity were considered.
Prominent among the considerations were the following:
(1) radiation degradation of cladding integrity, (2) high fuel temperature or temperature cycling leading to changes in the mechanical properties of the cladding, (3) corrosion or erosion of the cladding leading to thinning or other weakening, (4) mechanical damage resulting from handling or experimental use, and (5) degradation of safety components or systems.
The effects of fuel aging are discussed in Section 14.6 of this SER.
The staff's conclusions regarding these parameters, in the order in which they were identified above, are:
(1) Nearly identical fuel has been laboratory tested elsewhere and has been exposed in similar irradiations to higher total radiation doses in oper-ating reactors, such as at General Atomics and the University of Illinois.
No significant degradation of cladding has resulted in any of these reactors.
(2) The power density, coolant flow rates, and maximum temperatures reached in the UATR fuel are below similar parameters in s e other non power reac-tors using similar fuel.
No damage has occurred during normal operations in any of these reactors.
(3) Water flow through the core is obtained 5y natural thermal convection.
Therefore, erosion effects that might res01t from high flow velocity will be negligible.
Corrosion is kept to a minnum by careful control of the conductivity of the primary coolant.
(4) The fuel is handled as infrequently as possible, consistent with required
' surveillance and experimental program requirements.
Any indications of possible damage or degradation are investigated promptly, and damaged fuel will be removed from service in accordance with the Technical Specifica-tions.
All experiments placed near the core are isolated from the fuel cladding by a water gap and at least one barrier of encapsulation.
i NUREG-1390 17-1
(5) The UATR staff performs regular preventive and corrective maintenance and replaces components as necessary.
Nevertheless, some malfunctions of equipment have occurred.
The staff review, however, indicates that most of these malfunctions have been one-of-a-kind incidents.
There is no indication of significant degradation of the instrumentation, and there is strong evidence that any future degradation will lead to prompt reme-dial action by the VATR staff.
Therefore, there is reasonable assurance that there will be no significant increase in the likelihood of a reactor accident occurring as a result of component malfunction.
17.2 Conclusion In addition to the considerations above, the staff has reviewed event reports f' rom the licensee and inspection reports and informal comments from the NRC regional office.
On the basis of this review, the staff concludes that there has been no significant degradation of equipment and that facility management will continue to maintain and operate the reactor so that there is no signif-icant increase in the radiological risk to the employees or the public.
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18 CONCLUSIONS On the basis of-its evaluation of the application as set forth in the previous sections, the staff has determined the following:
(1) The application filed by the University of Arizona for the renewal at an increased operating power level of Operating License R-52 for.its research reactor, dated October 17, 1988, as supplemented on July 17 and September 15, 1989, and January 30, 1990, complies with the requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commis-sion's regulations set forth in 10 CFR Chapter I.
(2) The facility will operate in conformity with the application as amended, the provisions of the Act, and the rules and regulations of the Commission.
(3) There is reasonable assurance (a) that the activities authorized by the operating license can be conducted without endangering the health and safety of the public and (b) that such activities will be conducted in compliance with the regulations of the Commission set forth in 10 CFR Chapter I.
(4) The licensee is technically and financially qualified to engage in the activities authorized by the license in accordance with the regulations of the Commission set forth in 10 CFR Chapter I.
(5) The renewal of this license will not be inimical to the common defense and security or to the health and safety of the public.
NUREG-1390 18-1
19 REFERENCES Armed Forces Radiology Research Institute (AFRRI), " Final Safeguards Report for the AFRRI TRIGA Reactor," Appendix A Docket No. 50-170, November 1960.
Baldwin, N.
L., F. C. Foushee, and J. S. Greenwood. " Fission Product Release from TRIGA-LEU Reactor Fuels," Seventh Biennial TRIGA Users' Conference, San Diego, California, March 2-5, 1980.
Code of Federal Regulations Title 10. " Energy," U.S. Government Printing 6Tfice, Washington, D.C., revised annually.
Cof fer, C. 0., J. R. Shoptaugh, Jr., W. L. Whittemore, " Stability of U-ZrH TRIGA Fuel Subjected to Large Reactivity Insertion," General Atomic Report GA-6874, San Diego, California, January 1966 [ transmitted by letter dated July 25, 1967 (Docket No. 50-163)].
Foushee, F. C., " Release of Rare Gas Fission Products From U-ZrH Fuel Material,"
Gulf General Atomic Incorporated Report GA-8597, San Diego, California, March 1968.
--, and R. H. Peters, " Summary of TRIGA Fuel Fission Product Release Experiments," Gulf Energy and Environmental Systems, Inc., Report Gulf-EES-A10801, San Diego, California, September 1971.
General Atomic Company, "Thermionic Research TRIGA Reactor Description and Analysis," General Atomic Report GA-5400, Rev. C, November 1, 1965 [ transmitted by letter dated February 28, 1966 (Docket No. 50-227)].
Letter, R. L. Morgan, Department of Energy, to H. Denton, NRC, May 3, 1983.
--, George W. Nelson, University of Arizona, to Document Control Desk, NRC,
" Application for Renewal of Operating License R-52," October 17, 1988.
--, George W. Nelson, University of Arizona, to Theodore S. Michaels, NRC, providing additional information for the license renewal, July 17, 1909.
--, George W. Nelson, University of Arizona, to Theodore S. Michaels, NRC, providing additional information for the license renewal, September 15, 1989.
Simnad, M. T., "The U-ZrH Alloy:
Its Properties and Use in TRIGA Fuel,"
x General Atomic Report GA-4314, E-117-833, San Diego, California, February 1980.
--, F. C. Foushee, and G. B. West, " Fuel Elements for Pulsed TRIGA Research Reactors," Nuclear Technology, 28:31-56, 1976.
University of Arizona, "SAR for the University of Arizona TRIGA Research Reactor," Docket No. 50-113, 1988.
NUREG-1390 19-1
--, " Technical Specifications for the University of Arizona TRIGA Research Reactor," Docket No. 50-113, 1988.
West, G. B., " Safety Analysis Report for the Torrey Pines TRIGA Mark III Reactor," General Atomic Report GA-9064, San Diego, California, January 5,1970,
[ transmitted by letter dated January 29, 1970 (Docket No. 50-227)].
INDUSTRY CODES l
American National Standards Institute /American Nuclear Society, ANSI /ANS 15.1,
" Standard for the Development of Technical Specifications for Research Reactors," La Grange Park, Illinois, 1982.
--, ANSI /ANS 15.4, " Selection and Training of Personnel for Research Reactors," 1977.
1 i
--, ANSI /ANS 15.11, " Radiological Control at Research Reactor Facilities,"
- 1987, i
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96RC tons sat U.S. NUCLE AR RtGULAfoRY CotARAISGloII
- 1. RE PoMi NUMBE R In? im.
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- Rn.sm -
BIBLIOGRAPHIC DATA SHEET tsa 4,me,ws.as ei vu res==>
- 2. TITLE AND &USitTLE Safety Evaluation Report Related to the Renewal of the
' Operating License for'the TRIGA Training and Research jiE REPoRWBLISHED -
3.
Reactor at tne University of Arizona I
- rl May 1990 4 FIN on GRANT NUMBER 4
- 6. AUTHonts)
- 6. TYPE or REPORT
- E
" #I Same'as Box 8 7, PE Rloo CovERE D parausse osass N12 AT loN - N AME AND ADDRESS far mac, pm.e p=.m., one e, messen, u.s sucean aspubsery ce
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- 8. P Fy NG on Division of Reactor Projects III, IV, V & Special Projects Office of Nucicar Reactor Regulation
' U.S. Nuclear _ Regulatory Commission Washington, DC' 20555
- 9. EPoNSORING ORGANIZATION - NAME AND ADonESS fif sec, ever Seme es eseve~s #f re=reww.awmer WAC Deem Onm er esema, u.& Nester.-
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- 10. SUPPLEMENT ANY NOTES Pertains to Docket No. 50-113 11 A85 TRACT (Jeposee er mmJ This Safety Evaluation Report for the application filed by the IWrersity of Arizona for the renewal of Operating License R-52 to continue operati g I' research reactor at an' increased operating power level has been prepared by the Otiice of Nuclear
'RE. actor Regulation of the U. S. Nuclear Regulatory Commission. The facility is located on the University of Arizona campus in Tucson, Arizona. The staff concludes that the reactor can continue to be operated by the University of Arizona without ' endangering the health and safety of the public.
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