ML20247D947
| ML20247D947 | |
| Person / Time | |
|---|---|
| Site: | 05000113 |
| Issue date: | 07/17/1989 |
| From: | Nelson G ARIZONA, UNIV. OF, TUCSON, AZ |
| To: | Michaels T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8907260008 | |
| Download: ML20247D947 (59) | |
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TH E U NIVE RSITY_ OF
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TUCSON, A RIZON A 85721 COLLEGE OF ENGINEERING AND MINES
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' DEPARTMENT oF NUCLEAR AND ENEriGY ENGINEERING l
i July 17. 1989 l
. United States Nuclear Regulatory Commission l
Document Control Desk Washington, D.C.
20555 ATTN:
Theodore S. Michaels, Project Manager C
Standardization and Non-Power Reactor Directorate
- Office of Nuclear Reactor Regulation License R-52, Docket'No.-50-113 r-
Dear.Mr.~Michaels:
Enclosed is the additional information for the' license renewal for the. University of Arizona TRIGA reactor which you. requested in your letter of May 16,.1989.
The enclosed material includes the answers to the 16 questions, the ALARA statement signed by the Vice-President for Research, a' revised copy of the Facility Technical Specifications with the changes in-corporated which you suggested at our meeting in Tucson, and a revised copy of page 50 of. the SAR.
Please let me know if there is'any other additional.information needed.
Sinc ely, 7
eorge W. Nelson, Director
- Nuclear Reactor Laboratory GWN/sh
- . i ;
Oi g 8907260008 890717 PDR. ADOCK 05000113 P
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__________..______._____._..___._..________.=____._..___.._m.__-_.m___m
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9 United States Nuclcar Regulatory Commission
- uly 10,1989 Document Control Desk Wasieinatcn, Df. 20555 ATTN
- Theodore S. Michaels, Project Manager Standardization and Non-Power Reactor Project Directorate Office of Nuclear Reactor Regulation License R-52, Docket No. 50-113
SUBJECT:
Additional information for License Renewal of the University of Arizona TRIGA Reactor requested by your letter of May 16,1989 Question 1:
How close is the nearest residence to the reactor and where is it locattd?
Answer 1:
The nearest residence is Yuma Hall, a student dormitory, located 300 feet west of the Reactor
' Laboratory. The nearest private residences which are not under University of Arizona control (thus I
excluding dormitories or fraternity houses) are to the west of the University. These are houses on the west of Tyndah Avenue south of First Street, which are located approximately 1300 feet west of the laboratory.
The nearest private residences in the direction of the prevailing winds (southwest to northeast) are located on the north side of Speedway Blvd. between Highland and Rincon Avenues, and are also approximately 1300 feet from the laboratory. There are niso houses on the north side of Helen Avenue, north of Speedway Blvd., which are about 1500 feet due north of the Reactor Laboratory.
l Question 2:
Is there any heavy industry in the vicinity of the University of Ari:ona campus and if so, where is it located?
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How far is Interstate 10, Southern Pacific and AMTRAK from the reactor?
Answer 2:
e There is no heavy industry in the vicinity of the campus. Interstate 10 is, at its nearest,1.75 miles
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west of the Reactor Laboratory. The tracks for AMTRAK and Southern Pacific railroads are 1.1 miles south of the reactor laboratory. There are light industries in the area adjacent to the railroad tracks, about one mile south of the laboratory.
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Question 3:
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I The hydrology section, Section D. pg.7,'should contain information that quantifies the flood potential at the i
l site. For example, provide the maximum historic flood elevation of nearby streams and compare them to the elevation of the site. Provide judgements as to whether or not there is any potential to flood the site and f
reactor building or other safety related equipment, then discuss the effects of flooding such as the concentration of radionuclides at the secure boundary and how they compare to 10 CFR 20 limits.
Answer 3:
j The Engineering building is located on North Campus Drive in the northwest quadrant of the University. This location is near the top of the local watershed area. Precipitation falling near the building drains westward on North Campus drive, eventually entering Tucson Arroyo,, which flows into the Santa Cruz River. Precipitation falling 1000 feet north of the building drains northward on Mountain Avenue I
into Navajo Wash, which flows westward into Cemetery Wash and Flowing Wells Wash, and then into the Santa Cruz River. Precipitation falling 500 feet to the south of the building flows south into High School Wash, which flows into Tucson Arroyo and thence into the Santa Cruz River.
The nearest stream bed, which is normally dry, is High School Wash, which at its nearest is 0.6 miles directly south of the Engineering Building. The 100-year flood level at this location is 2407 feet above saa level. The areas which are 100 feet or more north of High School Wash (including the University of Arizona campus), are outside the 500-year flood level.
Tucson Arroyo at its nearest is approximately 0.9 miles southwest of the Engineering Building. At l
this location the 100-year flood level is 2387 feet above sea level.
l The Santa Cruz River is, at its nearest,1.8 miles west of the Engineering Building. At this location, the 100-year flood level is 2328 feet above sea level.
Navajo Wash, at its nearest, is 2.1 miles north of the Engineering Building, and its 100-year flood level is 2369 feet above sea level.
The first floor of the Engineering Building, at which the reactor is located, is 2439.5 feet above sea level and 6 feet above the level of North Campus Drive. The physical contours of the site and the drainage path provided by North Campus Drive will not permit accumulation of water around the building. for example, the elevation of North Campus Drive at the east end of the building is 2433.26 feet, and at the 1
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o west end of the building is at 2430.56 feet. Thus water cannot accumulate on the ground outside the I
reactor laboratory, but will instead drain into North Campus Drive and flow to the west.
From these data, it is clear that there is no potential for flooding of the reactor or the Engineering I
Building originating either from the near or distant stream beds or from accumulation of local precipitation.
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Water originating within the building, as for example, from a broken utility water pipe, would drain into the basement floor level of the building, and cannot accumulate on the first floor where the reactor i
j laboratory is located.
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Should flooding of the reactor laboratory occur, the only effect would be to transport some of the i
reactor pool water, diluted by the flood waters, into the hall of the building and out of the outside doors, l
l Table 7.4 on page 70 of the SAR lists the isotopes measured in reactor pool water after operation at 100
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I kilowatts for two hours. Undiluted, these concentrations in water are below the 10 CFR 20 Appendix B limits, and do not present a significant hazard if released beyond the secure boundary. Because of their j
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short halflivet these isotopes cannot increase in concentration due to biological or chemical processes. The 1
undiluted tritium content in pool water is a factor of 200 lower than the MPC allowed for release of tritium
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1 to unrestricted ares.
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l Question 4:
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i Submit justification for a power increase that would permit reactor trip testing and power level setpoint l
calibration costsistent with operation at a power level of 100 Kw.
Answer 4:
Under the previous licewe for the University of Arizona TRIGA Reactor, steady-staa cperation of the reactor was conducted at or below a maximum power level of 100 kilowatts thermal. The safety channels (reactor trips) have been set to scram at less than 110 kilowatts. (They are set at 107-108 kilowatts I
to provide a margin for power calibration variation). The facility management and the reactor committee
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agree that renewal of the operating license must permit continued operation at the same steady-state power l
l level, that is,100 kilowatts. This is the standard level for most irradiations performed for research and
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teaching using neutron activation analysis, and is the level at which, for example, neutron flux levels, l
magnitudes and spectra have been determined.
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In order to permit reactor trip testing and power level setpoint calibration, the renewed license must permit operation at a maximum power level of 110 kilowatts. Foc this reason, the safety analysis calculations have been repeated for postulated operation at 110 kilowatts. The results of these calculations
'are included in the answers to questions 6, 9,12 (a),12 (b),13 (a), and 13 (b). No additional safety considerations are significant for 110 kilowatt operation compared to 100 kilowatt operation.
Question 5:
IVhat is the maximum allowed core excess reactivity in the Technical Specifications?
Answer 5:
The maximum core excess reactivity in the Technical Specifications is $3.25. This amount of reactivity is sufficient to provide a positive period to bring the reactor to critical at a desired low power -
level with the transient rod fully inserted. This is necessary in preparation for pulsing. An extrapolation 0:
peak powers, energy releases, and fuel temperatures from the measured data in Table 3.2 on page 28 of the SAR has shown that for a reactivity insertion of $3.25, the peak power will be 1660 Mw, the energy release 33.0 Mw-sec, and the maximum fuel temperature will be 4600 C. Thus, this amount of reactivity is low enough that it will not exceed the safety limit of 1000oC if it were all inserted instantaneously.
Question 6:
Are there credible paths for reactor pool water to get into the campus water or the sewer system? If so, please assess the radiological consequences.
Answer 6:
There are no credible paths for reactor pool water to get into the campus water system or sanitary sewer system.
A limited amount of water could enter the storm sewer system if it es< aped from the demineralized system because of pipe breaks or leakage from the pump. The amount of leakage from the pool is limited by underwater siphon break holes in the pool inlet and outlet pipes. With the normal level of pool water, the pool level would decrease 5 inches and would release a maximum of 100 gallons of water. If the pool
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l water were at its maximum allowable level before leakage, a maximum of 280 gallons could escape.
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As recorded in the Environmental Impact Appraisal, the tritium content of reactor pool water is J
1 about 0.015 microcurks per liter, which is low in comparison to the 3.0 microcuries of tritium per liter which is the MPC for water discharged to an unrestricted area (10 CFR 20 Appendix B Table II).
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Since there is no mechanism by which trace amounts of tritium entering the storm sewer could become concentrated, the maximum possible release,280 gallons (1060 liters) of water, containing.015 microcuries of tritium per liter into the storm sewer, does not constitute a radiological hazard.
1 Table 7.4 on page 70 of the Safety Analysis Report lists the other radioisotopes measured in reactor pool water after two hours of operation at 100 kilowatts. These data were used to compute the saturated concentration of the isotopes with the longest halflives, Na-24 and Mn-56, which would be obtained after l
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operation for infinite duration at 110 kilowatts. In each case, the concentrations are lower than the MPC i
permitted for release to an unrestricted area without dilution. For Na-24, the MPC is.0002 microcuries/ml and the saturated concentration would be 0.0001 microcuries/ml. For Mn-56, the MPC is.0001 i
i microcuries/ml, while the saturated concentration would be.000034 microcuries/ml. These isotopes have j
short halflives, which would make reconcentration in a biological or chemical system impossible. Thus the I
l release of up to 280 gallons of this water into the storm sewer does not constitute a radiological hazard.
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Question 7:
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?! case discuss the operation of the HVAC systems serving the reactor laboratory, including at least the following:
(a) The air flow rates into and out of the laboratory for both normal and ci>. vmal reactor operation, including accident scenarios.
l Answer 7 (a):
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Air flows into the reactor, control, and storage rooms through ducts from a building HVAC system, l
l This system is not dedicated to the Reactor Laboratory, but also serves other rooms on the north and west sides of the Engineering Building. There are no air return ducts back into the building HVAC system.
l The exit for air which has entered these rooms is through the opening for the window fan, located in the north wall of the reactor room (room 124),9.2 feet above floor level in the room, and 12.5 feet above the i
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outside ground level. A.ir flows out of this opening at the rate of about 1190 cfm when the fan is on (it is I
required to be on during reactor operation). When the window fan is off and the stack fan is also off, air still flows out of this window fan opening, being driven out by the force of the air entering the rooms from l
l the building HVAC system. Under these conditions, some of the room air also escapes through the filter and up the stack and some escapes through spaces around doors and windows.
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If radioactivity is detected in the air by the particulate air monitor, the activation of the air monitor I
i alarm will automatically cause the stack fan to be turned on, exhausting air through the filter and up th, stack. If the window fan is on at the time of the air monitor alarm, it will be turned off.
Testing of flow patterns with an artificial smoke system designated for such testing shows that when
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I the stack fan is operating at 580 cfm, there is no flow of room air out of the window fan opening, but rather flow of outside air into the room through this opening. When the stack fan is operating at its low speed,380 cfm, most of the flow is up the stack. but there is a small amount of flow detectable out the window fan opening. To prevent this flow, a minimum air removal rate of 500 cfm will be required from the stack fan.
The stack fan exhausts air up to the stack on the roof of the Engineering Building, releasing it about 50 feet above ground level. The fan unit and motor are located in the attic of the Engineering Building. Keys are not available to students, faculty, or staff for access to the roof or attic. These keys are evallable to University Police, Radiation Control Office personnel (who maintain TLD and environmental dosimetry on the roof), and maintenance personnel.
A local on/off switch in the control room permits testirig of fan operation prior to reactor operation.
The following summarizes the air flow in the Reactor Laboratory-Normal situation, reactor shutdown:
Air enters through the building HVAC system incoming air ducts on the south wall of the reactor room, on the east wall of room 124A, and on the ceiling of the control room. Air exits through the l
window fan opening on the north wall of the reactor room, being driven by the pressure of the incoming air. Some air also exits through the stack fan duct to the roof, and through spaces around windows and I
i doors.
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Normal operations, reactor in operation:
Air enters through the building HVAC system incoming air ducts on the south wall of the reactor room, on the east wall of room 124A, and on the ceiling of the control room. ' Air is pulled out by the 1
window fan on the north wall of.the reactor room and is exhausted outside 12.5 feet above ground level.
No air leaves the laboratory by any other route.
Abnormal operations, stack fan activated:
Air enters through the building HVAC system incoming air ducts on the south wall of the reactor room on the east wall of room 124A, and on the ceiling on the control room.. Air is pulled out by the stack fan on the west wall of the reactor room, passing through a prefilter and absolute filter, and is exhausted 50 feet above ground from the stack on the building roof. No air leaves the reactor room by any other route.
- Question 7 (b):
How it is ensured that airborne radioactivity cannot enter the HVAC system of the rest of the engineering building under any conditions of either the HVAC system and the reactor and its special ventilation systems.
Answer 7 (b):
During conditions of airborne radioactivity, the stack fan is operated. It has been demonstrated with artificial smoke that air is pulled into the stack under these conditions and that air does not leave the reactor room to the outside through the window fan opening. It has also been demonstrated that no air leaves the reactor room through the spaces around the doors to the outside or by spaces around the door to the hall of the building.
Question 7 (c):
How is flow from the reactor laboratory to adjacent regions of the Engineering Building prevented during reactor operations?
i Answer 7 (c):
l l-During conditions of reactor operation, the window fan is in operation. It has been demonstrated l
w2th artificial smoke that air is pulled out of the reactor room with this fan operating, and is prevented I
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1 fr m c:*:rin;; the bui! ding through the HVAC ducts or through spaces around doors.
Question 8:
Please provide a description of the ALARA program at the University of Ari:ona reactor facility. Also a
- definite ALARA statement signed by high University officials is necessary.
Answer 8:
a) An ALARA statement has been issued from the Vice President for Research, who is the University official in charge of all research programs and facilities. A copy of this statement is attached.
b) The document, RADIATION CONTROL PROGRAM FOR THE UNIVERSITY OF ARIZONA I
TRIGA REACTOR LABORATORY includes the following ALARA policy statement:
"It is the policy of the University of Arizona TRIGA Reactor Laboratory that all radiation doses shall be kept as low as is reasonable achievable, taking into account the state of technology for measuring and controlling radiation exposure, and the feasibility and economics of improvements relative to the benefits to health and safety. By this it is meant that radiation will be restricted to a small part of that allowable by federal and state regulations. As a goal, "a small part" for this facility will be ten percent."
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This program provides a summary of policies and requirements for controlling radiation exposure in the Reactor Laboratory, including health physics services; radiation surveys and monitoring; instrumentation requirements and calibration; control, inventory, and storage of radioactive materials; training of personnel; contamination control; review of personnel radiation exposures which are greater than the minimum detectable level; and visitor control.
A training memorandum which is issued to, read by, and discussed with every applicant for I
personnel radiation dosimetry film badges for use in facilities of the Department of Nuclear and Energy
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l Engineering gives as a guideline that "...all unnecessary radiation doses should be avoided, and all necessary i
work which might cause radiation exposure should be engineered in a way to minimize the dose." This guideline has been effective in maintaining personnel doses within the department as low as reasonable
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achievable.
Question 9 a):
' Assuming :ero escape of the Ar-41 from the pool water in your analysis is nonconservative, assume a
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. conservative and defendable release fraction and recalcidate dose rates in the restricted and unrestricted areas.
Answer 9 a):
As reported in the SAL (page 64), the production rate of Argon-41 in the reactor pool water was measured 'to be 0.201 microcuries/ liter / hour. The rate of exponential decrease in time for Argon-41 concentration after shutdown at the surface of the pool was determined to be.00627 +/.00019 per minute
. (page 65) and the rate of radioactive decay is.00631 per minute. The 3-sigma upper limit for the rate of decrease in Argon-41 concentration is 00053 per minute greater than the rate of radioactive decay. At 3 feet below the surface, the 3-sigma upper limit is.00016 per minute greater than the radioactive decay rate.
This indicates that the leakage of Argon-41, if present, includes only the water near the surface. Assuming that for the whole quantity of water from 0 to 2 feet (with volume 1880 liters) the release rate is.00053 per minute, a very conservative estimate, and for the water from 2 to 4 feet it is.00016 per minute, then for the top 1880 liters of water the leakage is 0.201 x 1880 x.00053/(.00631+.00053) = 29.2 microcuries per hour, and for the next lower 1880 liters it is 8.8 microcuries per hour. Thus one may conservatively assume that Argon-41 is released from the pool at less than 38 microcuries per hour. Assuming 2080 hours0.0241 days <br />0.578 hours <br />0.00344 weeks <br />7.9144e-4 months <br /> of operation at 100 kilowatts per year (which is extremely conservative, the actual hours being about 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> per year), the release would be 79 millicuries per year.
Question 9 b):
IVhat is the predicted Ar-41 exposure to the nearest residence?
Answer 9 b):
A calculation was performed to determine the dose rates due to a continuous release of 30 microcuries per hour of Argon-41 into the air of the reactor room. Within the reactor room, the beta plus gamma dose rate would be 0.058 mrem / hour with the fan turned off, and 0.0021 mrem / hour with the window fan operating at 1100 cfm. With the fan operating at i100 cfm, exhausting air at 12 feet above ground level and assuming outside air speed of 1 meter /second with the most conservative weather conditions (Pasquill condition F--moderately stable), an atmospheric dispersion calculation shows that the maximum dose rate outside the reactor laboratory would be 0.00011 mrem / hour (0.11 microrem/ hour) at a distance of 100 meters from the point of release in the direction of the wind. This is on the order of one
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percent of natural background radiation in the area. The maximum dose rate at the distance of the nearest private residence (360 meters) in the direction of the wind would bo 0.000033 mrem / hour (0.033 microrem/ hour) for the 30 microcurie / hour Ar-41 release. This calculation is based on extremely
' conservative calculations based on measurements conducted at 100 kilowatts. As an estimate for operation at 110 kilowatts, the l>ostulated releases and dose rates should be multiplied by a f act of 1.1.
j, Question 10:
In Table 4.1 of the Environmental Impact Appraisal:
(a) in 1983-84 one person received a total dose in the range 101-500 mR. What was the source of l
1 this radiation?
Answer 10 (a):
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a) In 1983 one person in the Department of Nuclear and Energy Engineering received a total dose 1
of 130 mrem whole-body in film badge exposures during a seven-month period from en experimental research setup containing a radioactive tape head. This research was not associated with the reactor 1
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laboratory and was conducted in a separate laboratory of the Department. Autoradiography of magnetic tape driven across this head was used to study the head wear pattern. The radioactive tape head was obtained from a commercial supplier. An exposure of 45 mrem was recorded during the montil the system was constructed and the tape head installed, and lesser exposures totalling 85 mrem were recorded during the following six months of operation, maintenance, and disassembly of the system.
Question 10 (b; Please provide the information requested in the Table in 10 CFR 20.497 for the years 1984,85,86,87,and 1988.
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. Answer 10 (b):
4 The information requested in the Table in'10 CFR 20.407 for 1984-88 for the Department of -
Nuclear and Energy Engineering is given below:
Number of individuals with whole body exposure in rems for the calendar year.
j 1984-1985 1986 1987 1988 1
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No measurable '
exposure '
95 71 57 62 53
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0.1 rem 1:
0 0
0 2
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. Greater than 0.1 rem.
0 0
0 0
- 0 1.
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Questioti ! !"
IVhat is the natural background radiation level in the University of Arizona area?-
Answer 11:'
~ The natural background radiation level in the University of Arizona area has be'en determined from readings of TLD's which are placed on the roofs of ten University. Buildings'and read monthly by an independent contractor. This background level in 1988 was 86.9 mrem / year, or 10 microrem/ hour, relative to control TLD's which are shipped, stoad, and analyzed together with the building TLD's but are stored in a lead pig'(for shielding) while at the University. The readings of these control TLD's, about 90 1
mrem / year, has been subtracted from the readings of the TLD's placed on the University buildings to-1 L
obtain the 86.9 mrem / year. Most of the control background reading which is subtracted may be due to j
i expos.re during shipping to and from the contractor's facility in Sunnyvale, California and storage prior to I
j and after the monthly exposure period.
l Question 12:
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The NRC position, on the basis of the Columbia Unirersity hearings is that the maximum hypothetical l
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- accident for a TRIGA is the instantaneous failure in air and consequent release into the reactor room air of
.all the fission products in 'the fuel element gap, containing the highest power density, immediately following operation at the maximum authori:ed powsr level of sufficient length for all these fission products to reach their saturated activity levels. Assuming this scenario and using defendable conservative techniques:
a) Calcidate the IVhele-body Immersion Dose and the Thyroid Committed Dose in the restricted area
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over s time span sufficient to evacuate this restricted area (1 to 5 min):
Anrver 12 (a):
A conservative calculation was made to determine the maximum hypothetical dose to persons in the reactor room at he time of a fuel element failure, in this calculation, it was assumen that the reactor had operated continuously for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day for 100 days at the power level of 110 kilowatts. It was then assumed that failure of the clad of a B-ring fuel element caused the release of all fission product gases in the gap between the fuel meat and clad. It was further assumed that all of the fission product gases bubbled to the pool surface withuut being retained in the water and were dispersed in the air of the reactor room. The fraction of fission product gases accumulating in the gap of the TRIGA fuel element was taken to be 1.5 x 10-5, a value which has been established by measurements with TRIGA fuel (Simnad, General Atomic Report E-117-833,1980). The total release of iodine isotopes 131 through 135 under these circumstances is 6.5 millicuries.
The calculation assumed that a person remained in the reactor room for a period of 60 seconds after the bubble of fission product gases reached the surface of the pool and dispersed in the room, and that the stack fan was operating 'at 500 cfm at this time. Because of its size and the location of exit doors,60 l
l seconds is adequate time for a person to evacuate the reactor room.
The dose rate due to both beta particle and gamma radiation was computed by modelling the geometry of the room as an equal-volume hemisphere. This assumption is conservative, since it brings all parts of the distributed source as near as possible to the person. For this hypothetical situation, it was calculated that the person would be exposed to an average gamma dose rate of 2.91 mrem / hour and an average beta dose rate of 19.74 mrem / hour for 60 seconds, for a total beta plus gamma dose of 0.38 mrem.
Had the person remained in the room for 5 minutes instead of 60 seconds, the total beta plus gamma dose would be 0.94 mrem.
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The person was also assumed to breathe air containing the fission product gases at the ICRP
" standard man" rate of.000347 cubic metert per second for 60 seconds before evacuating the reactor room.
This breathing rate and the assumed concentration of iodine isotopes in the room air were used to determine the ingestion of radioactive iodine to the lungs and from this, to determine the dose commitment to the thyroid. Literature values of dose commitment factors were used for the iodine isotopes which give a significant dose. (Lamarsh, Introduction to Nuclear Engineering,1983). These values assume that 23 percent of the iodine ingested by breathing is deposited in the thyroid, that the mass of the thyroid is 20 grams, and that the biological halflife of iodine is 138 days. Breathing air with the assumed radioiodine concentration would lead to the ingestion of 0.877 microcuries, with a dose commitment to the thyroid of 203.5 mrem. Of this dose coramitment, the majority (107 mrem) is due to I-131, even though the ingested activity of this isotope (0.072 microcuries) is les; than 10 percent of the total.
This hypothetical calculation gives a much greater dose and dose commitment than is physically possible with the University of Arizona TRIGA for several reasons The computation assumes that during the time span of 100 d'
, twice as much 11 ermal power is produced by the reactor as has been produced l
in the entire 30-year history of reactor aperation at the University. A highly conservative but realistic reactor power history reduces tha iodine inventory and thyroid dose commitment by more than a factor of
- 10. The hypothetical computation described above also assumes that all fission product gases are released instantaneously from the gap. However, a hole in the clad would release fission product gas only until the absolute pressure in the gap is reduced to the external pressure, which is about 1.5 atm at the depth of the core. The computation also assumes that iodine in the gap between fuel meat and clad is in a gaseous form.
However, the melting point of iodine is 113.5 degrees Celsius, and its boiling point is 184.35 degrees Celsius. The clad temperature of TRIGA fuel at 100 kilowatts remains far below 100 degrees Celsius under any conditions of operation, so little of the iodine in the gap which is contact with the clad is in gaseous form. Furthermore, if iodine were released from the fuel, it would not remain in gaseous form, but would l
be captured by the water, whose temperature is well below the boiling point of iodine. Thus the above calculations, while demonstrating that the hypothetical release of iodine from the University of Arizona TRIGA reactor does not present a significant radiological hazard, do yield risk values which are orders of magnitude greater than could be present in a real situation of fission product release.
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Question 12 (b):
Calculate the same doses for the public exposure immediately outside the restricted area over 'a time span I
i sufficient to evacuate this area (1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />).
Answer 12 (b):
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The assumptions listed above were used to compute the exposure of persons remaining outside the 1
reactor room for two hours aftes the hypothetical gaseous fission products were released.' It was assumed that the fan was exhausting room air containing the fission product gases at the rate of 500 cfm, and that
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l there was no turbulence to facilitate dispersal, but only a wind of 1 meter /second. Beta and gamma dose
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1 rates were computed with an equivalent-volume hemisphere model, the volume consisting of the total J
amount of outside air which ha+ e d with the fission product gases since the time of first release. Under these conditions, a person et. side the reactor facility would receive a beta dose of.00034 mrem and a i
gamma dose of.0021 mre a during the two hours. During this time 5950 picoeuries of iodine would be ingested by breathinh, givmi a dose commitment to the thyroid of 1.99 mrem. Since the same conservative assumptions were used to de ti rmine the source of radiciodine as were used in the previous calculation, these 1
I calculated radiological riske s also larper by orders of magnitude than the risks which could be present in a real situation.
The tabulations demonstrated that neither the doses or dose commitments in 12 (a) or 12 (b) are i
significantly different for other fan speeds of 400,1000, or 1250 cfm.
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Question 13:
Assuming an instantaneous loss of vessel water:
a) IVhat is maximum fuel temperature?
I Answer 13 (a):
For a loss of pool water, fission power will cease, and the instantaneous heat generated from delayed 1
fission product decay may be computed by the Borst-Wheeler model. A D-ring fuel element in the
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University of Arizona TRIGA has the highest steady-state power generation of 1.83 kilowatts, and thus will have the highest power level due to fission product decay after shutdown. The maximum temperature of 3
this fuel element during operation at 110 kilowatts is 120 degrees Celsius, this may be taken as the initial i
temperature of the fuel before additional heat is added from fission product decay.
The heat capacity of the TRIGA fuel element is well known to be.0012 degrees C/ watt-second.
4 Step solutions using Newton's law of cooling were performed as a function of time using two different air temperatures (50 degrees and 100 degrees Celsius) as boundary conditions for natural convection cooling of the fuel. The area for natural convection heat transfer was taken as the fuel surface area,0.17 square meters.
The step solution showed that after shutdown, following infinite operation at 110 kilowatts, the instantaneous power decreased from 120 watts after I second to 75 watts after 10 seconds, to 48 watts after 100 seconds. For natural convection by 50 degree C air, the fuel reaches a maximum temperature of 128.9 i
degrees C after 300 seconds, and then decreases in temperature. For natural convection by 100 degrees C air, the fuel reaches a maximum temperature of 160 degrees after 1500 seconds, and then decreases.
i.
Neither of these maximum temperatures will cause the fuel to lose its integrity. TRIGA fuel may remain at temperatures far in excess of 160 degrees for prolonged periods without failure.
It should be noted that NUREG/CR-2387 (page 11) reports that the maximum temperature of air-1 l
cooled TRIGA fuel after infinite operation at 250 kilowatts would be less than 150 degrees C. However, l
the temperature of the air providing natural convection cooling for those calculations is not specified, so the i
values cannot be compared.
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Question 13 (b):
What is the resultant dose rate, i.e., what would Table 7.1 values be at 10 sec.?
Answer 13 (b):
In order to determine the dose rate as a result of complete loss of pool water, the computation detailed on pages 54 through 57 of the SAR were repeated with the following modifications.
a) The assumed power level was 110 kilowatts, and the reactor was assumed to operate continuously for 10,000 days prior to the loss of water shielding.
b) The times for which dose rate calculations were performed were 10 seconds, I hour, and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
c) The correction for attenuation within the fuel meat was refined, and attenuation in the graphite
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end plugs and stainless steel end fitting, with Taylor buildup factors, was included.
The data provided in Table 7.1 on page 57 for the postulated power history (40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per week,52 l.
weeks per year) are replaced by the following table for continuous operation as outlined above:
1 Table of Calculated Gamma Ray Dose Rate with no water l^
Shielding the Reactor Core at Various Times after Shutdown After 10 sec.
After I hour After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> At floor level above pool 195 rem /hr 58 rem /hr 28 rem /hr In room 216 above reactor room 5.8 rem /hr 1.7 rem /hr 0.8 rem /hr
- In reactor room 2 meters from pool 1.5 rem /hr 0.9 rem /hr 0.4 rem /hr in hall 4 meters from pool' 0,4 rem /hr 0.2 rem /hr 0.1 rem /hr Question 14:
Page 11, last paragraph, Give details of the history of the used stainless-clad elements. When were they new, where were they used, how were they used--pulses, steady-state power level, any history of failures from among the " lot"?
Answer 14:
The TRIGA fuel e!,.nents are stainless steel clad, 8.5 wt percent uranium in zirconium hydride (Zr.H ratio equal to 1:1.7), corresponding to General Atomic Catalog No.103. The enrichment of uranium in U-235 is slightly less than 20 percent. As new elements, they contained nominally 38 grams each of U-235.
These fuel elements when new were installed in the General Atomic Mark III reactor in 1966 and operated nearly continuously at powers from 1 to 1.5 Mw. They were subsequently removed from the reactor and put in storage, the latest of them in July,1968. They were stored in storage casks at the General Atomic facility until December 1972, when they were shipped to the University of Arizona. The
9:
. fuel was in the Mark Ill to run a thermionics experiment which was funded by AEC. The Mark III h
operated at a maximum steady state thermal power level of 1500 kilowatts, with 127 grid positions (a G-i.
w ring grid plate). These fuel elements were not pulsed during their time of use at General Atomic, because the Mark Ill reactor did not have pulse capability.
u The total burnup of the 87 fuel elements when received from General A1,nic was 422 grams, or an average burnup of 4.9 grams (of the original 38 grams) of U-235 per element. This corresponds to a thermal energy release from these 87 elements cf about 340 Megawatt-days. The fuel element with minimum burnup of the lot, according to General Atomic records, had 0.8 grams U-235 consumed, or about 37,2 grams remaining. The fuel element with the maximum had a burnup of 9.1 grams U-235.
. While this latter amount of burnup reduces the reactivity available from a TRIGA fuel element,it does not signify a reduction of cladding strength or integrity. It is still useful for operation of a low-power TRIGA reactor, where a large amount of excess reactivity is not needed to overcome the large temperature coefficient of reactivity which is characteristic of TRIGA fuel.
The single failure among the 87 previously-used fuel elements obtained from General Atomic in -
1972 occurred in 1975 after an element was accidentally dropped from the fuel element handling tool.
About two weeks after the fuel element was dropped, a leak in the bottom weld occurred. This fuel element has been isolated with a special cap which precludes its transfer from its permanent storage location J
in the reactor pool into the core. The fuel element handling tool which was in use when the fuel element was dropped has been replaced.
In addition to the 87 previously-used fuel elements obtained in 1972, five new fuel elements H
L (TRIGA Catalog No.103) were obtained in 1978, and an instrumented fuel element (TRIGA Catalog No.
l 503) was obtained in 1975. Except for the failed element described above, these fuel elements have been operated very successfully since 1973 at powers up to 100 kilowatts and for 1600 pulses up to $2.45 in reactivity. During this time the reactor has been operated for teaching and research at the University of Arizona for 134.8 Megawatt-hours, or about 8.4 Megawatt-hours per year. The additional burnup as a result of operation at the University of Arizona is about 7 grams for the whole core, or about 0.07 grams of U-235 per fuel element.
j With water purity maintained at the present level, and with operation at or below 100 kilowatts l'
steady state and pulses below $2.50, the fuel should continue to maintain its integrity far beyond the license
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period.
Qeestion 15:
Please resubmit page 50 of your SAR so that the last line trads "at a maximum of 110 kilowatts"&
Answer 15:
As requested, a corrected version of page 50 of the SAR is attached.
Question 16:
Resubmit your Technical Specifications to include the changes agreed upon during our meeting of May 3,
- 1989, 1-l.
Answer 16:
A revised copy of the Technical Specifications, with the changes agreed upon during our meeting of May 3,1989, is attached.
The content of this response and revisions to the Technical Specifications have been reviewed by the Reactor Committee for this facility. If other information is needed for your license review, please let me know. My telephone number is (602) 621-2565, and the FAX number for my department office is (602) l l
621-8096.
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George W. Nelson Reactor Director GWN/sh ec: Reactor Committee i
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TH E UNIV ERSITY OF ARIZON A b
d TUCSON, ARIZONA 85724 l
RADIATION CONTROL OFFICE 1827 E. MABEL(B/210)
(602) 626-6850 j
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i-J A.L.A.R.A. Policy Statement
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'The principle of keeping radiation doses from all University of Arizona licenced operations As Low As Reasonably Achievable (A.L.A.R.A.), forms the basis for the University's radiation protection program. The principle of A.L.A.R.A. will be applied wherever possible in order to minimize occupational exposures to radiation and releases of radioactive material.
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ichael A. Cusanovich, Ph.D.
Vice President for Research l
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..,, Maximum Fuel Temperature -
In a review 2 of the analysis of credible accidents for TRIGA reactors the-fuel-moderator temperature is identified as the crucial safety parameter. The limiting temperatures are given as ll500C for ZrH,, and 1000oC for ZrH,7 TRIGA -
3 3
fuel. For this reason, for the University of Arizona TRIGA, fuel temperatures must remain below 1000oC under any mode of operation. The fuel temperature limit of.
1000 C will be the safety limit for the facility.
Fission Product Retention Measurements at General Atomic 2 have demonstrated that a very small fraction of fission products are released from the ZrH fuel material during and after reactor operation. This release fraction depends upon the fuel temperature, and has the value of 1.5 x 10-5 for fuel temperatures below 350'C. Operation of the University of Arizona TRIGA in the steady-state mode maintains fuel temperatures.
well below 350oC. For this reason, the release fraction value of 1.5 x 10-5 is appropriate to use for the calculations reported here.
Postulated Power History 4
' The license for the University of Arizona TRIGA Reactor is intended to permit legal operation of the reactor under all conditiota permitted by the Technical Specifications. This will include operation at power levels not exceeding 110 kilowatts thermal. The Technical Specification limitation on steady-state reactor power will give a requirement for safety channels to scram the reactor with setags at a maximum of 110 kilowatts. It is clearly not intended that an occurrence where 2Hu.vley, S.C., and Kathren, R. L., NUREG/CR-2387, Credible Accident Analyses for a TRIGA and TRIGA-Fueled Reactors,1982, page 15.
2M. T. Simnad The U-ZrH Alloy: Its Properties and Use in TRIGA Fuel, GA Report E-Il7-833, February 1980.
1 1.
j INDEX l
Page Number q
l F-1.0 DEFINITIONS 1
l 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 5
2.1 Safety Limit - Fuel Temperature 5
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2.2 Limiting Safety System Setting - Steady State Re.
Power Level 6
2.3 Limiting Safety System Setting - Pulse Mode Reactor Power Level 7
3.0 LIMITING CONDITIONS FOR OPERATION 8
3.1 Reactivity Limits 8
I 3.2 High Power Operation 10 j
3.3 Pulse Operation 11 3.4 Reactor Instrumentation 12 3.5 Reactor Safety System 13 3.6 Ventilation System 15 3.7 Experiments 16 4.0 SURVEILL4NCE REQUIREMENTS 17 4.1 Fuel 17 4.2 Control Rods 18 4.3 Reactor Safety System 19 4.4 Radiation Monitoring Equipment 20 4.5 Maintenance 21 4.6 Pool Water Conductivity 22 5.0 DESIGN FEATURES 23 5.1 Reactor Fuel 23 5.2 Reactor Building 24 5.3 Fuel Storage 25
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- 6.0' ADMINISTRATIVE CONTROLS 26 6.1 Organization 26 6.2 Review 27 6.3.1 Operating Procedures 28 6.3.2 ALARA Program 28 6.4 Action to be Taken in the Event a Safety Limit is Exceeded 29 6.5 Action to be Taken in the Event of a Reportable Occurrence 30 6.6 Plant Operating Records 31 6.7 Reporting Requirements 32 6.8 Review of Experiments 35 l
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- l. 1.0 DEFINITIONS Channel - A channel is a combination of sensors, electronic circuits, and output devices connected by the appropriate communications network in order to measure and display the value of a parameter.
Channel Calibration - A channel calibration is an adjustment of a channel such that its output corresponds with acceptable accuracy to known values of the parameter which the channel measures.
Calibration shall encompass the entire channel, including equipment, actuation, alarm, or trip and i
shall include a Channel Test.
Channel Check - A channel check is a qualitative verification of acceptable performance by i
observation of channel behavior. The verification shall include comparison of the channel output with previous readings or performance or with other independent channels or systems measuring the same variable, whenever possible.
Channel Test - A channel test is the introduction of a signal into the channel for verification that it is operable.
Cold Critical - The reactor is in the cold critical condition when it is critical with the fuel and bulk water temperatures the same (~200C).
Experiment - An experiment is any device or material, not normally part of the reactor, which is introduced into the reactor for the purpose of exposure to radiation, or any operation which is designed to investigate non-routine reactor characteristics.
Experimental Facilities - Experimental facilities are the thermal column, pneumatic transfer systems, central thimble, rotary specimen rack, beam tube, and the in-core facilities.
Limiting Conditions for Operatn - Limiting Conditions for Operation (LCO) are administratively established constraints on equipment and operational characteristics which shall be adhered to during operation of the reactor.
Limiting Safety System Setting (LSSS) - The LSSS is the actuating level for automatic protective devices related to those variables having significant safety functions.
Manual Mode - The reactor is in the manual mode when the reactor mode selection switch is in the manual or automatic position. In this mode, reactor power is held constant or is chaned on periods of approximately one second or longer.
Measured Value - The Measured Value is the value of a parameter as it appears on the output of a channel.
Movable Experiment - An experiment is movable when it is intended that all or part of the experiment may be moved in or tear the core or into and out of the reactor while the reactor is operating.
Opera _b_le_ - Operable means a component or system is capable of performing its intended function.
Operating - Operating means a component or system is performing its intended function.
. Pulse Mode - The reactor is in the pulse mode when the reactor mode selection switch is in the l
pulse posid~on. In this mode, reactor power may be increased on periods less than one second by motion of the transient control rod.
Reactivity Worth of an Experiment - The reactivity worth of an experiment is the maximum value of the reactivity change that would occur as a result of planned changes or credible malfunctions that alter experiment position or configuration.
Reactor Committee - The group of persons at the University who are assigned responsibility for review and audit of facility operation and review of changes and experiments in accordance with 10 CFR 50.59.
Reactor Operating - The reactor is operating whenever it is not secured or shutdown.
Reactor Safety Systems - Reactor Safety Systems are those systems, including associated input channels, which are designed to initiate automatic reactor protection or to provide information for initiation of manual protective action.
Reactor Secured - The reactor is secured when:
a.
It contains insufficient fissile material or moderator present in the reactor, adjacent j
experiments or control rods, to attain criticality under optimum available conditions of moderation and reflection, or b.
1.
The minimum number of neutron absorbing control rods are fully inserted or other safety devices are in shutdown position, as required by Technical Specifications, and 2.
The console key switch is in the off position and the key is removed from the lock, and 3.
No work is in progress involving core fuel, core structure, installed control rods, or control rod drives unless they are physically decoupled from the control rods, and 4.
No experiments in or near the reactor are being moved or serviced that have, on movement, a reactivity worth of one dollar or more.
Reactor Shutdown - The reactor is in a shutdown condition when sufficient control rods are inserted to assure that it is suberitical by at least $1.00 of reactivity.
Reportable Occurrence - A Reportable Occurrence is any of the following which occurs during reactor operation:
a)
Operatiott with actual safety-system settings for required systems less conservative than the limiting safety-system settings specified in Technical Specification 2.2.
b)
Operation in violation of limiting conditions for operation established in the Technical Specifications.
4
. c)
A reactor safety system component malfunction which renders or could render the reactor safety system incapable of performing its intended safety function unless the malfunction or condition is discovered during maintenance tests or periods of reactor shutdown.
d)
Any unanticipated or uncontrolled change in reactivity greater than one dollar.
e) -
Abnormal and significant degradation in reactor fuel, cladding, or coolant boundary which could result in exceeding of prescribed radiation exposure or release limits, f)
An observed inadequacy in the implementation of either administrative or procedural controls which could result in operation of the reactor outside the limiting conditions i
for operation.
g)
Release of radioactivity from the site above limits specified in 10CFR20.
Control Rod - A control rod is a device fabricated from neutron absorbing material or fuel which is used to establish neutron flux changes and to compensate for routine reactivity losses. A control rod may be coupled to its drive unit allowing it to perform a safety function when the coupling is disengaged.-
Transient Rod - The transient rod is a control rod with scram capabilities that is capable of providing rapid reactivity insertion to produce a pulse.
Safety Limit - A Safety Li.mit is a limit on an important process variable which is found to be necessary to reasonably protect the integrity of certain of the physical barriers which guard against the uncontrolled release of ' radioactivity. The principal physical barrier is the fuel element cladding.
l Secured Experiment A Secured Experiment is any experiment. experimental facility, or component l
of an experiment that is held in a stationary position relative te the reactor by mechanical means.
i The restraining forret must be substantinhy greater than those to which the experiment -might be i
subjected by hydraulic, pneumatic, buoyam, or other forces which are normh1 to the operatirag environment of the experiment, a by 'Ifrees which can arise as the result of credible (malfunctions.
Shall, Should, and May - Tha word "shall" is used to denote 3 requirement, the word thwld" denotea a recommendation, and the mord *may" denotes permission, neither a reavire62etit neer a recommendation, l
Shutdown Margin Shutdown hfargin is the reactivity existing when the most reactive control rod is Elly withdrawn from the core and the other control rods are fully inseited into the core.
Time Interyal + Th6 average over say titended period fo.r each surveilbnce time item shall be the normal surv31ance time, eg, for a two-year Diterval the sverage 1, hall not excetdt two years.
a)
Biennially - at two-year intervals (interval not to exceed 30 months) b)
Annually - at one-year intervals (interval not to exceed 15 months) c)
Semiannualb - at 6-month intervals (interval not n exceed seven and one-half months) d)
Quarterly - at 3 -runnth intervah Timerval not to cr.ceed four months) 1 e)
Monihty - at one month interva'is (interval not lo exceed cix weeks)
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Weekly - at seven-day intervals (interval not to exceed ten days) g)
Daily - (must be done during the calendar day) '
Any extension of these intervals shall be occasional and for a valid reason and shall not affect the average as defined.
Untried Experiment - An untried experiment is any experiment not previously performed in this reactor.
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1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS' j
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L 2.1 Safety Limit - Fuel Temperature l
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Applicability This specification applies to the reactor fuel temperature.
Objective.
The objective is to define a fuel temperature below which it can be predicted with confidence I
that no damage to the fuel elements will occur.
Specification The temperature'of the fuel shall not exceed 1000 oC under any conditions of operation.
Basis The recommended limiting design basis parameter for TRIGA fuel is the fuel temperature. A fuel temperature safety limit of 11500C for stainless-steel-clad U-ZrH.85 TRIGA fuel is 1
recommended as a design value to preclude the loss of clad integrity when the clad temperature is below 500oC (Simnad, GA Report E-117-833, The U-Zr H Alloy: Its -
Properties and Use in TRIGA Fuel, Feb.1980, p. 4-1).. The criterion for assuring the integrity of a TRIGA fuel element at the University of' Arizona is that the fuel temperature be maintained below 1000oC, which is well below the recommended value. It has been shown by analysis and by measurements on other TRIGA reactors that a power level of 1000 kw corresponds.to a' peak fuel temperature of approximately 400oC. Pulsing with a reactivity input of $3.25 will give a peak fuel temperature of approximately 4600C.
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2.2 Limiting Safety Smtem Setting - Steady State Reactor Power Level l
. Applicability '
f This specification applies to the reactor power level safety system setting for. steady state operation.
Objective-The. objective is to assure that the Safety Limit is not exceeded.
t,
' Specification The setting for the power level scram in steady. state operation shall be no greater than -
110 kw..
Basis Calculations and measurements show that at 110 kw, the peak fuel temperature in the core will be less than approximately 150oC which is well below the safety criterion of.1000aC and
- provides an' ample safety margin to accommodate errors in measurement and anticipated operational transients.
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... 2.3 Limiting' Safety System Setting - Pulse Mode Reactor Power Level Applicability This specification applies to the reactor power level safety system setting for pulse mode operation.
Objective The objective is to assure that the fuel temperature specified by the Safety Limit is not exceeded in pulse mode operation.
Specification The setting for the peak power level scram in pulse mode operation shall be no greater than 1100 Mw.
Basis Calculations and measurements show that at a peak power of 1100 Mw in pulse mode operation, the peak fuel temperature in the core will be less than approximately 4000C. This provides an ample safety margin to accommodate errors in measurements and anticipated operational transients.
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1 3.0 LIMITING CONDITIONS FOR OPERATION 3.1 Reactivity Limits l
Applicability These specifications apply to the reactivity condition of the reactor, and the reactivity worths
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of control rods and experiments, and apply for all modes of reactor operation.
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l Objective The objective is to assure that the reactor can be shut down at all times and to assure that the j
safety limit will not be exceeded.
I Specifications The reactor shall not be operated unless the following conditions exist:
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a.
The shutdown margin referred to the cold xenon-free condition is greater than $0.50 I
with the highest worth rod fully withdrawn and with the highest worth non-secured j
experiment in its most positive reactivity state.
j b.
Any experiment with a reactivity worth greater than $1.00 is secured so as to prevent unplanned reactivity removal from or insertion into the reactor; c.
The reactivity available to be inserted by the pulse rod is determined and il lim:ted by a mechanical block to a maximum of $2.50, d.-
The reactivity worth of an individual experiment is not more than $3.00; e.
The total of the absolute values of the reactivity worth of all experiments in the reactor is less than $5.00;
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f, A ramp or oscillating rod placed in the reactor cannot add more than $1.00 of reactivity; I
g.
The drop time of e_;h standard control rod from the fully withdrawn position to 90 l
percent of full reactivity insertion is less than one second; and j
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The neutron count rate on the startup channel is greater than one count per second.
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The maximum reactivity insertion rate by control rods for non-pulsed operation is less than $0.20/second.
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The maximum excess reactivity does not exceed $3.25.
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The height of coolant water above the core is 14 feet or greater.
1.
The bulk temperature of coolant water does not exceed 450 C.
m.
The conductivity of the coolant water, averaged over 30 days, does not exceed 5 micrombos/cm.
Basis The shut down margin required by specification 3.la is necessary so that the reactor can be shutdown from any operating condition and ' remains shutdown after cooldown and xenon decay even if one control rod should stick in the fully withdrawn position.
Specification 3.lb is based on pulse measurements and analysis at the University of Arizona which indicate that as much as $3.00 reactivity could be inserted without increasing fuel temperature by more than 415'C. By restricting each non-secured experiment to a reactivity worth of one do!!ar, an ample margin is provided.
Specifications 3.lc through 3.lf are intended to provide additional margins between those values of reactivity changes encountere.d during the course of operations involving experiments and those values of reactivity which, if exceeded, might cause a safety limit to be exceeded.
Specification 3.lg is intended to assure prompt shutdown of the reactor in the event a scram signal is received.
Specification 3.1h is intended to assure that sufficient neutrons are available in the core to provide a signal at the output of the startup channel during approaches to criticality.
Specification 3.li is based on analysis at the University of Arizona which demonstrates that with a reactivity ramp of S.20/second starting at delayed critical at any power below 100 kilowatts, the maximum fuel temperature increase will be less than 750C, and thus will not exceed the Safety Limit.
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... 3.2 liigh Power Operation Applicability This specification applies to operation of the reactor at high steady-state power.
Objective l
The objective is to prevent inadvertent pulse operation of the reactor while it is at a high power level.
Specification The reactor shall not be operated in the steady-state mode at power levels above 10 kw unless, in addition to the conditions of Section 3.1, the transient rod is fully withdrawn to the limit of its limiting switch.
Basis This specification is intended to prevent inadvertent pulse operation when the fuel temperature is above 500C (corresponding to a power level of 10 kw) as measured in the B-ring. See Specification 3.3b.
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3.3 E Polse Operation Applicability These specifications apply to operation of the reactor in the pulse mode.
. Objective The objective is to prevent the fuel temperature safety limit from being' exceeded during -
pulse mode operation.
Spec.ifications The reactor shall not be operated in the pulse mode unless, in addition to the requirements of-
' Section 3.1, the following conditions exist:
- a. -
The transient rod is set such that the reactivity worth upon withdrawal is not greater
' than $2.50; and i
o b.'
The temperature of the fuel immediately prior to the pulse is essentially in equilibrium with the bulk water temperature. This is controlled by limiting the reactor power prior to pulsing.
Basis
' Specification 3.3a will maintain the maximum temperature of the fuel after a pulse below 400oC above the bulk pool temperature, and thus well below the 1000oC fuel safety criterion.
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. 3.4 Reactor Instrumentation Applicability This specification applies to the information which must be aimilable to the reactor operator during reactor' operation.
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Objective l
The objective is to require that sufficient information is available to the operator to assure safe operation of the reactor.
Specification The reactor shall not be operated unless the measuring channels described in the foil twing table are operable and observable from the control room:
i Minimum Operating Mode Number in which Measuring Channel Operable Required Reactor Power Level (Linear) 1 Steady State Wide-range Log Power i
Steady State Level (Startup count rate)
Reactor Period 1
Steady State Reactor Power Level (high range) 1 Pulse Mode i
Reactor Tank Water Temperature 1
All Modes Area Radiation Monitors 2
All Modes Particulate Air Radiation Monitor i
All Modes Reactor Water Activity Monitor 1
All Modes Basis The neutron detectors assure that measurements of the reactor power level are adequately covered at both low and high ranges.
The radiation monitors provide information to operating personnel of radiation above a preset level so that there will be sufficient time to evacuate the facility or take action to prevent the release of radioactivity to the surroundings.
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4 h 3.5. Reactor Safety System Applicability 1
This specification applies to the reactor safety system channels and interlocks.
Objective The objective is to require the minimum number of reactor safety system channels and interlocks that must be operable in order to assure that the safety limits and the LCO's are not exceeded.
Specification The reactor shall not be operated unless the safety system channels and interlocks described in the following tables are operable.
Minimem Operating Mode Safety Systmem or Number in which Measuring Channel Operable Function Required Setpoint Rcactor Power Level 2
Scram Steady State not above 110 kw Reactor Period 1
Audible Alarm Steady State no shorter than 2 seconds Peak Reactor Power i
Scram Pulse Mode not above 1100 Mw Manual Scram 1
Scram All Modes Pool Water Level 1
Scram All Modes not less than 14 feet of water above core Safety channel switched to "zero" or " calibrate" 2
Scram All Modes Timer after pulsing operation 1
Scram Pulse Mode 15 seconds or less Power Failure 1
Scram All Modes loss of console power
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I Minimum Operating Mode Number in which Interlock Operable Function Required 1
Startup Count Rate 1
Prevent control Reactor Startup Interlock rod withdrawal when neutron count rate is less than i
1 per second l
Transient Rod Interlock i
Prevent withdrawal Steady State of a transient rod Mode when its shock ab-so:ber anvil is not fully inserted Simultaneous Control Rod 1
Prevent simultaneous All Modes Withdrawal Prohibit manual withdrawal of Interlock two control rods Reactor Power Level 1
Prevent transient Pulse Mode Interlock rod withdrawal when power is greater than 10 kw Basis I
The power level scrams are provided in n!! modes of operation as protection.against abnormally high fuel temperatures and to assure that the reactor operation stays within the licensed limits. The manual scram n!!ows the operator to shut down the system if an unsafe or abnormal condition occurs. The reactor period alarm alerts the operator to potential rapid transient power changes so limiting actions may be taken. The pool water level scram assures that sufficient water shielding is above the core during reactor operation.
The interlocks which prevent the withdrawal of the transient rod in the steady state mode and when the power level is greater than 10 kw prevent inadvertent pulses. The interlock to l
prevent startup of the reactor with less than one neutron per second indicated on the startup channel assures that sufficient neutrons are available to provide indication on the measuring channels.
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. 3.6. Ventilation System Applicability This specification applies to the operation of the reactor facility ventilation system.
Objective The objective is to assure that the ventilation system is in operation to mitigate the consequences of the possible release of radioactive materials resulting from reactor operation.
Specification The reactor shall not be operated unless the facility ventilation system is in operation with a minimum air withdrawal rate of 500 cfm except for periods of time not to exceed two days to permit repairs to the system. During such periods of repair:
a.
The reactor shall not be operated at power levels above 10 kw and; b.
The reactor shall not be operated with experiments in place whose failure could result in the release of radioactive gases or aerosols.
Basis It is shown in the Safety Analysis Report that operation of the ventilation system reduces doses in the reactor facility due to argon-41, and also in the event of a TRIGA fuel element failure. The specifications governing operation of the reactor while the ventilation system is undergoing repairs limit the generation of argon-41 and also reduce the probability of fuel element failure during such times.
. 3.7 Experiments Applicability These specifications apply to experiments installed in the reactor and its experimental facilities.
Objective The objective is to prevent damage to the reactor or excessive release of radioactive materials in the event of an experiment failure.
Specifications The reactor shall not be operated unless the following conditions exist:
a.
fueled experiments shall be limited such that the total inventory of iodine isotopes 131 through 135 in the experiment is not greater than 1.5 millicuries and the strontium-90 inventory is not greater than 5 millicuries; b.
experiments containing materials corrosive to reactor components, compounds highly reactive with water, potentiaUy explosive materials, or liquid fissionable materiais shall be doubly encapsulated; and c.
known explosive materials shall not be irradiated in the reactor in quantities greater than 25 milligrams. In addition, the pressure produced in the experiment container upon detonation of the explosive shall have been determined experimentally, or by calculations, to be less than the design pressure of the container.
Basis The limits of Specification 3.7a prevent the dose in unrestricted areas resulting from experiment failure from exceeding 10 CFR Part 20 limite. Calculations for the SAP.
demonstrate that the maximum release in the event of a fuel element failure would not exceed 6.5 millicuries of iodine isotopes 131 through 135. Specifications 3.7b and 3.7c are provided to reduce the probability of damage to reactor components resulting from experiment failure.
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. 4.0 SURVEILLANCE REQUIREMENTS 4.1
,lM Applicability l
This specification applies to the surveillance requirements for the fuel elements.
Objective The objective is to assure that the dimensions of the fuel elements remain within acceptable l
limits.
i Specifications a.
All fuel elements shall be removed from the core and visually inspected for evidence of deterioration of cladding, (including at least corrosion, erosion, wear, cracking, and weld integrity) at least once every five years, b.
The standard fuel elements shall be measured for length and bend at interws separated by not more than 500 pulses of magnitude greater than 52.00 of reactivity.
c.
A fuel element indicating an elongation greater than 1/4 inch over its original length or a lateral bending greater than 1/16 inch over its original bending shall be considered to be damaged and shall not be used in the core for further operation.
d.
Fuel elements in the B-and C-rings shall be measured for possible distortion in the event that thee is indication that the Limiting Safety System Settings may have been exceeded.
Basis The most severe stresses induced in the fuel elements result from pulse operation with high reactivity input, during which differential expansion between the fuel and the cladding occurs and the pressure of the gases within the elements increases sharply. The above limits on the allowable distortion of a fuel element correspond to strains that are considerably lower than the strain expected to cause rupture of a fuel element.
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"w ' 4.2,. Control Rods Applicability This specification applies to the surveillance requirements for the control rods.
Objective The objective is.to assure the operability of the control rods.
- Specification -
a.
The reactivity worth of each control rod sha!! be datermined annually.
b.
Control rod drop times shall be determined annually and after disassembly and reassembly of control rod drives or removal of control elements.
c.
The control rods shall be visually inspected for deterioration biennially.
- d.
On each day that pulse mode operation of the reactor is planned, a functional performance check of the transient (pulse) rod system shall be performed prior to pulse mode operation.
e.
Semiannually, the transient (pulse) rod drive cylinder and the associated air supply -
system shall be inspected, cleaned and lubricated as necessary.
f.
The maximum control rod reactivity insertion rates shall be determined annually.-
4 Basis The reactivity worth of the control rods is measured to assure that the required shutdown margin is available and.to provide a means for determining the reactivity worths of-experiments inserted in tM core. The visual inspection of the control rods and measurement of their drop times are i&de to determine whether the control rnds are capable of performing their functions properly.
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. 4.3 Reactor Safety System Applicability The specification applies to the surveillance requirements for the measudng channels of the reactor safety system.
Objective The objective is to assure that the safety system will remain operable and will prevent the fuct temperature safety limit from being exceeded.
Specification a.
A channel test of each of the reactor safety system channels required in the operating mode to be followed shall be performed prior to each day's operation or prior to each operation extending more than one day, b.
A channel check of the power level measuring channels required in the operating mode to be followed shall be performed daily whenever the reactor is in operation.
c.
A channel calibration by the calorimetric method shall be performed for the reactor power level measuring channels annually.
Basis The daily tests and channel checks will assure that the safety channels are operable. The annual calibration and verifications will permit any long-term drift of the channels to be corrected.
. 4.4 Radiation Monitoring Equipment l
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Applicability l
- n This specification applies to the radiation monitoring equipment required by Section 3.4 of these specifications.
l Objective l
i.
The objective is to assure that the radiation monitoring equipment is operating and to verify the :ppropriate alarm settings.
Specification a.
The alarm set points for the radiation monitoring instrumentation shall be verified prict to each day's run.
b.
The radiation monitoring equipment shall be calibrated annually.
Basis Verification of the alarm set points of radiation monitoring instrumentation will assure that J
sufficient information to provide protection against radiation exposure is available.
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.. s 4.5 Maintenance Applicability This specification applies to the surveillance requirements following maintenance of a control or safety system.
Objective The objective is to assure that a system is operable before being used after maintenance has been performed.
Specification a) Following maintenance or modification of a control or safety system or component, it shall be verif ed that the system is operable prior to its return to service. A system shall not be i
considered operable until after it is successfully tested.
b) Any additions, modifications, or maintenance to the ventilation system, the core and its associated support structure, the pool or its penetrations, the pool coolant system, the rod drive mechanism, or the reactor safety system shall be made and tested in accordance with the specifications to which the systems were originally designed and fabricated or to specifications approved by the Reactor Committee.
c) A licensed reactor operator shall be present during maintenance of the reactor control and safety system.
Basis This specification relates to changes in reactor systems which could directly affect the safety of the reactor. Changes or replacements to these systems which meet the original design specifications are considered to meet the presently accepted operating criteria.
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>; 4.6 Pool Water Conductivity Applicability This specific action applies to surveillance of pool water conductivity.
Objective
.The objective is to assure that pool water mineral content is maintained at an acceptable level.
Specification The conductivity of bulk coolant water shall be verified to be within specified limits at least monthly.
Basis Based on experience, in which pool water conductivity changes slowly with time, observation at these intervals provides acceptable surveillance of conductivity to assure that accelerated fuel clad corrosion does not occur.
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l i i 5.0 DESIGN FEATURES 5.1 Reactor Fuel Applicability 1
This specification applies to the fuel elements used in the reactor core.
Objective The objective is to assure that the fuel elements are of such a desig1 and fabricated in such a manner as.to permit their use with a high degree of reliability with respect to their i
mechanical integrity.
l Specificationss a.
Standard Fuel Element: The standard fuel element shall be of the TRIGA type and shall contain uranium-zirconium hydride, clad in 0.020 inch of 304 stainless steel. It shall contain a maximum of 9.0 weight percent uranium which has a maximum enrichment less than 20 percent. There shall be 1.55 to 1.80 hydrogen atoms to 1.0 zirconium atom.
b.
Loading: The elements shall be placed in a closely packed array except for exmrimental facilities or for positions occupied by control rods, elements fully loaded with graphite, a neutron startup source, or single positions within the array filled with water.
Basis This type of fuel element has a long history of successful use in TRIGA reactors.
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l-5.2 Reactor Building Applicability This specification applies to the facility which houses the reactor.
OJective.
The objective is to r,.ssure that provisions are made to restrict the radioactivity released into the environment.
Specifications a.
' The reactor shall be housed in a closed room of a facility designed to restrict leakage.
b.
The free volume of the reactor room shall be at least 6,000 cubic feet.
c.
All air or other gases exhausted from the reactor room during reactor operation shall be released at a minimum of 12 feet above ground level.
j d.
The reactor facility shall be equipped with a ventilation system capable of exhausting air or other gases from the reactor room from a stack at a minimum of 50 feet above ground level under emergency conditions.
Basis In order that the movement of air can be controlled, the facility contains no windows that can be opened. Under emergency conditions the room air is exhausted through a filter and discharged through a stack at a minimum of 50 feet above ground to provide dilution.
, 5.3 Fuel Storage Applicability This specification applies to the storage of reactor fuel at times when it is not in the reactor core.
Objective The objective is to assure that fuel which is being stored will not become supercritical and i
will not reach unsafe temperatur%.
Specifications a.
All fuel elements shall be stored in a geometrical array where the value of k rr is less e
than 0.9 for all conditions of moderation and reflection using light water, b.
Irradiated fuel elements and fueled devices shall be stored in an array which will permit sufficient natural convection cooling by water or air such that the fuel element or fueled device temperature will not exceed 800oC.
Basis Specification 5.3a assures that unplanned criticality will not occur in fuel storage racks.
Specification 5.3b is based on a fuel temperature limit of 950oC to assure fuel clad integrity when the clad temperatures can equal the feel temperature (Simnad, G. A. Report E-Il7-833, February 1980, p.4-1) i i
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,, 6.0 ADMINISTRATIVE CONTROLS 6.1 Organization a.
The reactor fr..cility shall be an integral part of the Nuclear and Energy Engineering Department of the College of Engineering and Mines at the University of Arizona as shown in the diagram below.
b.
The reactor facility shall be under the supervision of a licensed senior operator for the reactor. He shall be responsible for assuring that all operations.re conducted in a safe L.
manner and within the limits prescribed by applicable federal regulations, by the l
facility license, and by the provisions of the Reactor Committee.
c.
There shall be a Health Physicist responsible for assuring the safety of reactor operations from the standpoint of radiation protection, i
d.
An NRC-licensed operator must be present in the control room when the key switch is on. An operator and one other person authorized by the Reactor Supervisor must be present in the Reactor Laboratory whenever the reactor is not shut down.
COLLEGE of ENGINEERING andMINES RADIATION CONTROL DEAN DIRECTOR HEAD of NUCLEAR and ENERGY ENGINEERING -------t RADIATION DEPARTMENT CONTROL OFFICE l
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REACTOR U
T REACTOR
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REACTOR L~
HEALTH D
TO SUPERVISOR C0MMITTEE PHYSICIST REACTOR OPERATIONS l
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6.2 L ' Review-a.
There shall be a Reactor Committee which shall review reactor operations to assure that ~
the facility is operated in a manner consistent with public safety and within the terms.
of the facility license.
b.
The responsibility of the Committee includes, but is not limited to, the following-1.
' Review and approval of experiments utilizing the reactor facilities; 2.-
Review and approval of all proposed changes to the facility, procedures, and Technical Specifications;
' 3.
. Determination of whether a proposed change, test, or experiment would constitute.
an unreviewed safety question or a change in the Technica' Specifications as i
required by 10 CFR 50.59, and review and approval of required safety analyses; 1
'4.
Review of the operation and operational records of the facility;
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Review of abnormal performance of plant equipment end operating anomalies; j
and j
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Review of unusual or abnormal occurrences and incidents which are reportable under 10 CFR 20 and 10 CFR 50.
3 7.
Review and audit of the retraining and requalification program for the operating
-staff.
8.
Biennial audit of the Emergency Plan, c.
The Committee shall be composed of at least five members, and shall include a health physicist and members competent in the field of reactor operations, radiation science, or l
reactor engineering. The membership of the Committee shall be such as to maintain a -
high level degree of technical proficiency.
J d.
The Committee shall establish a written charter defining such matters as the authority I
of the Committee, review and audit functions, and other such administrative provisions as are required for effective functioning of the Committee. Minutes of all meetings of l
the Committee shall be kept and submitted to committee members and to the Head of the Department of Nuclear and Energy Engineering in a timely manner.
l e.
A quorum of the Committee shall consist of not less than three members of the L
Committee and shall include the chairman or his designee.
f.
The Committee shall meet at least quarterly.
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.t 6.3.1 Operating Procedures Written procedures, reviewed and approved by the Reactor Committee, shall be in effect and followed for the following items. The procedures shall be adequate to assure the safety of the reactor, but should not preclude the use of independent judgment and action should the situation require such.
a.
Startup, operation, and shutdown of the reactor.
b.
Installation or removal of fuel elements, control rods, experiments, and experimental facilities.
c.
Actions to be taken to correct specific and foreseen potential malfunctions of sysims or components, including responses to alarms, suspected primary coolant system leaks, and abnormal reactivity changes.
d.
Emergency conditions involving potential or actual release of radioactivity, including provisions for evacuation, re-entry, recovery, and medical support.
c.
Maintenance procedures which could have an effect on reactor safety.
f.
Periodic survel' lance of reactor instrumentation and safety system, area monitors and continuous air monitors.
Substantive changes to the above procedures shall be made only with the approval of the Reactor Committee. Temporary changes to the procedures that do not change their original intent may be made with the approval of the Reactor Laboratory Director. All such temporary changes to procedures shall be documented and subaquently reviewed by the Reactor Committee.
6.3.2 ALARA Program A program shall be established to assure that radiation exposures and releases are kept as low as reasonably achievable.
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1 6.4 Action to be Taken in the Event a Safety Limit is Exceeded In the event a safety limit is exceeded, or thought to have been exceeded:
l a.
The reactor shall be shut down and reactor operation shall not be resumed imtil authorized by the NRC.
b.
An immediate repor'. of the occurrence shall be made to the Chairman of the Reactor Committee and reports shall be made to the NRC in accordance with Section 6.7 of these specifications.
c.
A report shall be made which shall include an analysis of the causes and extent of possible resultant damage, efficiency of corrective action, and recommendations for measures to prevent or reduce the probability of recurrence. This report shall be submitted to the Reactor Committee for review, and a similar report submitted to the t.
NRC when authorization to resume operation of the reactor is requested.
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6.5 Action to be Taken in the Event of a Reportable Occurrence In the event of a Reportable Occurrence, the following action shall be taken:
a.
The Reactor Laboratory Director shall be notified and corrective action taken prior to resumption of the operation involved.
b.
A report shall be made which shall include an analysis of the cause of the occurrence, efficiency.of corrective action and reconimendations for measures to prevent or reduce the probability of reoccurrence. This report shall be submitted to the Reactor Committee for review, c.
A t; port shall be subinitted to the NRC in accordance with Section 6.7 of these specifications.
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/ 1 6.6 Plant Operating Records
' In addition to the requirements of applicable regulations, and in no way substituting therefor, records and logs of the following items shall be prepared and retained for a period of at least.
5 years (except as otherwise specified in the Commitsion's regulations);
)
a.
Normal plant operation (but not including supporting documents such as checklists, and recorder charts, which shall be maintained for a period of at least one year);
- -incipal maintenance activities; Reportable Occurrences; d.
Equipment and component surveillance activities required by the Technical Specification; e.
Experiments performed with the reactor; Loss and records of the following items shall be prepared and retained for the life of the facility.
f.
. Gaseous and liquid radioactive effluents released to the environs; g.
Off-site environmental monitoring surveys; h.
Fuel inventories and transfers; i.
Facility radiation and contamination surveys; j.
Radiation exposures for all personnel; and k.
Updated, corrected, and as-built drawings of the facility.
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6.7 L Reporting ' Requirements -
In addition to the requirements of applicable regulations, and in no way substituting therefor, reports shall be made to the NRC as follows:
a.
A report within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and telegraph to the USNRC Region V Office of Inspection and Enforcement of:-
- 1. -
Any accidental offsite release of radioactivity.above limits permitted by.10 CFR '
20, whether or not the release resulted in' property damage, personal injury, or exposure;
. 2.
Any violation of a Safety Limit; and 3.
Any reportable occurrences as defined in Section 1.0 (Reportable Occurrence) of' these specifications in writing.
- b. ~
A written report within ten days to the U. S. Nuclear Regulatory Commission A+tn:
' Document Control Desk,. Washington D.C. 20555, with a copy to the Director, Division of Reactor Safety and Projects, Region V, of :
1.-
- Any significant variation of measured values from a corresponding predicted value of previously measured value of safety-connected operating characteristics occurring during' operation'of the reactor; 2.
Incidents or conditions relating to operation of the facility which prevented or
. could have prevented the performance of engineered safety features as described in these specifications; 3.
Any reportable occurrences as defined in Section 1.0 of these specifications; and '
4.
Any violation of a Safety Limit.
I 5.
Any accidental offsite release of radioactivity above limits permitted by 10 CFR 20, whether or not the release resulted in property damage, personal injury, or
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exposure.
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l A written report within 30 days to the U.S. Nuclear Regulatory Commission, Attn:
c.
Document Control Desk, Washington D.C. 20555, with a copy to the Director, Division 1
of Reactor Safety and Pmjects, Region V, of:
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Any substantial variance from performance specifications contained in these specifications or in the Safety Analysis Report; l
2.
Any significant change in the transient or accident analysis as described in the Safety Analysis Report; 3.
Any changes in facility organization; and t
4.
Any observed inadequacies in the implementation of administrr.tive or procedural controls.
d.
A written report within 60 days after completion of startup testing of the reactor to the U. S. Nuclear Regulatory Commission, Attn: Document Control Desk, Washington D.C.
20555, with a copy to the Director, Division of Reactor Safety and Projects, Region V.
1.
An evaluation of facility performance to date in comparison with design predictions and specifications; and 2.
A reassessment of the safety analysis submitted with the license application in light of measured operu..ag characteristics when such measurements indicate that there may be substantial variance from prior analysis, e.
A written annual report within 60 days following the 30th of June each year to the U.S.
Nuclear Regulatory Commission, Attn: Document Control Desk, Washington D. C.
20555, with a copy to the Director, Division of Reactor Safety and Project *,, Region V.
1.
A brief narrative summary of (1) operating experience (including experiments performed), (2) change.s in facility design, performance characteristics, and operating procedures related to reactor safety and occurring during the reporting period, and (3) results of surveillance tests and inspections; 2.
Tabulation of the energy output (in megawatt days) of the reactor, amount of pulse operation, hours reactor was critical, and the cumulative total energy output since initial criticality; 3.
The numoer of emergency shutdowns and inadvertent scrams, including reasons therefort; 4.
Discussion of the major maintenance operations performed during the period, including the effect, if any, on the safety of the operation of the reactor, and the reasons for any corrective maintenance required; 1
5.
.A brief description including a summary of the safety evaluations of changes in the facility or in procedures and of tests and experiments carried out pursuant to j
Section 50.59 of 10 CFR Part 50; l
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A' summary of the nature and amount of radioactive effluents released or discharged to the environs beyond the effective control of the licensee as L
- measured at or' prior to the point of such release or discharge; Liquid Waste (summarired on a monthly basis) n.
Radioactivity discharged during the reporting period.
(1)
Total radioactivity released (in curies).
(2)
The MPC used and the isotopic composition if greater than 1 x 10-7 microcuries/cc for fission and activation products.
(3)
Total. radioactivity (in curies), released by nuclide, during the reporting period, based on representative isotopic analysis.
(4)
Average concentration at point of release (in microcuries/cc) during the reporting period.
b.
Total volume (in gallons) of effluent water (including diluent) during periods of release.
Gaseous Waste (summarized on a monthly basis) a.
Radioactivity discharged during the reporting period (in curies) for:
(1)
Gases.
(2)
Particulate with half lives greater than eight days.
b.
The MPC used and the estimated activity (in curies dacharged during the reporting period, by nuclide, for all gases and particulate based on representative isotopic analysis.
Solid Waste a.
The total amount of solid waste packeged (in cubic feet).
b.
The total activity involved (in curies).
c.
The dates of transfer or shipment and dispsition 7.
A summary of radiation exposures received by facility personnel and visitors, including dates and time of significant exposures, and a summary of the results of radiation and contamination surveys performed within the facility; and 8.
A description of any environmental surveys performed outside the facility.
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All proposed new experiments utilizing the reactor shall be evaluated by the experimenter and the Reactor Committee. The evnination shall be reviewed by a licensed Senior Operator of the facility (and the Health Physicist when appropriate) to assure compliance with ths provisions of the utilization license, the Technical Specifications,10 CFR 20, and the requirements of 10 CFR 50.59. If, in his judgment, the experiment meets with the above provisions and does not constitute a threat to the integrity of the reactor, he shall submit it to the Reactor Supervisor for scheduling or to the Reactor Committee for final approval as indicated in Section 6.2 above. When pertinent, the evaluation shall include:
1.
The reactivity worth of the experiment; 2.
The integrity of the experiment, including the effects of changes in temperature, pressure, or chemical composition; 3.
Any physical or chemical interaction that could occur with the reactor components; and 4.
Any radiation hazard that may result from the activation of materials or from external beams.
5.
A determination that for the maximum planned or inadvertent pulse, no credible mechanism exists which could cause the experiment to fail, b.
Prior to performing an experiment not previously performed in the reactor, it shall be reviewed and approved in writing by the Reactor Committee. This review shall consider the following information:
i 1.
The purpose of the experiment;
)
2.
A procedure for the performance of the experiment; and j
1 3.
The evaluation approved by a licensed Senior Operator.
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For the irradiation of materials, the applicant shall submit an "Irrad.iation Request" to the Reactor Supervisor. This request shall contain information on the target material including the amount, chemical form, and packaging. For routine irradiations (which do not contain known explosive materials and which do not constitute a significant threat to the integrity of the reactor or to the safety of individuals) the approval for the Reactor Committee may be made by the Reactor Supervisor.
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d.
In evaluating experiments, the following assumptions should be used for the purpose of determining whether failure of the experiment would cause the appropriate limits of 10 CFR 20 to be exceeded:
1.
If the possibility exists that airborne concentrations of radioactive gases or aerosols may be released within the facility,100 percent of the gases or aerosols will escape; 2.
If the ef 'luent exhausts through a filter installation designed for greater than 99 percent efficiency for 0.3 micron particles,10% of the particulate will escape; and 3.
For a material whose boiling point is above 130oF and where vapors formed by boiling this material could escape only through a column of water above the core, 10% of these vapors will escape.
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