ML20034E864
| ML20034E864 | |
| Person / Time | |
|---|---|
| Site: | General Atomics |
| Issue date: | 12/31/1992 |
| From: | GENERAL ATOMICS (FORMERLY GA TECHNOLOGIES, INC./GENER |
| To: | |
| Shared Package | |
| ML20034E861 | List: |
| References | |
| NUDOCS 9303020029 | |
| Download: ML20034E864 (24) | |
Text
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TRIGA REACTORS FACILITY TRIGA Mark I Reactor ANh1JAL REPORT for CALENDAR YEAR 1992 prepared to satisfy the requirements of U.S. Nuclear Regulatory Commission Facility License R-38 Docket No. 50-89 February 1993
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TRIGA REACTORS FACILITY TRIGA Mark I Reactor ANNUAL REPORT for 1992 l
TABLE OF CONTENTS Section Page I NTR O D U CTI ON............................................ 1
?.
SUMMARY
OF OPERATIONS............................. 2 1.1 Operating Experience................................. 2 1.2 Facility Changes and Modifications 5
1.3 Surveillance Tests and Inspections.......................
5 2.
ENERG Y G ENERATION................................. 9 3.
EMERGENCY SHUTDOWNS AND INADVERTENT SCRAMS.......
11 4.
M AINTENANCE ACTIVITIES............................
12 4.1 Reactor, Mechanical, and Auxiliary Systems..................
12 4.2 Instrumentation and Control System.......................
13 5.
10CFR50.59 FACILITY MODIFICATIONS AND SPECIAL EXPERIM ENTS......................................
16 5.1 Use of New Large Diameter In-Core Irradiation Tubes in the R-38 a
Facility
.............................16 5.2 Modification of the R-38 Facility Console Digital Communication Network Configuration.....................................
17 6.
RADIOACTIVE EFFLUENTS DISCHARGED TO THE ENVIRONMENT.
20 7.
ENVIRONMENTAL MONITORING.........................
21 8.
SUMMARY
OF RADIATION EXPOSURES AND RADIOLOGICAL S U R VEY S......................................... 2 2 8.1 TRIGA Reactors Facility Staff Whole Boo, Exposures...........
22 8.2 Nonfacility GA Staff Whole Body Exposures.................
22 8.3 Contractor and Reactor Users Whole Body Exposures............
22 8.4 Visitor Whole Body Exposures..........................
22 8.5 Routine Wipe Surveys of Mark I Reactor Facility...............
23 8.6 Routine Radiation Measurements of Mark I Reactor Facility........
23
INTRODUCTION This report documents operation of the General Atomics (GA) TRIGA Afark I non-power reactor for the period January 1 - December 31, 1992. The hiark I reactor - one of two reactors operated by GA at its San Diego, California facilities - is a pulsing type reactor with a licensed steady state operating power of 250 kilowatts, and with maximum reactivity insertions during transient operations of $3.00. It is operated by GA under License No. R-38 granted by the U.S.
Nuclear Regulatory Commission (Docket No. 50-89). The second reactor is a 1.5 hiW(t)
TRIGA hiark F reactor operated under License No. R-67; both reactors are housed in GA's reactor building with their own independent reactor rooms and control rooms.
This report is being prepared and submitted to satisfy the requirements of Section 9.6(e) of the R-38 Technical Specifications, as amended. This report is presented in eight parts, consistent with the information required by the applicable Technical Specifications..
1.
SUMMARY
OF OPERATIONS 1.1 Operatine Emerience.
The TRIGA Mark I reactor was operated during calendar year 1992 on an as needed basis, for numerous steady-state irradiations as well as pulsing operations. The following represents a summary of reactor operations during this period:
1.1.1 The reactor generated a total of 24.16 Mwh of energy.
1.1.2 The reactor was pulsed 81 times, for a total of 12,142 pulses to date.
1.1.3 The reactor consumed 1.51 grams of U-235.
1.1.4 A total of 300 irradiation requests were processed during the period.
1.1.5 Two applications for facility modifications under 10CFR50.59 were approved and implemented during this reporting period.
These modifications are described in Section 5 of this report.
1.1.6 No special experiments, as defined in the R-38 Technical Specifications, were conducted during this period.
1.1.7 No amendments to the facility license were issued during thic period.
1.1.8 There was one occurrence during the year which was reported to the U.S.
Nuclear Regulatory Commission (NRC) under the requirements of 10CFR73.71. On December 12, 1992, an unauthorized entry into the TRIGA Mark I Category III protected area occurred, when a contract employee was able to manipulate the door lock and gain entry. Since the facility was occupied at that time, the reactor operators on duty immediately intercepted and escorted the individual from the facility. Corrective actions included termination of the contract employee, replacement of the door lock in question with a new lock that cannot be manipulated in the same manner, and an assessment of all perimeter door locks at the facility to ensure that they were installed and operating as prescribed. A future planned upgrade.
of facility security will include the installation of additional perimeter intrusion detection devices.
1.1.9 During a one week period in September 1992, an additional administrative control on pulsing operation was imposed. This was a result of a generic software problem which allowed a control rod to continue its upward motion if the reactor mode was switched from steady state to pulse with the rod up button depressed, then the up button released. Even though this was not a normal operational sequence, the additional administrative control required all operators to stop all rod motion before switching modes.
Upon installation of corrected software, the administrative control was lifted.
1.1.10 The facility conducted one reactor operator training program during this period for new operators. As a result, one candidate was granted an instant SRO license by the NRC. Seven licensed operators completed biennial written examinations under the facility requalification program requirements.
1.1.11 The following types of operations were conducted to support the various users of the reactor:
neutron activation analysis (2180 samples).
Iow power radiation hardness testing of electronic piece parts and hybrid circuits.
neutron radiography.
irradiation at ambient as well as high sample temperatures, using electrically heated (in-core and ex-core) furnaces, of gas cooled reactor fuel, including New Production Reactor (NPR) test fuel capsules.
testing of commercial reactor instrumentation.
operator training and requalification exercises.
Table I summarizes pertinent reactor operating parameters for 1992. ______ -
TABLE I SUM 51ARY OF TRIGA MARK I OPERATING DATA Annual Values Operating Parameter January 1,1992 through December 31,1992 KWh of energy produced 24,159 MWD of energy produced 1.01 Grams U-235 consumed 1 51 Number of fuel elements removed from corem 0
Number of fuel elements added to core
- 0 Number of pulses 81 Hours reactor critical (steady state) 264 Number of start-up and shutdown checks 272 Number of irradiation requests processed 300 Number of facility modifications under 10CFR50.59 2
Fuel elements removed from the core represents fuel removed as a result of bending or length changes, or determined to be damaged or otherwise deteriorated.
Fuel elements added to the core represents fuel added to compensate for loss of reactivity, or to replace fuel removed from the core due to damage or deterioration. _ _ - _ _
1.2 Facility Changes and Modifications.
There were no ma_ior changes to the Mark I reactor facility during 1992. Several changes to the reactor instrumentation and control system were made in an effort to enhance its operation and usefulness. Where appropriate, such changes were approved by the facility safety committee on the basis of an application and supporting safety evaluation submitted by the facility under the provisions of 10CFR50.59.
The modifications are described in Sections 4 and 5 of this report.
1.3 Surveillance Tests and Inspections.
Surveillance tests and inspections were performed as required by Sections 3.0 (Reactor Pool),4.0 (Reactor Core) and 5.0 (Control and Safety Systems) of the R-38 Technical Specifications. A summary of the results are presented below:
1.3.1 Pool Water.
The pool water was sampled on a continuous basis for conductivity using an on-line sensor installed in the water treatment system flow. Water conductivity was maintained well below the limit of 5 micro-mhos per centimeter averaged over one calendar month as required by the Technical Specifications.
Water level sensors with audible and visual alarms were used to ensure that the pool water level always was maintained at acceptable levels. In addition, a visual check of pool water level was made as part of the Daily Start-up Checklist.
Redundant pool water temperature monitors were used to ensure that bulk pool water temperature is maintained within acceptable limits.
1.3.2 Reactor Core. During the month of December 1992, the reactor fuel was inspected visually for damage and deterioration and all uninstrumented fuel elements were inspected for length and bend changes. The growth test measures the elements average growth (which must be less than 0.500-inch for aluminum-clad elements and less than 0.100-inch for stainless steel clad elements). The bend test measures the sagitta of each element over a length l
of 23-inches along the cladding; the bend must be less than 1/16-inch for the element to be considered satisfactory. It is to be noted here that many older aluminum clad elements are still in routine use in the Mark I reactor; however, all presently manufactured TRIGA fuel now uses stainless steel clads. The present reactor core is a mixture of aluminum and stainless steel clad fuel elements.
There were no fuel elements removed from routine use due to observed clad deformities as a result of this fuel inspection. The inspection utilized an enhanced inspection device utilizing an underwater color CCD camera based system - put in use during calendar year 1991 - as opposed to an observation utilizing the older telescope-mirror system used in previous years. The new inspection device also gives the flexibility of recording a video image of the fuel for future reference.
1.3.3 Control RodL All control rods were removed from the core and visually inspected for deterioration during December 1992, and all rods were found to be in satisfactory condition. The next scheduled inspection of control rods is December 1994.
As part of the routine control rod surveillance procedures, the mechanical components of the central transient (pulse) rod (air piston, lip seal, anvil and accumulator) were inspected, cleaned, and lubricated twice during the calendar year as part of the routine surveillance activities (June and December 1992). No deterioration or undue wear were noted on the rod damper assembly itself, which had been completely overhauled in 1987.
1.3.4 Reactor Safety Systems. Surveillance and calibration of reactor safety systems was carried out as specified in the R-38 Technical Specifications and reactor operating procedures. The calibrations and checks on the scram functions of the minimum required safety system scrams were verified on a routine basis, with the surveillance on power level, fuel temperature measuring channels and manual scram capability performed on a daily basis prior to reactor start-up, to ensure that the channels are operating as intended, and that the set points for these channels are within the limits specified in the Technical Specifications. __
A calorimetric determination of reactor power was performed monthly to verify the calibration of the three power measuring channels.
In conformance with reactor operating procedures, the calibration of the power measuring channels was considered acceptable if the deviation of the measured value from the indicated power was less than five percent; the power measuring channels were adjusted to conform to the calorimetric value if the deviation was greater than five percent. During the reporting period, thirteen such adjustments to the power level channels were made.
However, it is noted that only on four of these adjustments was the deviation actually greater than five percent. For the other eight, where the deviation was less than 5%, the adjustments were made regardless because micrometer detector adjustment devices installed in 1988 now give greater precision and sensitivity in setting detector positions, which allows a more accurate indication of power level to be obtained.
1.3.5 Radiation Monitoring.
The primary instruments utilized during the reporting period for facility radiation monitoring were a continuous beta-gamma air monitor, radiation area monitors, water and air filter monitors, a control console monitor, and a variety of portable survey meters. Their use and calibration is described below:
Continuous Air Monitor (CAM). The CAM (Ludlum Model 333-4) alert and alarm set points were checked on a weekly basis by activating them with a check source. Calibration of the system was performed annually using two Sr-90/Y-90 sources with a calibration traceable to the National Institute of Standards and Technology (NIST). Two sources were used to allow calibration at low and high count rates.
Radiation Area Monitors (RAM). Two area monitors (Eberline Instrument Corp.) were used for monitoring area radiation levels in the reactor room.
The low level monitor was used to provide an alarm when the area radiation levels exceeded 20 mR/h; the high level monitor alarmed at levels exceeding 5000 mR/h. The alarm set points were checked daily, with alarm testing performed biweekly using a check source. Calibration was performed annually using a 4 mci Cs-137 source on a calibration range.
All calibrations were traceable to NIST...
Water and Air Radiation Monitors. Separate radiation monitors (Eberline RMS II) were used to monitor the radiation levels in the reactor pool water and the reactor room air ventilation system. Their operation and alarm set points (50 mR/h and 5 mR/h respectively) were checked daily with alarm testing performed on a weekly basis. The monitors were calibrated on an annual basis using the calibration range; all calibrations were traceable to NIST.
Console Radiation Monitor.
A radiation monitor (Eberline RMS II) monitors dose rates at the reactor console. The alarm set point (2.5 mR/h) was checked daily with alarm testing performed on a weekly basis. The monitor was calibrated on an annual basis.
Portable Radiation Monitors. Several types of portable radiation monitors were in use at the facility. Examples are the Eberline RO2 and RO2A beta-gamma survey meters, the Ludlum pancake probes, the Ludlum MicroR meter and the LFE SNOOPY neutron survey meter. All portable radiation monitors were calibrated on a semiannual basis, with the exception of the SNOOPY neutron survey meters, which were cahorated on an annual basis.
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ENERGY GENERATION i
t The total energy generated during calendar year 1992 as a result of Mark I operations was j
24,159 kilowatt-hours. Figure 1 is a bargraph showing energy generated on a monthly I
basis during the year.
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i 4500.0
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4000.0 i
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3000.0-
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F 2500.0-r o
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JAN FEB MAR APRIL MAY JUNE JULY AUG SEPT OCT NOV DEC 1992 i
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Figure 1. TRIGA Mark i Energy Production for 1992 i
3.
DIERGENCY SIIUTDOWNS AND INADVERTENT SCRAMS i
The total number of unscheduled scrams during 1992 operations was 18. None of the unscheduled scram experienced in 1992 had any efect on, or consequencefor, the safe operation of the Mark I reactor. In fact, all safety channels functioned as intended in shutting down the reactor when trip setpoints were reached, or an error condition was otherwise detected in the reactor operating systems. The causes of the scraras are grouped into the following general categories:
Scram Channel Cause Number Power level Operator er or, including training 10 Timer Operator error; accidentally activating scram timer 1
L Power level Noisy signal from one of three power channels 1
Power level Instrument being reset by operator. Noisy signal I
was causing intermittent high alarms Watchdog Timer Watchdog timeouts in system computers 3
l External Low pressure disconnect improperly secured on 1
King Furnace External I_oss of voltage to King Furnace Power Supply 1
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i 4.
MAINTENANCE ACTIVITIES All maintenance activities performed during the year generally fall into three categories-i l
(i) routine preventative maintenance, (ii) routine calibration activities, and (iii) ongoing
)
upgrade activities associated with replacement of older components and systems with state-of-the-art technology, or simply due to wear and tear from the many years of use.
i Significant activities in this area are described below:
i 4.1 Reactor. Mechanical. and Auxiliary Systems January 1992 The "O" ring air seal on the central transient rod (CTR) mechanism was replaced to correct an air leak problem in the CTR cylinder. At the same time, the piston shaft connector on the CTR was machined and refinished to allow for a proper union between the shaft and the actuator using a detent screw. Prior to this modification, the mating l
was against the round piston shaft rather than a machined flat in the shaft.
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February 1992 A leak in the city water supply pipes, supplying one of the cooling I
towers, was repaired by installing a new section of pipe in the affected area.
March 1992 The down limit switch actuator on the CTR cylinder was replaced.
This replacement affected only the portion of the dual switch for indication of cylinder down and not the interlock portion of the limit switch.
April 1992 A minor leak in the automatic bleed valve pipe in the water treatment system was located and repaired.
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Broken teeth on the main drive gear of the rotary specimen rack (RSR) were repaired. At the same time, the access port in the sample cavity used to detect moisture in the RSR cavities was relocated from position
- 25 to # l.
i June 1992 Semiannual maintenance was performed on the CTR drive. The air l i
system was inspected and no leaks were found. All control rods were calibrated as part of the maian1ual calibration and maintenance activities.
December 1992 Semiannual maintenance on the CTR drive, as well as biannual control rod inspection and maintenance was performed. The CTR air supply system and shock absorber were inspected and found to be satisfactory.
All four control rods were removed during the annual fuel inspection, visually checked for wear and tear, and found to be satisfactory.
A persistent air leak in the diffuser system was traced to the stem of a PVC ball valve in the discharge line, allowing air to leak in by venturi action during water flow. The valve and associated piping will be replaced during 1993.
4.2 Instrumentation and Control System January 1992 The "high" communications network boards were removed from the instrumentation and control (I&C) system for continued testing of network reliability issues when operating with dual networks. Two redundant and independent networks, labeled "high" and " low" networks are utilized for communication between the console and data acquisition computers.
May 1992 Both "high" and " low" networks were returned to service after tracking system reliability with a single network.
Memory chips in the data acquisition computer were replaced in an effort to fix frequently encountered " memory parity error" problems.
June 1992 An independent rod drop timer (RDT) was designed and installed as part of the I&C system. The RDT is utilized for a verification of control rod drop times from its fully withdrawn position to full insertion. Use of the RDT replaces the manual measurement of rod drop times using a stop watch. It automatically records the period between the time that the individual rod scram switch is depressed, to the time that the rod reaches bottom and the limit switch is tripped. t
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The " acknowledge" switch on the console was replaced. The old i
switch mechanism was found to have a broken spring apparently due I
to the high frequency of repetitive action encountered with this switch.
July 1992 Power, fuel temperature and auxiliary channels were calibrated as part of the semiannual maintenance and calibration activities.
The " action pack" for fuel temperature channel no. 3 (FT3) was replaced with a " Mighty Module" Hi/ Low trip type. This allows both high (-495C) and low (5C) channel trip set points. The action pack units on FTl and FT2 were similarly replaced during 1991.
Switch sections on all mode control panel switches (Prestart, Pulse, Manual, Automatic) were all replaced to eliminate occasional problems in operation of some of the switches.
September 1992 Modified software was installed which will halt rod movement if a standard rod is moving up upon switching reactor modes from steady state to pulse. All modes of operation tested satisfactory with this revision to the software.
I October 1992 Operation of the Mark I was permanently switched to a single network configuration (see section 5.2). The "high" network circuit boards were removed from the two system computers and the associated network communications fault warning was removed from the warning message window.
l A faulty relay in one of the two external scrams in the scram loop was replaced. External scrams inputs are used for reactor experiments and other testing requiring automatic reactor scram capabilities.
November 1992 A broken potentiometer on the CTR drive was replaced to restore CTR position indication at the console.
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December 1992 The control rod drive designations were changed in the I&C system to make the highest worth control rod the regulating rod. This change was made for finer control of reactor power when operating in the automatic, steady-state mode.
Power, fuel temperature and auxiliary channels were calibrated as part of the semiannual maintemnce and calibration activi ies.
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i 5.
10CFR50.59 FACILITY MODIFICATIONS AND SPECIAL EXPERIMENTS f
i Two (2) applications for facility modifications under the provisions of 10CFR50.59 were approved and implemented for the R-38 facility during the 1992 reporting period. In the implementation of these facility modifications, the applications for the proposed changes were reviewed by the TRIGA Reactors Facility Safety Committee, among others, and approved only after making the determination that the proposals for the changes did not; 1
i (a) involve a change to the R-38 Technical Specifications, or (b) create any unreviewed safety questions as defined in 10CFR50.59.
No new Special Experiments were submitted for the R-38 facility during 1992.
5.1 Use of New Large Diameterin-Core Irradiation Tubes in the R-38 Facili.ly 5.1.1 Description. New,2.5-inch O.D. cadmium-lined and non Cd-lined in-core dry tubes were proposed for use in the R-38 facility to allow for the conduct of tests on larger sized samples. These tubes are identical on the outside, with the only difference being in the cadmium lining that extends ~43 inches from the bottom of the tube along with a second inner tube to prevent piece part contact with the cadmium. To accommodate this need, a new top gnd plate with three (3), 3-fuel element and one (1) 4-element cutout j
locations was installed in 1990.
l The tubes are built with an angle of ~0.5* between the upper and lower sections to prevent gamma ray streaming.
Two (2) locator pins on the bottom of the tube fit into the holes in the bottom grid plate providing consistent positioning, as well as lateral support for the bottom end of the tube. The dry tubes are secured to the bridge by means of a clamp which is securely fastened in place on the bridge with a nut and threaded stud permanently installed on the bridge.
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5.1.2 Safety Evaluation. The application for the proposed modifications was l
reviewed and it was concluded that this modification does not involve I !
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changes to the R-38 Technical Specifications, or mvolve any unreviewed safety questions.
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The primary concern with the installation of these new dry tubes was the possibility of inhibiting cooling flow around the cutouts, causing excessive heating, and thus clad damage, to the surrounding fuel elements. However, it was shown that the minimum water gap is 0.093 inches, much greater that j
the 0.02 gap required in the hottest region of the core for operation at 250 KW.
The second concern was with touching fuel elements if fuel is improperly reinstalled in the triad. However, it was shown that the removal of heat from touching fuel elements is such that it does not lead to deletenous effects with respect to clad damage from potential hot spots.
l The reactivity worth of the new cadmium-lined dry tube with respect to water is slightly less than that of the smaller dry tube presently used at the facility (-51.90 vs. -52.04). Therefore, the sudden inadvertent removal of this tube at power will have consequences less severe than those already considered for the small cadmium-lined tube currently being used.
Flooding of the new tubes was also considered and determined to have consequences less severe than for the existing tube.
5.2 Modification of the R-38 Facility Console Dicital Communication Network Conficuration.
5.2.1 Description. The R-38 Facility instrumentation and control system (ICS) was configured with two (2) networks between the control system computer (CSC) and the data acquisition computer (DAC). At any given time, only a single network is utilized for communications between the CSC and DAC.
The second redundant network was reserved as a backup to which communications would automatically be switched should the primary network fail.
Through four (4) years of operating experience with this system, it became clear that the redundant network offered little in the way of a backup, as.--
most network failures caused a lock-up of the ICS (in which case the ICS scrammed the reactor if it is operating). In fact, the failure of the unused redundant network was also likely to cause a lockup of the ICS and a system scram (if the reactor was operating). The only network failure which will allow the redundant network to become the primary network was if the physical wiring between the normal primary network cards in the CSC and DAC was disconnected or otherwise violated. This is not considered a credible event.
After four (4) years of operation, it has been determined that inclusion of the second network for CSC-DAC communications decreased the reliability of the digital ICS. This is due to problems with the IC-DOS operating environment which is utilized by the digital ICS software. The problem revolves around IC-DOS operating system's inability to handle " interrupted system calls", which occur frequently when operating with two (2) networks. Since the network drivers are an integral part of the IC-DOS kernel, it is not possible for GA to correct this problem in the applications software.
The network reliability was tested by GA product engineers through the use of a Network Hardware Diagnosis Program, and the results indicated that system performance was degraded rather then improved by using a redundant network configuration.
5.2.2 Safety Evaluation. The application for the proposed modifications was reviewed and it was concluded that this modification does not involve changes to the R-38 Technical Specifications, or involve unreviewed safety questions.
The primary issue with respect to the conduct of reactor operations without a backup network is the ability to cause a reactor scram should the single network fail to provide the necessary communications between the CSC and DAC. In other words, can reactor operations continue safely should the single network fail. The presence of two (2) additional scram circuits, timeout of either the CSC or DAC watchdog timers, or by the detection of loss of communications between the CSC and DAC computers, makes the single network configuration similar to the dual network configuration, in that a watchdog timeout will scram the reactor when a network failure occurs, regardless of the number of networks that are available for communication between the CSC and DAC.
In addition, the deletion of the redundant communications network in no way affected the hardwired safety system scrams since these do not depend on the communications network.
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6.
RADIOACTIVE EFFLUENTS DISCIIARGED TO TIIE ENVIRONMENT During the calendar year 1992,1.74 millicuries of Argon-41 were discharged from the Mark I reactor facility to the atmosphere.
All low level radioactive wastes were transferred to GA's Nuclear Waste Processing Facility - which operates under NRC license SNM696 and GA's California Radioactive j
Materials License - for disposal. All waste was measured at the facility for specific radionuclide activity using high resolution gamma-ray spectroscopy prior to the transfer.
Solid wastes were then repackaged as necessary and shipped to an authorized disposal facility by GA's waste processing facility. Liquid waste was first subjected to volume reduction by evaporation, and the residue waste was packaged for disposal as solid waste.
Trace quantities of liquid low level waste may also be released into the municipal sewer system, if such waste is found to be within the limits and criteria specified by applicable local, state and NRC regulations.
During calendar year 1992, GA's TRIGA Reactors Facility (R-38 and R-67 licenses) shipped 246 cu. ft. of compacted as well as noncompactable low level radioactive waste to an authorized disposal facility.
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7.
ENVIRONMENTAL MONITORING There were no significant changes in the GA Environmental Surveillance Program during 1992.
The environmental monitoring program for the TRIGA Reactors Facility included the following:
Five (5) emergency air samplers situated on the roof and around the reactor building.
Fifteen (15) environmental air samples adjacent to, and near the GA site in accordance with GA's SNM-6961:.ense.
Daily liquid effluent monitoring from GA's main pump house, for gross alpha and beta concentrations.
Annual soil, vegetation, and water sampling at sixteen (16) stations on the GA site, including stations around the GA reactor building.
External radiation monitoring of the reactor facilP.ies using four (4) area dosimeters (26 locations around the entire GA site), as well as radiation meter surveys conducted periodically.
Air samplers located in the reactor room to routinely sample room air for airborne radioactivity.
Additional radiation monitors as described in Section 1.3.5 of this report.
8.
SUMMARY
OF RADIATION EXPOSURES AND RADIOLOGICAL SURVEYS The following data summarizes personnel radiation exposures (rem) and radiological surveys of the facility during 1992.
8.1 TRIGA Reactors Facility Staff Whole Body Exposures
- Number of employees monitored:
21 High Exposure:
0.145 Low Exposure:
0.000 Average Exposure:
0.029 2 Nonfacility GA Staff Whole Body Exposures
- Number of employees monitored:
21 High Exposure:
0.090 Low Exposure:
0.000 Average Exposure:
0.004 8.3 Contractor and Reactor Users Whole Body Exposures
- Number of persons monitored:
83 High Exposure:
0.185 Low Exposure:
0.000 Average Exposure:
0.006 8.4 Visitor Whole Body Exposures
Number of persons monitored:
59 High Exposure:
0.185 Low Exposure:
0.000 Average Exposure:
0.004. _ _ _ _ _ _ _ _ _ _ _
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8.5 Routine Wine Surveys of Mark I Reactor Facility High Wipe:
174.7 Beta dpm/100 cm2 Low Wipe:
2.3 Beta dpm/100 cm2 Average Wipe s1 Beta dmp/100 cm2 8.6 Routine Radiation Measurements of Mark I Reactor Facility High Measurement:
22 mrem /hr @ l foot Low Measurement:
0.5 mrem /hr @ 1 foot Average Level:
<0.1 mrem /hr @ l foot m
Includes reactor operations staff facility support staff and experimenters assigned to work full-time or near full-time at the reactor facility.
Includes GA support staff and experimenters who were granted periodic access to the reactor facility for the performance of work.
Includes non-GA personnel who were granted periodic access to the facility for the performance of work.
Includes GA staff who routinely work in other GA radiation facilities, and who were granted visitor access to the reactor facility. Most if not all, of the radiation exposure received by the GA staff was from these other radiation facilities. _ _ _ _ _ - _.
+ CENERAL ATOMICS R O. Box 85608
- San Diego, CA
- 92186-9784 ~ (619) 455-3000
_ _ _ _ _