ML20034E873

Triga Mark F Reactor Annual Rept for CY92

ML20034E873
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Site:	General Atomics
Issue date:	12/31/1992
From:	GENERAL ATOMICS (FORMERLY GA TECHNOLOGIES, INC./GENER
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TRIGA REACTORS FACILITY TRIGA Mark F Reactor ANNUAL REPORT for CALENDAR YEAR 1992 prepared to satisfy the requirements of U.S. Nuclear Regulatory Commission Facility License R-67 Docket No. 50-163 February 1993

++ CENERAL ATOMICS

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l TRIGA REACTORS FACILITY TRIGA Mark F Reactor ANNUAL REPORT fo: 1992 l

TABLE OF CONTENTS Section Page INTRODUCTION I

1.

SUMMARY

OF OPERATIONS.

2 1.1 Operating Experience...............................

2 1.2 Facility Changes and Modifications 4

1.3 Surveillance Tests and Inspections 6

2.

ENEP.GY GENERATION 10 3.

EMERGENCY SHUTDOWNS AND INADVERTENT SCRAMS.

12 4.

MAINTENANCE ACTIVITIES 13 4.1 Reactor, Mechanical and Auxiliary Systems

..............13 4.2 Instrumentation and Control System...

..............14 5.

10CFR50.59 FACILITY MODIFICATIONS AND SPECIAL EXPERIMENTS.. 16 6.

RADIOACTIVE EFFLUENTS DISCHARGED TO THE ENVIRONMENT..

17 7.

ENVIRONMENTAL MONITORING 18 8.

SUMMARY

OF RADIATION EXPOSURES AND RADIOLOGICAL SURVEYS 19 8.1 TRIGA Reactors Facility Staff Whole Body Exposures............

19 8.2 Nonfacility GA Staff Whole Body Exposures..

19 8.3 Contractor and Reactor Users Whole Body Exposures 19 8.4 Visitor Whole Body Exposures 19 8.5 Routine Wipe Surveys of Mark F Reactor Facility..........

20 8.6 Routine Radiation Measurements of Mark F Reactor Facility 20

INTRODUCTION This report documents operation of the General Atomics (GA) TRIGA Mark F non-power reactor for the period January 1 - December 31, 1992. The Mark F 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 1500 kilowatts, and maximum reactivity insertions during transient operations of 55.50. It is operated by GA under License No.

R-67 granted by the U. S. Nuclear Regulatory Commission (Docket No. 50-163). The sec-ond reactor is a 250 kW(t) GA Mark I reactor operated under license No. R-38. Both reactors are housed in GA's reactor building with their own independent reactor rooms and control rooms.

This report is presented in eight parts, consistent with the information required by Section 9.6(e) of the R-38 (Mark I) Technical Specifications, as amended. The administrative requirements in the R-67 (Mark F) Technical Specifications, as amended, do not have annual reporting requirements. i

1.

SUMMARY

OF OPERATIONS.

l 1.1 Operatine Experience. The TRIGA Mark F reactor was operated during calendar year 1992 in the steady-state mode only, primarily for in-core irradiations of direct conversion (thermionic) devices. Irradiations of other types i

of samples were carried out as necessary. Operations were continuous, except for shutdowns - typically one to two weeks - required for annual reactor related inspection and maintenance activities, neutron radiography inspection of the f

thermionic experimental devices and extended holiday periods. The following l

represents a summary of reactor operations conducted during this period:

i 1.1.1 The reactor generated a total of 9,379 MWh of energy. Total operating time was 7,176 hours0.00204 days <br />0.0489 hours <br />2.910053e-4 weeks <br />6.6968e-5 months <br /> during calendar year 1992. A total of 13,856 cumulative test hours were logged on five irradiation capsules with direct conversion (thermionic) devices during this period.

i 1.1.2 The reactor was not pulsed. Pulsing capability was removed in 1985 I

for the performance of continuous, in-core irradiations of direct conversion devices.

1.1.3 The reactor consumed 542.34 grams of U-235.

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1.1.4 A total of 13 irradiation requests were processed during the period.

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1.1.5 There was one occurrence during the year which was reported to the U.S. Nuclear Regulatory Commission on April 28,1992. The occurrence as related to the violation of an administrative con.ation in l

the R-67 Technical Specifications, related to the testing frequency of l

the blower used to operate an activated charcoal filter system. The a

system is required to be tested once a week (at intervals not to exceed 10 days). During one period, the interval between tests was eleven days. Corrective action (s) to prevent a recurrence included appropriate i

postings to remind the operations staff of not exceeding the ten-day l

interval between testing, which is performed as part of the weekly j

checklist.

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I 1.1.6 No new applications for facility modification under 10CFR50.59 were approved during the reporting period.

1.1.7 No special experiments, as defined in the R-67 Technical-Specifications, were conducted during this period.

i 1.1.8 No license amendments were submitted or granted during this period.

i 1.1.9 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 were l

required to take - and successfully completed - biennial written exams under the facility requalification program requirements.

i 1.1.10 Two new irradiation capsules (3H5 and UC2-1) with thermionic j

devices were fabricated and installed in in-core positions during 1992.

i 1.1.11 Three irradiation capsules (lH2, lH3 and UC2-1) were removed from j

their Mark F in-core irradiation positions during 1992, and were stored i

in the reactor pool for transfer to the GA hot cell facilities for destructive, post-irradiation examination at a future date. These capsules were removed after irradiation cycles of 14,147,20,998 and j

936 hours0.0108 days <br />0.26 hours <br />0.00155 weeks <br />3.56148e-4 months <br /> respectively.

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i 1.1.12 There was one apparent tramp uranium related reoccurrence with the Mark F in 1992 that was first observed and reported in 1978. A

'l written report of this occurrence was submitted to the NRC on l

t December 9,1992, following a verbal report made on November 10, 1992.

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Table I summarizes pertinent reactor operating parameters for 1992.

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l.2 Facility Chances and Modifications. There were no major changes made in reactor performance characteristics or mechanical design during the reporting period. The facility continued to operate primarily as a thermionics test facility, an operating mode which has prevailed since early 1985. Several changes to the reactor instrumentation and control system to upgrade and install state-of-the-art components and systems were made during the course of the year.

The modifications are described in Sections 4 and 5 of this report. _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ - -

TABLE I

SUMMARY

OF TRIGA MARK F OPERATING DATA Annual Values January 1,1992 Operating Parameter through December 31,1992 MWh of energy produced 9,379 MWD of energy produced 390.8 Grams U-235 consumed 542.34 Number of fuel elements removed from core

• 4 Number of fuel elements added to core m 4

Number of pulses 0

Hours reactor critical (steady state) 7,176

. Number of start-up and shutdown checks 45 Number of irradiation requests processed 13 Number of facility modifications under 10CFR50.59 0

Number of direct conversion device capsules 5

irradiated during calendar year

• Number of cumulative test hours on direct conversion 13,856 device capsules

(

(1)

The number of fuel elements (including FFCRs) removed from the core repre-sents fuel removed as a result of bending or length changes, or otherwise determined to be damagcd or otherwise deteriorated.

(2)

The number of fuel elements (including FFCRs) 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.

(3)

During the course of the year, two new capsules were installed in in-core irradiation positions, and three capsules removed from their core positions. _ _ _ - _ _ _ _ _ _ - _ - _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ - _ _ - - _

l.3 Surveillance Tests and Inspections. Surveliiance tests and inspections were performed as required by Sections 4.0 (Reactor Pool),5.0 (Reactor Core) and 6.0 (Control and Safety Systems) of the R-67 Technical Specifications. A summary of the results are presen'ed below:

1.3.1 Papi Water. The pool water conductivity was measured continually using a sensor installed in the demineralizer input line. Water conductivity was maintained well below the limit of 5 micro-mhos per centimeter averaged over one calendar month required by the Technical Specifications.

Water level sensors were used to ensure that the pool water level

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always was maintained at acceptable levels. In addition, a visual check of pool water level was made as part of the Daily Startup or Shift Change Checklists.

1 A water temperature monitor was used to ensure that bulk pool water temperature was maintained within acceptable limits. During continuous irradiations of thermionic fuel elements in-core, the bulk pool water temperature maintained at 35 1*C.

1.3.2 Reactor Core. The reactor fuel was inspected for bending and length I

changes, as well as visually for deterioration and damage during April 1992. In addition, the inspection also used a plate gage test to check for swollen or deformed cladding, and to ensure that the elements can be removed and reinserted from the core grid plate without sticking.

To perform these inspections, each fuel element is removed from the core, and then reinserted following the necessary inspections. Four FLIP fuel elements were removed from service du..ng 1992 as a result of these inspections. In addition,27 elements failed the 1/32" bend test but passed the 1/16" test and were returned to service. The elements removed from service were: l l

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Element Core Reason Sh Location For Removal 5876 F11 Difficult to insert into grid plate (April 1992) 6346 F25 Difficult to insert into grid plate (April 1992) 5868 D15 Difficult to insert into grid plate (April 1992) 6340 F3 Difficult to insert into grid plate (February 1992)

The four FLIP elements were replaced with four new LEU (30-20) fuel elements respectively. The addition of the LEU elements brought the total number of such LEU fuel elements in the Mark F core to 19.

1.3.3 Control Rods. All five fuel follower control rods (FFCR) were removed from the core and visually inspected for deterioration in April 1992. All were found to be in satisfactory condition.

1.3.4 Reactor Safety Systems. Surveillance and calibration of reactor safety systems I

was carried out as specified in the R-67 Technical Specifications and reactor i

operating procedures. The calibrations and checks on the mam 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

.l manual scram capability performed on a daily basis (except during continuous operations) prior to reactor startup, to ensure that the channels are operating as l

intended, and that the set points for these channels are within the limits specified in the Technical Specifications.

j A calorimetric determination of reactor power is required at least semiannually, and can be performed more often if dictated by the needs of the experiments being carried out. This procedure involves increasing reactor power to an indicated value of approximately 1000 kW, and holding this power level for approximately one hour while the rise in pool temperature is recorded as a function of time. The ratio of the pool heatup rate to the pool constant gives the reactor thermal power. During the reporting period, three _ _ _ - - _ _ - _ _ _ _ _ _ _ _ - _..

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power calibrations were performed, and the calibration constants on each channel (amps / watt) used to convert detector current output to reactor power were recalculated and posted for each of the three power indicating and safety channels.

1.3.5 Radiation Monitorine. 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 thermionic cell top monitor, a control room monitor, and a variety of portable survey meters. Their use and calibration is described below:

Continuous Air Monitor (CAM). During 1992, a continuous air monitoring system was in use for monitoring the air above the reactor pool except for short periods of time necessary for calibration or repair, during which times the CAM was temporarily replaced by a Ludlum Model 300 portable monitor.

The CAM 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 Na-tional 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 f

a Corp.) were used for monitoring area radiation levels in the reactor room.

j The low level monit~ wr

., 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 itself 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 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 on a weekly basis. The monitors were 8

i

calibrated on an annual basis using the calibration range; all calibrations were traceable to NIST.

Thermionic Cell Top Monitor. A radiation monitor is present above the reactor pool water level to indicate a gross leakage of fission products into the purgeable secondary containment for the encapsulated thermionic devices.

The alarm set point was checked on a weekly basis, with NIST traceable cal-ibrations performed on an annual basis.

Control Room Moniter. A radiation monitor is present in the reactor control room to monitor dose rates at the control console. The alarm set point (2.5 mR/h) was checked on a weekly basis, with calibrations performed 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 meter, which was calibrated on an annual basis.

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2. ENERGY GENERATION i

l The total energy generated during calendar year 1992 as a result of Mark F operations was 9,379 megawatt-hours. Figure 1 is a bargraph showing the reactor operation on a monthly basis during the year. The relatively lower energy generation during the months of April, May, July and September reflects reactor shutdowns for reactor inspection and maintenance, thermionic device inspections via underwater neutron radiography, holiday i

shutdowns, or a combination of the above.

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3. EMERGENCY SIIUTDOWNS AND INADVERTENT SCRAMS The total number of unscheduled scrams during 1992 operations was 13. None of the 13 scrams experienced in 1992 had any efect on, or consequencefor, the safe operation of the Mark F reactor. In fact, all safety systems functioned as intended in shutting down the reactor when trip setpoints were reached, or an error condition was otherwise detected in the reactor operating or experimental systems. The causes of the scrams are summarized below:

Scram Channel Cause Number External Loss of site power 1

External CAM improperly bypassed during scram I

checks External Noise in CAM signal from detector 1

Power Level Operator Error 1

Manual Unscheduled shutdown caused by malfunction 1

of control rod 2 drive UP switch contacts Manual Unscheduled shutdown for repairs to 1

experiment power linec External Thermionic device driving power supply 1

malfunction External Thermionic device low voltage 3

External Thermionic device low current 3

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4. MAINTENANCE ACTIVITIES All maintenance activities performed during the year generally fall into three categories:

(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.

Significant activities in this area are described below:

4.1 Reactor. Mechanical and Auxiliary Systems.

January 1992 A new surface skimmer for the water treatment system was installed in the reactor pool.

May 1992 Preventative maintenance on several components of the pit cooling systems was completed. This included cleaning ofinside sections of the cooling tower, replacement of inlet and sump strainers, cleaning or replacement of spray nozzles and lubrication of all bearings on pump motors.

The reactor room crane controller was replaced with an inverter driven, variable speed type, to allow better control of removal and insertion of experiments and components from the core or reactor pool.

July 1992 The core shroud water diffuser pump was replaced.

The reactor room activated charcoal filter required for operations with direct conversion devices was replaced; this was a regularly scheduled replacement.

September 1992 A leak in the main city water supply line to the cooling towers was repaired.

One of the four fuel supports used to install fuel elements in the in-core experimental cutouts was replaced (Position B6-C10-Cll).

The new support is a redesigned version with an inner diameter of _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

l.515 inches, compared to previous ID of 1.507 inches. The slightly larger ID allows easier movement of fuel into and out of the experimental cutout positions.

4.2 Instrumentation and Control System.

January 1992 The self powered detector (SPD) circuit for the in-core SPD readouts in the control room was redesigned and relocated to the main control console from an auxiliary cabinet location.

February 1992 The control rod 2 drive UP switch was replaced. This switch failed in the open position during reactor operation, causing the control rod to withdraw to its full up position, and preventing any UP motion of rods 4 and 5. The reactor was manually scrammed to repair the switch (Section 3).

April 1992 Routine calibration activities of power, fuel temperature and other auxiliary systems instrumentation was performed.

May 1992 Maintenance on CAM circuits was performed in an attempt to locate the cause of the noise in the signal which caused a reactor scram (Section 3). No obvious failure of any component or source of the noisy signal was found; the detector itself was replaced during this activity.

The 5-decade (0.1-10,000 mR/hr) console readout units for the air filter and control room radiation monitors were replaced with 4-decade (0.01-100 mR/hr) units.

August 1992 Trouble in the control rod drive DOWN indicator circuit was traced to a broken jumper wire in the rod control panel and repaired.

September 1992 Test points to measure the power channel detector high voltage settings during routine checks were relocated for easier access. _ _ _ _ _ - _

The reactor room air intake louvre closure mechanism was repaired and upgraded to allow remote actuation of the closure mechanism from the reactor control room.

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5.10CFR50.59 FACILITY MODIFICATIONS AND SPECIAL EXPERIMENTS There were no applications for facility modi 6 cations under the provisions of 10CFR50.59 that were approved for the R-67 facility during the 1992 reporting period. In the implementation of such facility modi 6 cations, the applications for the proposed changes to the R-67 facility are reviewed by the TRIGA Reactors Facility Safety Committee, among others, and approved only after making the determination that the proposal for the change (a) did not involve a change to the R-67 Technical Specifications, or (b) did not create any unreviewed safety questions as de6ned in 10CFR50.59.

No Special Experiments as defined in the R-67 Technical Specifications were conducted during 1992.

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6. RADIOACTIVE EFFLUENTS DISCHARGED TO TIIE ENVIRONMENT During the calendar year 1992,3.127 curies of Argon-41 were discharged from the Mark F reactor facility to the atmosphere.

All low level radioactive wastes were transferred to GA's Nuclear Waste Processing Facility - /hich operates under NRC license SNM696 and GA's California Radioactive 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 during 1992:

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 l

accordance with GA's SNM-696 license.

Daily liquid effluent monitori g 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 facilities 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 Exposuresm Number of employees monitored:

21 High Exposure:

0.145 1Aw Exposure:

0.000 Average Exposure:

0.029 8.2 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

• l l

Number of persons monitored:

83 High Exposure:

0.185 Low Exposure:

0.000 Average Exposure:

0.006 i

8.4 Visitor Whole Body Exposures

• Number of persons monitored:

59 High Exposure:

0.185 Low Exposure:

0.000 Average Exposure:

0.004 I

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8.5 Routine Wioe Survevs of Mark I Reactor Facility High Wipe:

174.7 Beta dpm/100 cm 2

2 low Wipe:

2.3 Beta dpm/100 cm 2

Average Wipe s1 Beta dmp/100 cm r

8.6 Routine Radiation Measurements of Mark I Reactor Facility High Measurement:

22 mrem /hr @ l foot Low Measurement:

0.5 mrem /hr @ l foot Average Level:

< 0.1 mrem /hr @ 1 foot 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.

(3' Includes non-GA personnel who were granted periodic access to the facility for the performance of work.

l 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.. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

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