ML20246N760

Annual Rept Covering Period Jul 1988 - June 1989 for Activities of Triga Mark I Reactor

ML20246N760
Person / Time
Site:	05000113
Issue date:	08/31/1989
From:	Nelson G ARIZONA, UNIV. OF, TUCSON, AZ
To:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
References
NUDOCS 8909080258
Download: ML20246N760 (5)

Text

. , f7j f 4; f THE UNIVERSITY OF ARIZON A d 3M_%x,I l T U C S O N, ARIZON A 85721

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~" COLLEGE OF ENGINEERING AND MINES DrPARTMENT of NUCLEAR AND ENERGY ENGINEERING i

August 31, 1989 I l

U. S. Nuclear Regulatory Commission l Region V l Office of Inspection and Enforcement 1450 Maria Lane, Suite 210 Walnut Creek, California 94596 i RE: Annual Raport for License R-52, Docket 50-113 Gentlemen:

This is the Annual Report covering h period July 1,1988 through June 30,1989, for the activities of the TRIGA Mark 1 Reactor at the University of Arizona, Tucson, Arizona. This report is submitted in compliance with Section 6.7e of the Facility Technical Specifications and Paragraph 50.59(b) of Title 10, Code of Federal Regulations.

1. During the reporting period, the reactor was operated for research and education. It was used i

for reactor operator training of replacement operators at this facility. The reactor was used for graduate thesis research and undergraduate Nuclear Engineering course experiments, including approaches to critical, control rod calibrations, measurements of the dynamic response of the reactor to step and periodic changes in reactivity, and flux mapping. The reactor was also used for neutron activation analysis and production of short-lived radioisotopes for teaching and research. Less than 1.0 percent of total reactor operating time was used for non-university-related purposes.

Reactor upgrading and modification included the repair and upgrading of the Fast Irradiation Facility (fir). The boron layer of neutron absorbing material, which had been formed by a slurry of boric acid powder, was replaced with a sheet of boron impregnated rubber. This change provides more uniform flux distribution in the fir.

Routine surveillance tests of the power channels, including recalibration, showed only minor changes in zero adjustment and full scale trip setting. The total worths of the regulating, shim, and transient rods were measured to be $3.85,53.04, and $2.43 respectively. The largest change in worth was 2.3 on the regulating rod which is consistent with the small changes in rod worth due to rotational changes of position of individual fuel elements from fuel movement during approach to critical l experiments.

No fuel elements were measured for length or bend during the reporting period because the number of pulses since the last measurement is below the surveillance requirements in the facility technical specifications.

The transient rod drive assembly was inspected twice during the reporting period. Both piston seals were found to be in satisfactory condition and no wear or rust accumulation was present in the air cylinder.

Rod drop times from full out to full insertion were measured to be 0.35,0.33, and 0.88 seconds for the regulating, shim, and transient rods, respectively. There were no appreciable changes in the C

h 1 I 9

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• Augpst 31, 1989-drop times of the regulating and shim rods since the last rod drop measurements. The drop time of the transient rod increased slightly from the previous measurements. All three drop times were less than the time required by the facility technical specifications.

The regulating shim, and transient rods were visually inspected during the reporting period. The rods showed normal signs of wear and no deterioration. The inspection was performed at an interval greater than that allowed by the facility technical specifications. This matter was brought to the attention of Region V in a letter dated May 9,1989.

2. The reactor was critical for a total of 132 hours0.00153 days <br />0.0367 hours <br />2.18254e-4 weeks <br />5.0226e-5 months <br />, producing 6031 kw-hours (0.251 Mw-day) of thermal energy. The cumulative energy output since the facility was commissioned is 8.078 Mw-days. During the reporting period 57 pulses with input reactivity greater than $1.00 were performed. The cumulative number of pulses greater than $1.00 since the time pulsing was initiated is 1567.

The reactor was in operation 110 days during the reporting period, with approximately 438 hours0.00507 days <br />0.122 hours <br />7.242063e-4 weeks <br />1.66659e-4 months <br /> of operating time.

3. There was one inadvertent reactor scram during the reporting period. On March 8,1989, a left safety channel scram occurred when an experimenter inserted a sample into the pneumatic transfer system while the reactor was operating at 100 kw. At the time, the F1r was out of the reactor core causing the left safety channel to read 104 kw at a thermal power level of 100 kw. As this was a conservative condition, the left safety channel had not been readjusted to read correctly when the fir was removed for modification. With the reactor operating in automatic mode, there was a slight increase in local flux at the instant the sample entered the core position (approximately 2 kw of indicated thermal power). The left safety channel scram setting was 106 kw (also conservative under the current facility technical specifications). The combination of indicated power of 104 kw and a scram setting of 106 kw (both conservative) on the left safety channel caused the inadvertent ,

scram when the sample was inserted in the pneumatic transfer system. No safety limit was exceeded and the reactor was brought back to 100 kw after review by the Senior Operator on duty.

The left safety channel was adjusted to read 100 kw and was later readjusted when the Flr was '

returned to the reactor core.

4. Major maintenance includes the replacement of bearings in the pneumatic transfer system blower motor. Minor maintenance items include replacement of a 15 volt power supply in the control console, greasing of reactor room exhaust fan bearings, checking the thermecouples in the instrumented fuel element for continuity, changing filter cartridges in the wates purification system, '

adding pool water lost by evaporation, replacing burned out light bulbs in the reactor pool, replacing burned out annunciator bulbs in the reactor control console, and making periodic adjustments to reactor control console circuitry.

5. The reactor committee met six times during the reporting period, (9/28/88,10/14/88,11/16/88, 2/17/89, 4/28/89. 6/16/89). The following items of business were covered-The Committee approved the relicensing application for submittal to the NRC after offering changes in the Safety Analysis Report and The Emergency Plan.

The Committee reviewed a report on the repair of the recirculation pump in the water purification system during the previous reporting period. This repair was performed using a procedure which had been temporarily approved by the Director of the NRL as permitted by facility procedures.

The Committee reviewed and approved UARR136 (Procedure for Repair or Replacement of the Recirculation Pump in the Water Purification System) with minor modifications.

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. . August 31, 1989 The Committee reviewed and gave temporary approval to amendments to UARR101 (Emergency Procedure 3).- These procedures were to be evaluated in the upcoming annual emergency drill.

The Committee reviewed a report on the replacement of a 15 volt power supply. They concluded 4 that the replacement supply exceeded the manufacturer's specifications of the original supply.

The Committee reviewed three individual critiques of the annual emergency drill. They approved the suggestions for improvement in the summary and suggested an additional checklist for those persons or support organizations that had been contacted.

!' The Commlttee reviewed and temporarily approved UARR139 (Procedure for Repair or Modification of the Fast Irradiation Facility (fir)). Subsequently, discussing the beta radiation doserates contributed by the irradiated absorber materials in 'the fir, the procedure was approved.

The Committee reviewed and approved UARR140 (Procedure for Pulse Chamber Calibration).

The Committee reviewed and approved a change in UARR107 (Procedure for Control Element Removal and Inspection) reducing the waiting time from reactor shutdown to removal of a control element.

The Committee reviewed a report from the Director of the NRL regarding a surveillance inspection l item that had not been completed in the time frame mandated by the facility technical i specifications. The visual inspection of the control rods was to have been completed by January 19, 1989. This was discovered on April 28, 1989 by a Committee member performing a facility operations review. The Director notified Region V of the NRC and terminated all reactor operations until the inspection had been completed. The Committee discussed ways to prevent future occurrences of this nature.

The Committee reviewed a report on the repair of the fir using the approved procedure. Flux l calibration is to be done on the fir following the visual inspection of the control rods.

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The Committee reviewed a report on the construction and testing of a scif-powered neutron detector that will be used to measure the peak power of reactor pulses for a Master's thesis.

Committee members observed reactor operation, reactor pulsing operation and fuel movement by the NRL staff. At its meetings and in individual reviews by Committee members, the Committee reviewed operations and operational records of the facility as specified by the Committee charter.

This included audit of preliminary check sheets, pulsing check sheets, approach to critical and termination check sheets, operations and maintenance logbooks, monthly and annual checksheets, irradiation records, and experiments performed with the reactor. The Committee received reports about new operator trainees, laboratory class enrollment, and reviewed the 1988 annual report to the l

NRC.

6. No liquid waste was discharged from the facility during the reporting period. 4.0 cubic feet of solid waste (floor sweepings, tissues, plastic gloves, decayed activation samples and standards) containing trace quantities of mixed irradiation products was collected by the University Radiation Control Office to be held in storage for pickup by Thomas Gray and Associates. Thomas Gray collects the stored radioactive waste with their own vehicles and transports it to a holding area in Orange, CA. From their it is shipped for burial to either the U.S. Ecology Site at Richland, WA or Beatty, NV. The collections from the Reactor Laboratory by the Radiation Control Office were made in January, April, and h!ay of 1989.

Measurements of the Argon-41 concentration in the reactor pool water have demonstrated that the

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l rate of decrease of Argon-41 from the surface layer of water is the same as the radioactive decay l- rate, and thus that there is no appreciable release of Argon-41 from the pool surface into the I

reactor room air. Using the 3-sigma lower limit for these concentration measurements, the maximum rate of release of Argon-41 from reactor pool water is test than 0.34 pCi per kilowatt-hr of reactor operation, and certainly much less. The Argon-41 release from the pneumatic transfer system for this reporting period is based on a production rate from measured activity of small vials of air that have been irradiated in the system. Resident air in the pneumatic transfer system I produces approximately 0.05 pCi of Argon-41 per kw-min of reactor operation prior to inserting a i sample in the system. Presented below are the monthly releases of Argon-41 from the reactor pool surface, the pneumatic transfer system and the totals.

Argon-41(pCi) Argon-41(pCi) Argon-41(pCi)

Month Pool Surface Pneumatic Transfer System Total (using upper limit)

July 1988 0 73 310 383 August 1988 0 1 510 1 511 September 1988 0 1 277 125 402 October 1988 0 t 152 0 152 November 1988 0 i 119 0 119 December 1988 0 i 19 0 19 January 1989 0 i 21 0 21 February 1989 01 40 252 292 March 1989 01214 1009 1223 April 1989 0 1 320 253 573 May 1989 0 i 197 9 206 June 1989 0 i 109 0 109 The calculations for the Argon-41 release from the pneumatic transfer system did not include decay of the isotope prior to inserting a sample in the system, and therefore, represents a very conservative estimate of Argon-41 relense. The maximum total estimated Argon-41 release from the facility during the reporting period using the 3-sigma upper limit for release from the pool surface, is 4 millicuries. This is less than 0.6 percent of the allowable Argon-41 release allowed under 10 CFR

20. There were no other gaseous effluents from the facility during the reporting period.

7. Fifty-five (55) persons were issued film badses on a monthly basis for all or part of the reporting period in the Department of Nucles and Energy Engineering. The persons receiving badges included all reactor operators, faculty and staff members using the reactor laboratory, l researchers and all students in laboratory courses. Two badged individuals were reported as having received 25 mrem and 15 mrem whole body exposure of penetrating radiation. Three badged individuals were reported as having received 80 mrem,60 mrem, and 30 mrem ring badge exposure of penetrating radiation. Two of these individual received their reported exposure peforming activities not associated with the reactor. In each case of reported exposure, the individual was interviewed by the Director of the NRL and methods of reducing or eliminating further exposure were discussed and implemented. Documentation of these interviews form a basis for training facility personnel as part of the facility's ALARA program.

Three hundred sixty-two (362) non-badged persons were admitted to the Reactor Laboratory in classes, tours, or on official business during the reporting period. All groups were issued pocket dosimeters or were admitted only after completion of a radiation survey which showed all dose rates to be less than that in an unrestricted area as required by the facility procedures. No radiation  :

exposure was received by any visitors, as measured by the pocket dosimeters.

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8. Radiation surveys of reactor room,' control room, and experiment set-up room were conducted monthly during the reporting period by members of the University of Arizona Radiation Control Office using direct measurement and wipe tests. The results show little detectable activity except where expected (i.e., irradiated samples in storage areas and internal wall surfaces of the irradiation i facilities). Other radiation surveys were performed by members of the reactor laboratory staff when necessary. . During the repair of the fir, the activated cylinders of gold and cadmium were stored in a 3-sided lead cave in the reactor room. During one monthly reporting period, the area monitor film badge on the opposite side of the wall which is the fourth side of the lead cave, showed an exposure of 170 mrem. The room containing the area monitor is adjacent to the reactor room and is used for storage. A two inch thick layer of lead bricks was placed at the back of the lead cave to prevent radiation passing into the adjacent room.

9. Environmental TLD monitors at 3 locations on the building housing the reactor and at 10 other locations on the University campus were replaced and read monthly. For the 12-month period from July 1,1988 through June 30, 1989, the average reading of the 3 TLD's located on the reactor building, was 14.3 mrem and the average of the 10 other TLD's was 21.9 mrem, relative to two control TLD's which were kept in a shielded container in a non-radiation area. This is consistent with similar measurements for these locations in previous years. Thus, there is no evidence that radiation exposures in the vicinity of the reactor are higher than normal, but rather, as indicated by the TLD's, are slightly lower than the average for the University.

In writing this report, I have tried to be both complete and as brief as is reasonable, and still satisfy the requirements of the 10 CFR 50.59, the Facility Technical Specifications, and the needs of the Commission. If other or more detailed information is needed, please contact me at your earliest convenience.

Sincerely Y 'A&

George W. Nelson, Director Nuclear Reactor Laboratory GWN/sh I cc: Document Control Desk USNRC Washington D. C. 20555 Ernest T. Smerdon, Dean College of Engineering and Mines I

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