ML20248E288

Provides Operator Requalification Program to Comply W/Amend to 10CFR50

ML20248E288
Person / Time
Site:	05000356, University of Illinois
Issue date:	01/10/1975
From:	Beck G, Wehring B, Wyman M ILLINOIS, UNIV. OF, URBANA, IL
To:	US ATOMIC ENERGY COMMISSION (AEC)
Shared Package
ML19324D199	List: ML20248E108 ML20248E133 ML20248E147 ML20248E288
References
NUDOCS 8910050138
Download: ML20248E288 (3)

Text

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2H Xuclear Engineering Laboratory -

' NUCLEAR ENGINEERING' PROGRAM AT URBANA-CHAMPAIGN Urbana, Illinois 61801 jrea Code 217-333 2295 January 10, 1975 Director

-Directorate.of Licensing Office .of Regulation .

U.S.A.E.C.

Washington, D. C. 20545 9

SUBJECT:

OPERATOR REQUALIFICATION PROGRAM University of Illinois TRIGA and LOPRA Reactors Docket Nos. 50-151 and 50-356 The following is submitted to comply with the amendment to 10 CFR 50 which became effective on September 17, 1973.

General Provisions A. Implementation: A program of continuous training of personnel who have a current Senior Operator or Operator License from the A.E.C. has been followed since December,1973.

B. Administration:

1. The Reactor Supervisor, or a designated alternate, is in charge of supervising the program and specifying scheduled times for meetings and tests.

2. Records on attendance to lecture or discussion periods are recorded for each meeting. Evaluations on individual performance is included in the individual summary.

3. The individual in charge of the evaluation is exempt' from taking tests for the purpose of the individual evaluation.

Requalification Program A. Operation: Each licensee shall spend a minimum of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> at the console manipulating the controls. These manipulations shall include, but not be limited to: completion of the daily check list and start-up from shutdown to 15 watts; changes in the power level above 10 kilowatts in both natural and forced convective cooling; operation of either square wave or pulsed transients; and the shutdown of the reactor w'ith the completion of the shutdown check list.

The performance of each operator shall be evaluated by the Reactor Supervisor, or his designate, at least once each year.

B. Lecture and Discussion Periods: There shall be a minimum of six lecture or discussion periods each year. During the first year, the topics to be covered are those given in 10 CFR 55.21 and 55.22. Any changes in plant design, controls, instrumentation,

i. or procedures will be distributed to all licensed operators.

l A review of any changes will be made during a discussion period.

An annual review of the emergency procedures, abnormal occurrences, emergency plans, and the building security plan will be made at these meetings. ,

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i Page-2 j Oper. Requal. Program Univ. of Illinois During subsequent years, emphasis will be placed on topics where the performance on oral or written tests has shown a i weakness. Licensed operators will be expected to attend the sessic,ns during the first year after obtaining their license and during, subsequent years depending on the results of the exams as given in Part C q

Individuals who have not manipulated the controls for a period of four months or greater will be required to demonstrate their operational skills and knowledge before assuming the

. direct responsibility as the reactor operator. If the operator has been absent from the facility during this period, he will be required to take the equivalent of an annual written examination or shall spend a minimum of five hours manipulating the controls under the supervision of a licensed senior operator before taking over the controls.

An evaluation of his performance will be documented as in Part C. This will meet the requirements of 10 CFR 55.31(c).

C. Examinations: A written examination will be given to all licensed operators, except the individual in charge of the requalification program, on an annual basis. The examination will contain questions similar to those on an A.E.C. exam.

Evaluation shall be on the following basis:

1. Less than 70% on the overall examination will result in a termination of any licensed duties until the ability of th'e individual is shown to be adequate from further examinations.

These individuals will be required to attend the lecture and discussion periods during the following year in order to be recommended for an extension of his license when it expires.

2. A grade of 80% on each of the individual sections of the examination will be used as che criterion for participation in the class sessions during the following year. If an

, individual receives less than 80% on a given section, he will be required to parti'cipate in the session where the material is covered.

3. The individual evaluating an operator's performance on the oral examination (Part A) will make a written judgement on the adequacy of the examinee. The examiner may require additional training in specific areas or may recommend the termination of the licensed duties of the examinee until his performance is shown to be adequate. ,

Note: Because of the limited number of licensed operators at the facility, class sessions on specific topics may not be necessary due to the results on the written examinations.

Thus the actual number of classes may vary depending on new operators and exam results.

Page.3 Oper. Requal. Program Univ. of Illinois D. Records: The following records shall be kept on those individuals who have a current senior operator or operator license from the A.E.C.

1. Operation time including the type of operation.

2. Lecture and discuss. ion periods including the individuals attending same.

3. Copies of written exams administered.

4. Answers to exams by the licensees including the evaluation by the grader. (this includes the annual exam)

5. Evaluation of. oral exams by the examiner including the date of the exam and the amount of time utilized.

6. A documentation of any additional training which has been administered in an area where a reactor operator or senior reactor operator has exhibited deficiencies.

Yours truly,

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. ' ,6L.-dA4 a .4 &

C6rald j P. Beck, Reactor Superi/isor Nuclear Reactor Laboratory b Y NcN $/MPW .

W Bernard S. Wehring, Chanrman Marvin E. Wyman, Chairmth Nuclear Reactor Committee Nuclear Engineering Program Approved: 11-15-74*

• The content of this document was approved on this date.

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ANNUAL REPORT August 1, 1987 - July 31, 1988 ILLINOIS LOPRA REACTOR Facility License R-117 1

SUMMARY

OF OPERATING EXPERIENCE The reactor was scheduled for operations a total of 53 hours6.134259e-4 days <br />0.0147 hours <br />8.763227e-5 weeks <br />2.01665e-5 months <br /> and was in actual operation a total of 21.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Usage was similar to what it was in the previous reporting period although the pattern of usage was different than the last reporting period. Student Experiments include an " Approach to Critical fuel loading experiment in the LOPRA and the determination of Control Rod Worth by sub crstical multiplication. The time involved with Surveillance Requirements includes power calibrations, control rod worth determinations and fuel inspections. Scheduled time includes time for activities such as fuel inspections which don't require actual operation of the LOPRA. Over 99.9 % of the energy listed in Section it was generated during the two power calibrations performed during this reporting period.

Maintenance performed in this reporting period includes the removal and repair of the LOPRA Compensated lon Chamber (CIC) and the repair of the pulley system used to position the LOPRA Assembly relative to the thermal column.

Maintenance 31 %

Student Experiments 15 %

Survesilance Requirements 45 %

Laboratory Measurements 9%

Total 100 %

11. TABULATION OF OPERATIONS Hours Critical

• end Enerav 21.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 6.33 kW-hrs

• This time includes that for loading fuel elements during the Approach to Critical experiment and sub-critical time during the control rod calibration. The control rods are calibrated using sub-crstical multiplication so that a large portion of the experiment is done with the reactor sub-critical. The actual erstical time was about 3.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or 6 % of the above time.

I lit. EMERGENCY SHUTDOWNS AND INADVERTENT SCRAMS There was one unplanned scram and no emergency shutdowns dureng this 3 per t od. The unplanned screm was due to a trainee (student) switching error.

The trainee mid-posetioned the Range Selector Switch on the Linear Power Channel, resulting in an up-scale response by the Linear Channel Chart Recorder which initiated the scram. Students are routinely advised to make {

range changes smartly to avoic switching problems.

IV. MAINTENANCE 1

l Maintenance on the LOPRA included the removal, repair and re-installation of the LOPRA CIC. This was accomplished by partially unloading the LOPRA core and sending a diver into the Bulk Shielding Tank to first remove and later rt.

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install-the Guide' Tube for the LOPRA CIC. The final report on this evolution; is attached as sn addendum to thi?. report.

.Another significant stem of maintenance performed during this reporting pertod wat the repair of the pulley system which is used to position the LOPRA core relative to the thermal column.

Several years ago, the anchors that held the pulley system in place pulled away from the south wall and the bottom of the Bulk Shielding. Tank.

When this occurred, the LOPRA Assembly was manually positioned to allow enough clearance for the boral curtain to be raised and lowered to de-couple the

.LOPRA whec it was shutdown. Four graphite dummy elements were added to'the I-row of the LOPRA to gain back some of the -reactivity lost due, to the positioning of the LOPRA assembly. This provided twelve to fifteen' cents excess reactivity for operation of the 'LOPRA. Typically fourteen cents of.

excess reactivity is needed for 10 kW operation of the LOPRA.

The new pulley system is anchored to the south wall near the top of the Bulk Shielding Tank 'and held in place by the tracks upon which the LOPRA.

Assembly is moved back and forth. The LOPRA is free to move over its entire l range of travel,

.The last item of maintenance performed during the reporting period was the overhaul of'the Poison Control Rod. Position Indication' system. This system relied on an obsolete servo driver similar to the one which used to drive the Poison Control Rod servo motor. The replacement for this obsolete driver uses one-tenth the power of the old unit and is electronically superior to the old driver. This unit is less susceptible to electronic noise than the old unit.

V. CONDITIONS UNDER SECTION 50.59 0F 10 CFR 50 There were no changes to procedures or new experiments during this period. The excess reactivity of the LOPRA core was increased to about 22 cents during this reporting period as a result of the repast of the pulley system for moving the LOPRA Assembly relative to the thermal column. This will allow the flexibility to use the pulsing feature of the LOPRA. This 6ncrease leaves the excess reactivity well below the maximum Technical Specification limit of sixty cents.

VI., Vll., Vill., RADIOACTIVITY Because of the lower power and infrequent use of the LOPRA, its operation j doet, not contribute to the release of effluents. Personnel radiation exposures l -. for the laboratory are given in the Annual Report for the Advanced TRIGA Reactor. License No. R-115, Docket No. 50-151, dated February 29, 1988.

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ANNUAL REPORT August 1, 1986 - July 31, 1987 ILLIN0IS LOPRA REACTOR Facility License R-117

1.

SUMMARY

OF OPERATING EXPERIENCE The reactor was. scheduled for operations a total of 46 hours5.324074e-4 days <br />0.0128 hours <br />7.60582e-5 weeks <br />1.7503e-5 months <br /> and was in actual operation a total of 25.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Usage was a little over double what it was in the previous reporting period, but less than half what it has been in the recent past. This apparent decrease over the recent past is due to the fact that the local utilities who usually send power plant trainees through our training program are either involved in ascension to commercial operation or adjusting training to match projected fuel loading and did not send any trainees during this reporting period. The training program includes an

" Approach to Critical " fuel loading experiment in the LOPRA. The time involved with Surveillance Requirements includes power calibrations and control rod worth determinations. 99.7 % of the energy listed in Section it was generated during the three power calibrations performed during this reporting period.

Operator Training (Facility - new RO's) 18 %

Student Experiments 26 %

Surveillance Requirements 43 %

Laboratory Measurements 13 %

Total 100 %

11. TABULATION OF OPERATIONS Hours Critical

• end Enerov 25.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 5.69 kW-hrs

• This time includes that for loading fuel elements during the Approach to Critical experiment and sub-critical time during the control rod calibration. The control rods are calibrated using sub-critical multiplication so that a large portion of the experiment is done with the reactor sub-critical. The actual critical time was about 6.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> or 24 % of the above time.

111. EMERGENCY SHUTDOWNS AND INADVERTENT SCRAMS There was or.e unplanned scram and no emergency shutdowns during this period. The unplanned scram was due to system noise believed to be caused by a bad connector in the Compensated lon Chamber (CIC) channel.

l IV. M AJ NT E N ANC E No maintenance on the LOPRA was required during this reporting period except for the replacement of the aforementioned bad connector in the CIC l Channel. Considerable maintenance was performed on the LOPRA during the i previous reporting period. This effort eliminated most difficulties with the l

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V. CONDITIONS UNDER SECTION 50.59 OF'10 CFR 50 There were -no changes .to procedures or.new exper imen t s- during this -

p e r i o d ., The excess reactivity of_'the LOPRA core was about 12 cents during this reporting . period. This would have . allowed- a. maximum . power of -about 10 ktiowatts.

'The'LOPRA was made inoperative on.the 23rd of' July 1987 in anticipation-of removal of the CIC.for inspection of its' connectors and determination of

'the insulation-resistance of its leads. This action was also taken- to allow

'for modification of the pulley and winch system which positions the LOPRA-platform; relative to the thermal column. One set of pulleys has pulled away from the Bulk Shielding Tank well. The modified-system will not concentrate so-much stress on the pulley anchoring.

The above modification of the LOPRA pulley system will be reviewed by the Reactor' Committee before it is cartned out.

It is felt.that these two items should be performed before re-licensing of.the LOPRA is requested in 1989.

VI., Vll., Vill., RADIOACTIVITY Because of the lower power and infrequent use of the LOPRA, its operation does not' contribute to-the release of effluents. Personnel radiation exposures for the laboratory are given in the Annual- Report for the Advanced TRIGA' Reactor, License No. R-115, Docket No. 50-151, dated' February.27, 1987.

t'

214 Nuclear Engineering laboratory

' NUCLEAR ENGINEERING PROGRAM UNIVERSITY OF ll.LINOIS

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K Urkna, Ilknou- 61801 AT URBANA-CHAMPAIGN area code 217-333 2295 May 7, 1975 l Director Division of Reactor Licensing U. S. Nuclear Regulatory Commission Washington, D. C. 20555

SUBJECT:

Leakage of Air from Aluminum-Clad TRIGA Type Fuel Elements Utilized in LOPRA Reactor Assembly.

License No. R-117, Docket No. 50-356.

Incident During the annual surveillance check on the LOPRA fuel elements, it was noted that as several of the fuel elements were lifted out of the grid plate, air bubbles were coming from the low portion of the element.

On a check of 11 elements on April 23, the last one that was removed gave indications of air being released from the lower portion. This element was removed from the core matrix and placed in a rack at the side of the bulk shielding ~ tank. The checks on the elements were continued on April 29 when'4 more elements indicated leakage out-of 17 that were checked.

In all cases the leakage appeared to come from the lower part of the element.

One of the elements was removed on April 30 to inspect the lower part where the end fixture was attached to the cladding. This check showed-that the leakage was coming from the welded portion. Some corrosion in this area was evident, but the cause of same has not been fully determined.

There was no detectable release of radioactive effluents from the fuel elements. A NMC Continuous Air Particulate Monitor showed no gerease in the counting rate - this instrument has a sensitivity of about 10- uCi/ml.

A 200 cc water sample was taken from the bulk shielding tank (location of the fuel elements) and evaporated for a radioactivity check with a flow counter.

The. counting. rate that was obtained was essentially background. For the size of sample that was used gnd the counting time utilized, the sensitivity for this check was about 10' uCi/cc.

Information on Fuel Elements The fuel elements involved were received from AFRRI in May,1965 for use in suberitical experiments. With the exception of 6 standard elements and 4 thermocouple elements, tle remainder had either damaged end fixtures or had exceed the 100 mil limitation on growth. The tubing on the TC elements had been cut about 6-inches above the top fixture in order to fit in the shipping cask. After receipt of the elements, the end fixtures were repaired by securing the elements in a lead fuel cask and then bending the fixture back into place. A cap was welded on the TC elements with a ring at the top to allow movement under water.

The six standard elements, which were within damage specifications, were used in the Mark II TRIGA assembly for brief periods of time in 1967-68. These were used as additional fuel for making reactivity measurements with the B-ring fuel elements removed.

After the LOPRA license was received in December,1971, the elements have been used for the operation of this facility. To extend measurements being made in the area of " Reactor Coupling," an instrumented element that had OyGgOt / g

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Page 2 been retained when the Mark II fuel elements were transferred to Michigan State was repaired to obtain temperature information. In September, 1974, during the movement of fuel elements for a sub-critical experiment, air bubbles were seen coming from the upper portion of the TC elements that had been obtained from AFRRI. The suspected elements were placed in the racks at the side of.the tank for later checks. In January, 1975, repairs were made on the upper portion of these elements. However, two of the three elements that were resealed at that time were found to be leaking during the checks on April 23 and April 29.

On April 1,1975, the Mark II instrumented fuel element was found to have a severe bend in the thermocouple extension tube. The element was removed and a crack was found in the extension tube where the bend was located.

Since the crack was below the area where the tube is filled with epoxy, it was surmised that some water had leaked into the fuel portion. Before repairs were initiated on this element, the cladding was heated and the moisture in the element was collected in a balloon to check on any activity that might be released in this process. Data on this is given in the following section. i Radiation Hazards The fuel elements in question have a dose rate of 20-100 mr/hr/M, which indicates a total activity of about 0.2 - 1.0 curies per fuel element. The fractional portion of this activity that one would expect to je released from the fuel meat is estimated to be from 2 x 10 -6 to 2 x 10 . (

Reference:

Pages XIV-16 to XIV-22 of the Safety Analysis Report for the Illinois Advanced TRIGA, August, 1967.)

Approximately 10 cc of water was collected from heating the Mark II instrumented element. This water was evaporated and an activity of 0.01 uCi was measured with a flow counter. An attempt was made to possible identify the isotopes with a 4000 channel analyser, however the background in the laboratory was found to overshadow a source of this strength.

This element has a dose rate of 20 mr/hr/M.

With proper precautions, work can be done on the elements with only a moderate amount of exposure to the individual. 'Ibe mechanic, who repaired the three AFRRI fuel elements, received a total dose of about 270 mrem.

It should be noted that the LOPRA has not been operated above 2 watts since March 25, 1975 so that there would be essentially no short half'. life q fission products involved. Since the license was received in December, 1971, 4 the total operation has been 39 kilowatt-hours or les.s than 1.0 kilowatt- l hours per element. The last operation above 10 kilowatts (licensed power for LOPRA) occurred in February,1963 at which time they were removed from '

the AFRRI facility. ,

t Review The instrumented fuel elements described above had defects in the aluminum extension tube that is used for the thermocouple wires. The actual j cladding was not involved as is the case for the standard fuel elements. j Region III, Directorate of Regulatory Operations was notified by phone of j the difficulties with the AFRRI instrument elements in September,1974. j It was agreed that this would not be a reportable incident, but that the  ;

elements should not be utili::ed in the facility until the expected leakage j was corrected. '

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, Page 3 Region III, Office and Inspection and Enforcement, USNRC was notified '

by phone .on ' April- 30; 1975 concerning the difficulties with'the standard fuel-elements. It was. concluded that a report to the Division of Reactor _ Licensing, USNRC, h'ashington, D.C. should be filed giving the present situation as well as any conclusions concerning possible repair of the elements with the defects near the cladding.

The personnel, in charge of the TRIGA facilities at General Atomic Company,L San Diego, California,- have been notified. The impression obtained from the GA personnel was that although ' difficulties should be expected in re-welding the.end plug to the cladding, an individual experienced in aluminum welding could very likely obtain a good seal at this . juncture. Drawings of the original AFRRI elements have been obtained, and before any welding is attempted, communications will be made to the personnel at General Atomic who perfom these duties.

An approval of a basic procedure to be utilized has been approved by the Nuclear Reactor Committee at a meeting on.May 1, 1975. This approval is contingent on the notification of the Division of Reactor Licensing of the plans. One of the fuel elements was inspected by two staff members from the Mining'and Metallurgy Department at the University of Illinois. Their general conclusions were that t,he openings could be due to a combinations of a slight defect in the original weld and later stresses placed at this j uncture. The appearance of corrosion then results from possible impurities in the weld defect. These individuals will also check on other possibilities of sealing the plug. to the cladding other than by aluminum welding.

Conclusions

. At the present time it is planned to attempt to repair the fuel elements following the basic procedure given below. It should be noted that this procedure is based on the fact that the activity is of a nature that can easily be handled with some precautions. The procedure will be tried on a single element and may be altered on future elements depending on the results.

, Steps: 1. Remove approximately 1/16 inch of the plug-cladding interface.-

This should remove the present weld and allow the plug to be removed from the fuel element.

2. Check inside portion of cladding and outer portion of the plug for any evidence of corrosion or possible' damage.

3. Place fuel element in furnace and heat to about 150 C to remove any moisture that may have gotten into this area. During this process, the moisture will either be collected or air particulate will be used to check on any activity that is released. In either case, the NMC Air Particulate Monitor will be in operation to check on any gaseous effluents released.

4. Clean out inner portion of the cladding and the outer portion of 4 the plug.

y 5. Re-weld the aluminum plug to the cladding. Check element for '

any leaks.

Note: Shielding will be placed around the fuel section of the element during any inspection, cleaning, and welding. Personnel i

monitoring will.be done with a self-reading dosimeter and a film badge for those individuals doing the work.

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page 4 Of major ' consideration is the fact that these ' fuel elements will be used in the LOPRA, which has a power limitation of 10 kW and an excess reactivity limgtation of 60 cents. The maximum fuel temperature at 10 kW is about 15-20 C. above the water temperature, and there is essentially no change in the temperature next to the cladding. Later damage to these fuel elements is far more likely to occur from moving the elements in and out of the core than from reactor operation.

f Yours truly, atl, ' 4 . ^I Gdrald P. Beck, Reactor Supervisor

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Nuclear Reactor Laboratory University of Illinois Urbana, Illinois 61801 Copy: Region III, Office of Inspection and Enforcement U. S. Nuclear Regulatory Commission 799 Roosevelt Rd.

Glen Ellyn, Ill. 60137

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ANNUAL REPORT-August 1, 1985 - July 31, 1986 ILLINOIS LOPRA REACTOR Facility License R-117 j

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SUMMARY

OF OPERATING EXPERIENCE i The reactor was scheduled for operations a total of 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br /> and was'in actual. operation a total of 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br />. Usage was only about 20 % of what it was in the previous reporting period. This was due to several reasons: the usage in the. previous period was much higher than average and the local utilities who usually-send power. plant trainees through our training program are either nearing fuel load or adjusting training to match projected fuel loading. The

- training program includes an " Approach to Critical

• fuel loading experiment in the LOPRA. The time involved with Surveillance Requirements includes power calibrations and control rod worth determinations. 99.8 % of the energy listed in Section 11 was generated during the two power calibrations.

Operator Training ( Facility - new SRO ) 37 %

Student Experiments 21 %

Surveillance Requirements 42 %

11. TABULATION OF OPERATIONS

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Hours Critical

• and Enerov 16.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 5.69 kW-hrs

• This time includes that for loading fuel elements during the Approach to . Critical experiment and sub-critical time during the control rod calibration. The control rods are calibrated using sub-critical multiplication so inat a large portion of the experiment is done with the reactor sub-critical. The actual critical time was about 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or 24 % of the above time.

lit. EMERGENCY SHUTDOWNS AND INADVERTENT SCRAMS There were no inadvertent scrams or emergency stautdowns during this period.

IV. MAINTENANCE The maintenance activities this reporting period involved the high voltage power supplies for thi LF3 tube and the Compensated lon Chamber and the Poison Control Rod Drsve System. The high voltage supply for the RIOL Scaler was overhauled and returned to service. Several interrelated problems with this supply had forced the use of an external supply.

Once the AIDL high voltago supply was repaired, the external high voltage supply that had been used in glace of the RIDL supply was then installed for use with the Compensated lon Chamber. The hegh voltage for the Compensated lon Chamber had been supplied by battery, since its power a- r_ i nn fd1 0& L

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. supply had' failed' several years ago. The failed supply was vintage '1945 and i

- repair' parts for It: were,not' available any. more, AL similar vintage: power supply was'found to' beifunctional so it is now a back-up unit-for 2

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' Compensated'lon. Chamber. power. supply.

The.DC power supply for the stepper 1 motor on the Poison Control Rod was overhauled The stepper motor drive was found .to have an intermittent open-which caused the stepper motor- to chatter. This driver was replaced and another driver was ordered. '

1 The results of the overhaul of the LOPRA equ6pment were as expected, the replacement of capacitors and identification of poor solder joints produced smoother operation with much less noise.

l V. CONDIT10NS'UNDER SECTION 50.59 0F 10 CFR 50 There were no changes to procedures- or new experiments during this period. The present excess reactivity of the LOPRA core is-about 14 cents.

- This' excess reactivity would allow a maximum power of about 12 kilowatts.

VI., Vll... Vill., RADIOACTIVITY Because of the lower power and infrequent use of the LOPRA, its operation does not contribute to the release of effluents. Personnel radiation exposures for the laboratory are given in the Annual -Report for the Advanced TRIGA

. Reactor, License No. R-115, Docket No. 50-151, dated February 26, 1986.

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