ML20012A335
| ML20012A335 | |
| Person / Time | |
|---|---|
| Site: | University of Illinois |
| Issue date: | 12/31/1989 |
| From: | Miley G, Pohlod C, Williams J ILLINOIS, UNIV. OF, URBANA, IL |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9003090268 | |
| Download: ML20012A335 (9) | |
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, University of Illinois
.D'Partment of C2Hege d Engineering Nuc!:ar Engl:eering v.
- at Urbana-Charnpaign 214 Nuclear Engineering -
217 333-2295 Laboratory -
103 South Goodwin Avenue Urbana, IL 618012984 j
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February 28, 1990
. Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 1 White Flint North 11555 Rockville Pike
- Rockville, Md 20B52 4
= Attention Document Control Desk
Dear Sir:
SUBJECT '
ANNUAL REPORT: Illinois Advanced TRIGA Reactor License No. R-115 Docket No. 50-151
.The following is written. to comply with the requirements of Section-6.7.f.
of the Technical Specifications and the conditions of Section 50.59 of '
10 CFD.
The. outline of the report follows the numbered sequence of section.
'6.7.f of the. Technical Specifications.
Yours truly, Craig S."Pohlod Reactor Supervisor hl John G. Wi11iams Reactor Laboratory Director i
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deorgef.Miley,Chapr%n 7Nuclear Reactor-Committee W. ' 1/
DarclayG. jones lead Department 7of N, lear Engineering cc: Regional Administrator, Region III',
- 9003090268 891231
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i di ANNUAL REPORT JANUARY 1, 1909-DECEMDER 31,1989 ILLINDIS ADVANCED TRIGA FACILITY LICENSE R-ll5
SUMMARY
OF OPERATING EXPERIENCE A. Summary of Usace t
The reactor was scheduled for use 34.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> per week and was in operation 22.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> per week. Scheduled time increased about 15% from last year, and actual operational hours increased about 20%. These increases are due to increased research in nuclear pumped lasers, the investigation of radiation effects on optical materials and neutron activation work involving air pollution. Air filter irradiations are usually long irradiations which preclude scheduling several activities during the same time period.
In the following table, the per cent of time for different activities is listed.
Scheduled time is that time reserved for a given operation and it includes scheduling more than one reactor facility for use at the same. time while operating time is from start-up to shutdown for all the scheduled activities.
CATEGORY SCHEDULED OPERATING Research Projects 22.8%
24.0%
frradiations (samples) 48.4%
54.4%
Education and Training 20.3%
- 17. 5Y.
Maintenance and Measurements 8.5%
4. l Y.
Presently there is one individual with a Senior Operator License and three individuals with an Operators License. The facility operates with a 40
. hour week schedule, a staff of one and three quarters full time equivalent operators and one full time reactor health physicist.
One individual is preparing to take the Senior Reactor Operator (SRO) examination and one individual is preparing to take the Reactor Operator examination. It is expected that these individuals will be examined during the early summer of 1990.
B. Performance Characteristics 14 Fuel Element Lenoth and Diameter Measurements These checks were made on the B & C Hexagonals during the month of March. The pulse number at the time of the checks was 9348. For the eighteen elements in this region, there was an average increase in the length of about 8 mils. This increase is to be expected as six of these elements are relatively new, having been added to the core last year. The accuracy of a given measurement is estimated at i 5 mits. There was no change in the diameter of the fuel elements checked.
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- These checks.were made on' the D-G Hexagonals~ during the month of ~
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-September. ;The-pulse-number at the time of these checks was 9693. For the-93~
elements in this region,.there was an average decrease in length of 16 mils.
This decrease is thought to be due;to annealing associated'with long, high
-power runs. The~ accuracy of a'given ' measurement is as stated above. There was no change in the. diameter of,the fuel elements -checked.
L There were 781 pulses in 1989, bringing the total since 1969-to 9962.
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= The: values =for pulse height, reactor period and fuel temperature were the same as measured in previous. years.
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Reactivity-y Control Rods: The measured reactivity values of the control rods have shown essentially no change. Variations between successive measurements are seldom greater than 5'4.
Core Reactivity: The net loss of reactivity attributed to fuel burnup during the year was $ 0.47. This value was determined by a comparison <
of the: cold critical xenon-free control rod position at the beginning and-at the end of the_ year.
Based on an estimated-2 (10,5) cents per MW-day of E
' operation, the reactivity loss for the year would have been approximately 50.43.
II. 1TABULAT10N OF ENERGY AND PULSING 1
A. Hours Critica1 and Enerov Tvoe of Operation Time (hrs)
Eneroy (MW-hrs) 0-10 kW 289.9 0.05 10kW-250kW 105.3 13.89 250kW-1.5MW 525.1 491.55 Pulsing 250.4 4;65 Total 1,170.7 510.14 B.
Pulsino Pulse Size Number
$1.00-1.70 6
1.71-2.00~
27 2.01-2.30 0
2.31-2.80 25 2.81-3.19 723 Above $3.19 0
= Total-781 1Because of the-type of operation, the Hours Critical time includes j
instances where the reactor is not critical in the normal sense. These include
- l the. time'to get critical during a start-up, the time between pulses during j
continuous pulsed operation and-short periods of time during sample irradiations when samples may be removed or added.
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'!![. REACTOR SCRAMS-F There'were 43' unplanned scrams and no emergency shutdowns during this time period.
.hese scrams were attributed to Instrument Malfunction- (20),
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Operator / Operator Trainee Error (19) and External Causes (4). This is about-1 average for the facility over the years.
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Linear Power (14)
This is a power level scram required'by the Technical Specifications. It occurs when the signal on any power range exceeds about 1087.'of that range.
Eight (8) of these scrams were due to electronic noise problems. In four cases
'the-noise was generated when the Mode Switch was moved-from the Automatic F
position-to the -Steady: State position. In the other four cases the. scram occurred' at low power. These type of noise scrams are usually caused by capacitor-problems in the High Voltage Power Supply to the neutron chambers.
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The Mode. Switch :is cleaned periodically and old capacitors, usually electrolytic are replaced as they are identified as failed or failing.
Six (6) scrams occurred due to an operator or-operator trainee turning the range switch the wrong direction or switching to the Automatic Mode at too
. low a power level, t
Period Scram (16)
This scram is not required by the Technical Specifications. It occurs when-the period is 3 seconds or less with the Mode Selector Switch in Automatic or Steady State position. Four (4) of these scrams occurred either; when the period circuit ~was placed in operation at too low a power level or when.the true power level'was masked by high gamma current and therefore the
_ period limiter-circuit could not effectively drive the Regulating Rod in_to prevent-the Period Scram. Adjustment of the-Log-N Channel is not practical p
_after every shutdown. -This scram is usually caused by operator trainees or student operators who are not yet familiar with this behavior.
Twelve (12) of these scrams were due to circuit noise of an undetermined origin. The High Voltage Power Supply is suspected as the cause of these scrams. The problem appears to be intermittent.
Loss of Power (4)
Three (3) of these loss of power scrams-were caused by momentary interruptions;of electrical power to the building due to an electrical storms.
h The fourth resulted from an operator accidentally de-energizing the Control Panel while cleaning around the On/Off button for the Control Ponel.
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4 Primary Flow (8)
This scram is not required by Technical. Specifications. The scram occurs if power level exceeds-1.0 MW without adequate coolant flow (approx. 550 gpm)
-in. the primary coolant loop. This scram will also occur when adequate g
L secondary cooling :(opprox. 900 gpm) is not available and power is greater than 1.0 MW.-Four (4) scrams occurred due to the operator / operator trainee failing to establish Primary Coolant flow before increasing power level above 1 MW.
Three (3) scrams occurred due to difficulties in maintaining secondary flow while starting up the secondary system in the winter line-up. The cooling
. tower'is drained over night and it is not refilled until it is needed. Start up while the tower is filling sometimes leads to the introduction of air slugs into the coolant loop.
An air slug in the secondary coolant system will allow the secondary flow switch to trip the secondary pump which in turn trips the
' primary pump and this will cause a scram if the power level is greater.than
.l.0 MW.
One (1) scram occurred when the Primary / Secondary Cooling System Control E
Panel was accidentally de-energized. The operator reached for.the switch for the Isolation Valves without looking and turned off the panel.
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Hiah Water Level (1) l 1
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-i This scram occurred because the Bulk Shielding Tank, which is filled from the TRIGA Tank, was filled too fast-and the make-up system over-filled the TRIGA Tank in response. The over-fill left the water level in the TRIGA Tank an inch higher-than normal and several inches below the top of the TRIGA lank.
i IV. Maintenance It is estimated that about 480 hours0.00556 days <br />0.133 hours <br />7.936508e-4 weeks <br />1.8264e-4 months <br /> (40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per month) were. spent on maintenance-related activities. Only 150 of these -hours are reflected in the
-Summary of Operaticns.
These hours account for time when normally scheduled activities could not be carried out due to the need to make necessary repairs to: the reactor system and for that time when the reactor was needed to _ perform surveillance activities. The significant item of maintenance is given below.
L Primarv isolation Valves: The inlet isolation Valve to the TRIGA Tank failed. This failure was identified by analyzing the suction and discharge
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pressures. of the Primary System Pump and observing the operation of the pump-just before full flow was established.
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Observed indication of the suction and discharge gauges showed-suction pressure
' dropping' about 10-15 inches of' water before full flow was established, while the discharge pressure initially dropped below normal full J
flow pressure and then returned to normal full flow pressure as flow-came up-1to the normal full flow (1500 gpm) value.
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. of the' inlet Primary Isolation Valve required isolating'the Replacement inlet to-the TRIGAD Tank with the Manual Primary Isolation Valve, draining f
about 320 gallons from the Primary System, isolating the Primary. Heat Exchanger and removing the failed valve.
p The failed valve was a butterfly valve. It' exhibited excessive wear on the upper support surfaces, the corners had become rounded due to twenty years use. This valve was replaced with a new valveaof the same-type. The TRIGA was not operated during the time this valve was being replaced.
V.
Conditions Under Section 50.59 of 10 CFR There were three 50.59 reviews carried out during 1989. The addition of a' cadmium lined pneumatic transfer tube to be installed-in the NAT/CLNAT core position,.the addition of an automatic valve actuator to the Secondary Throttling Valve and the replacement of the stainless steel Fast Transient Rod (FTR) Guide Tube with a segmented aluminum Guide Tube were all reviewed under the provisions of 10CFR50.59.
The results of these reviews are presented.
below.
Cadmium Lined Pneumatic Transfer Tube (CLPS):
This cadmium lined-pneumatic -transfer tube was designed to be installed in the TRIGA-core position which accommodates the Neutron Activation Tube / Cadmium Lined Neutron Activation Tube (NAT/CLNAT) which was subject to 50.59 review in 1988. This tube allows for the irradiation of Neutron Activation Analysis samples by epithermal and fast neutrons.
Internally it is identical to the pneumatic transfer tube that has been used in the TRIGA since initial start-up. Externally the tube is sized to fit in.the NAT/CLNAT Support Plate. Reactivity measurements on the Cadmium Lined Pneumatic Transfer Tube revealed that it is worth less than the CLNAT (CLNAT-$
0.41 verses $ 0.33 for the CLPS).
Review of the CLPS reveals that'the use of the CLPS does not involve a-change to the facility Technical Specifications ar e..e reactivity worth of the CLPS is substantially less than the limits on in co<e facilities and no other Technical Specification item addresses the use of in core facilities.
Use of.the CLPS does not involve an unreviewed safety question as the CLPS does not increase the probability of occurrence or the consequences of an accident or~ malfunction of. equipment important to safety previously evaluated 3
in the safety analysis report. The.CLPS is internally identical to the.
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. pneumatic systems previously evaluated in the facility safety analysis report.:
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Externally. the CLPS is identical to the CLNAT, which has been previously l
reviewed.
I Use of thel CLPS does not introduce the possibility for an accident or a l
malfunc t ion of a different type than previously analyzed in the facility
-l safety analysis report because of its similarity to the original pneumatic transfer system and the CLNAT which have been previously evaluated.
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Use of the CLPS does not reduce any safety margin as defined in the basis for any technical specification. The reactivity worth of any single experiment is limited to $ 3.00 and the limit at which an experiment is required to be locked in place in the core is $ 1.00. The reactivity worth of the CLPS is S 0.33. The reactivity change caused by melting of the cadmium is far below the $ 4.60 limit placed on pulsing in the TRIGA.
FTR Guide Tube This segmented, aluminum guide tube is identical in
' dimension to the stainless steel tube it replaces. The original tube was heavy and difficult to remove itom the core due to the lack of overhead space. It is felt that the failures experienced with the FTR over the years are due to the difficulties in removal of this heavy guide tube. The details of this problem are discussed in an Information Letter to Region 111, USNRC dated Jarcary 3, 1989.
Conversations with General Atomics revealed the University of Illinois TRIGA is the only TRIGA in the country with a stainless steel Guide Tube for the FTR. Further conversation revealed that no one at General Atomics could explain this selection of material. General Atomics personnel could offer no reason for not replacing the stainless steel guide tube with a segmented aluminum guide tube.
This guide tube is not described in the facility safety analysis report.
The installation and use of this guide tube does not involve a change to the facility technical specifications as there is no mention of the guide tube in the technical specifications. Use of the segmented, aluminum guide tube does not involve. an unreviewed safety question. Safety should be enhanced as failure of the FTR should be reduced.
Secondary Throttle Valve Actuator: The addition of a small electric motor and associated positioning and indicating circuitry used in conjunction with the existing flow meter allows the operator to establish and adjust flow in the Secondary Cooling System from the Control Room. This motor can be quickly disconnected from the Secondary Throttle Valve to allow manual operation of the valve.
Position indication of the Secondary Throttle Valve is provided by an optically coupled linear system, with digital read out in the Control Room.
The Secondary Throttle Valve is-described generically in the facility safety analysis report.as a throttling valve with no mention of how it is operated. The Secondary Throttling Valve is not described or addressed in the facility technical specifications for operation. The TRIGA is designed to operate without Primary or Secondary Cooling to a power level of 1.0 MW. Loss of Coolant in the Secondary Coolant System will scram the-reactor if power level is above 1.0 MW at the time Secondary Coolant is lost. The reactor will scram if the operator attempts to increase power to greater than 1.0 MW without Secondary Flow.
Use of the motor actuator for operation of the Secondary Throttling Valve does not involve a change to the f acility technical specifications nor does it involve an unreviewed safety question.
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VI.. Release of Radioactive Materials
- Argon-41
Average concentration to environs via exhaust =:
.9.7 E-8 uti/ml-Total release = 3003 mci Monthly. range = 103-766 mci 9
Tritium -
Estimation of 1.0 mci release from the evaporation of water' in the reactor tank. This is based on the measured-concentration of H-3 and water usage for the year.
Effluent (sanitary sewer): Less than -10 uCi of beta-gamma emitting' material.
l Vll. Environmental Surveys
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There were no environmental surveys performed in' 1989 other - than rou' tine -
-j radiological-monitoring.
Contamination surveys are performed -in the Laboratory. See Section Vill.
Vill. Personnel Radiation Exposure and Surveys Within the Facility A. Personnel Exposure Twenty-two persons wer e assigned _ film badges at the f acility. Three.
were full time employees, while the others averaged less than twenty 1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per week in the. facility. The badges are sent to R.S.
Landauer-of. Glenwood
. Illinois 'for processin9 The table below gives the-whole--body deep dose
. equivalent' received by'those who were assigned.. film badges duringL1989.
Dose Eauivalent (REMS)
Number of Individuals No.Neasurable Exposure 0
0.01-0.10 15 -
O.10-0.25-5
-Above 0.25
_2_
Total-22
.g The -highest individual dose equivalent was 400' millirems.
This was received by an experimenter. whose research involves handling an experimental Lapparatus which';is-inserted. and -removed from-a beam port. Five other individuals received'a dose equivalent above 100 millirems. They received this.
dose. equivalent as 'a result of handling radioisotopes and/or special experimental apparatus. Students and visitors doses are recorded on self-reading dosimeters and1were less than 10 millirems.
'B. Contamination Surveys 10.-
Smear samples from various locations around the laboratory are taken at periodic intervals. The removable contamination is determined by counting the smears with a gas flow proportional counter.
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,The maximum gross beta _ contamination is.usually found in the vicinity
'.where the irradiated-sample. containers are handled. There were 4,376 samples-irradiated _during the year. In_ the sample-area the contamination varied from F
'approximately 100-25,000 dpm/ LOO; cm" or 4.5E-07 to. 1.lE-4'uCi/cm".
In the g
control room area' the maximum was 200 dpm/100 cm" or 9.0E-07 uCi/cm". Smears L
_from other. areas of the laboratory showed a maximum of 2,000 dpm/100 cm" or
'B.7E-06 uCi/cm". Clean up of these areas always results in levels of less than 1000 _dpm/100 ce".or 4.5E-6 uCl/cm". Surveys for alpha contamination were less p.
y than 10dpm/lOOcm" or 4.5E-8 uCi/cm".
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~IX'-
Nuclear Reactor Committee D r'. George -H.
Miley continues as Chairman-of-the Nuclear Reactor E
Committee for the 1980-1990 term. Dr.
Micklich has lef t the University of Illinois. 'He has-been replaced by Dr.
Abderrafi Ougouag. Dr.
Dugouag is an l Assistant Professor of Nuclear. Engineering who has served on the Reactor' Committee before.
Dr. Miley is a Professor'of Nuclear Engineering.-He is the former Chairman of.the Nuclear Engineering Program from the mid 1970's through the mid~1980's and has served on the Nuclear Reactor Committee previously.-The following_ members remain on the Nuclear Reactor Committeet Dr.
John G.
Williams, Associate Professor of Nuclear Engineering; Dr. Sheldon Landsberger, Assistant Professor of Nuclear Engineering). Mr.
Hector Mandel,
~ Campus Radiation Safety Officer; Mr. Craig Pohlod, Reactor Supervisor (ex-officio);
Mr.fNeil Barss, Reactor Health Physicist (ex-officio). All are previous members of this committee.
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