ML20236J743
| ML20236J743 | |
| Person / Time | |
|---|---|
| Site: | Pennsylvania State University |
| Issue date: | 06/30/1987 |
| From: | PENNSYLVANIA STATE UNIV., UNIVERSITY PARK, PA |
| To: | |
| Shared Package | |
| ML20236J739 | List: |
| References | |
| NUDOCS 8711060262 | |
| Download: ML20236J743 (7) | |
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PENN STATE BREA2EALE REACTOR 7 ' ANNUAL OPERATING REPORT, FY 86-87 PSBR Technical ~ Specifications 6.6.1 i License CDW R-2, Docket No. ' 50-5 .
Reactor Utilization ,
The Penn State Breazeale Reactor (PSBR) is a TRIGA Mark III facility capable of 1 MW steady state operation, and 2000 MW peak power pulsing operation. Utilization of the reactor and its associated facilities falls j into three major categories:
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EDUCATION' utilization is primarily in the _ form of laboratory !
classes conducted for graduate, undergraduate, associate degree f candidates,.and numerous high>achool science groups. These classes vary from neutron activation analysis of an unknown sample to the calibration e
of a reactor control rod. In aedition, an average of 2000 visitors tour the PSBR facility each year. '
RESEARCH accounts for a large portion of reactor time which l involves Radionuclear Applications, Neutron Radiography, a myriad of research programs by faculty and graduate students throughout the University, and various applications by the industrial sector.-
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TRAINING programs for Reactor Operators and Reactor Supervisors are .j continuously offered and are tailored to meet the needs of the participants. Individuals taking part in these programs fall into such categories as power plant operating personnel, graduate students, and foreign trainees.
The PSBR- facility operates on an 8 AM - 5 PM shift, five days a week, with an occasional 8 AM - 8 PM or 8 AM - 12 Midnight shif t to accommodate reactor operator training programs or research projects. I l
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'8711060262 871026 l gDR ADOCK 05000005 1 PDR 1 L
t Summary of Reactor Operating Experience -
Technical specification requirement 6.6.1.a.
Between July 1,1986, and June 30, 198.7, the PSBR was l critical for 533 hrs or 2.2 hrs / shift suberitical for
- l'95 hva or 1.9 hrs / shift used while shutdown for 532 hrs er 2.1 hrs / shift net available for 39 hrs or 0.2 hrs / shift Total Usage 1600 hrs or 6.4 hrs / shift 547 hrs of subcritical time involved fuel movement The reactor was pulsed a total of 166 times with toe following reactivities:
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less than $2.00 74 j l
$2.00 to.$2.50
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84 greater than $2.50 8
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The square wave mode of operation was used 85 times to power levels between 100 and 500 KW.
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Total energy produced during this report period was 245 MWH with a consumption of 13 gms of U-235.
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Unscheduled Shutdowns !
Technical specification requirement 6.6.1.b.
There were 18 unplanned scrams during this period. Power range switching j errors by students or industrial trainees account for 7 of the 18 scrams. The i i
r other scrams are described below.
July 3,1986 -- Manual scram by operator as per procedure when a pneumatic transfer system I capsule returned to the laboratory terminus broken. J With health physics staff assistance, the missing parts of the capsule were retrieved. The polyethylene encapsulated sample which travels in the capsule !
was also retrieved intact, i
July 24, 1986 -- A manual scram was initiated when a string used to support a central thimble oscillating sample slipped from a hook on the oscillator mechanism. The sample was retrieved intact from the central thimble and irradiation resumed. 4
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1 December 8,1986.-- A manual scram was initiated when the conscle recorder pens stopped moving during a startup. The Pen / Record switch was found to be in a neutral position which deactivated the pen drives. The switch lever was removed to prevent a recurrence of the event due to an operator inadvertently bumping the lever.
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February 13, 1987 -- Period scram with the reactor suberitical in the standby position when the operator turned the period test knob to the I bi-stable test position to defeat the period scram during fuel movement. A previous operator had not returned the period test control potentiometer to l
its counterclockwise position .following the daily reactor checkout. 1 February 20, 1987 -- Manual scram by operator as per procedure when an I experimenter reported that a pneumatic transfer . system I capsule failed to return to the laboratory terminus. The capsule did then return intact; the slower than normal capsule travel time was attributed to an askew capsule threaded cap.
March 3, 1987 -- Manual scram by operator as per procedure ~'.ien an experimenter reported that a pneumatic transfer system I capsule failed to return to the laboratory terminus. The capsule did then return intact but slower than normal. Imprecise t'ubing connections suspected; critical tubing connections remade.
March 3,1987 -- External scram occurred when the wrong evacuation alarm defeat switch was activated while interchanging the beam hole laboratory and cobalt-60 facility area monitors' reactor console readout modules during monitor calibrations.
March 5,1987 -- Manual scram by operator as per procedure when a radiation monitor on pneumatic transfer system I alarmed. With health physics assistance, the capsule was returned to the laboratory intact. Since the I
alarm cleared within ten seconds and there was no release of radioactivity L from the sample, the alarm was thougnt to be an electronic noise spike.
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March 5, 1987 -- Manual scram by operator as per procedure when an experimenter reported that a pneumatic transfer system I capsule failed to !
return to the laboratory terminus. Capsule found intact in curved aluminum !
l section of system. Cap was crossthreaded and askew, causing capsule to wedge in tubing. Also, the aluminum section was inspected and found to be somewhat flattened's this section was apparently damaged at some time. The curved aluminum section was replaced with poly tubing with an inside diameter slightly larger than the aluminum tubing. i' April 10,1987 -- External reactor scram and building evacuation alarm initiated by reactor bridge east area nonitor. Operator failed to activate Nitrogen-16 diffuser pump as per procedure when exceeding 200 KW; intended power for this startup was 900 KW. Building evacuation and re-entry.as per procedure.
June 19, 1987 -- Reactor linear scram when the neutron source was removed from the reactor core to check for criticality during a student laboratory.
The source was inadvertently moved too near to the linear detector.
Major Maintenance With Safety Significance Technical specification requirement 6.6.1.c.
No major preventative or corrective maintenance operations with safety significance have been performed during this report period.
Major Changes Reportable Under 10 CFR 50.59 .
Technical specifications requirement 6.6.1.d.
Facility Changes July 29, 1986 -- Fuel element thermocouple that previously read to the l 1
console recorder red pon in pulse mode only, was reconnected to read only to a j 1
new green pen on the recorder in all modes of operation. J J
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August 5,1986 -- New dual high voltage power supply systems installed .
for the core radiation detectors. The gamma chamber and fission chamber are supplied with 700 volts and 320 volts, respectively, by a common power supply.
The log N chamber and linear chamber are supplied with 800 volts by a second i
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. common power supply. When tha voltage supplied by either power supply.
decreases by 0-20 volts, a reactor scram occurs accompanied by a console panel
'HIGH VOLTAGE SCRAM indication.
September 8, :986 -- A drain line discharge from the radioactive waste evaporator was changed from a storm sewer to a sanitary sewer to conform to new state waste disposal permit regulations.
t October 23, 1986 -- Reactor Pool Level High indicator installed with an alarm to the console annunciator panel. Police Services also receives thib alarm during off-duty hours when the Intrusion Alarm System is activated.
January 29, 1987 -- Pneumatically operated valves were installed in the primary side of the-heat exchanger system. These valves close when the primary pump is off to isolate the pool from the heat exchanger system. d i
March 26, 1987 -- A console recorder fuel temperature channel check circuit was added for the green pen indication, i
March 26, 1987 -- A vacuum breaker was installed in the secondary side of the heat, exchanger to prevent water hammer when the system is turned off. l April 27, 1987 -- A steel expansion tank was installed in the section of f
the primary side of the heat exchanger that is isolated from the pool when the primary side is turned off. Because of temperature changes in the isolated f
primary section, the expansion tank is needed to prevent meaningless f
differential pressure alarms when the system is turned off.
i May 14, 1987 -- The original General Atomic reactor console !
s electro-mechanical rod position indicators were replaced with new Veeder-Root l
digital indicators. the existing eending potentiometers on the control rod {
drives were replaced by optical Veeder-Root sending units on the transient, safety, and shim rod drives. An optical Veeder-Root sending unit was also added to the regulating red but the original sending potentiometer was also retained to provide rod position indication to the auto control s; stem. On May i
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'20, 1987, this new rod position indicating system was connected to an uninterruptable power supply.
l f Radioactive Effluents Released Technical specifications requirement 6.6.1.e.
Liquid ]
There were no liquid effluent releases under the reactor license for the report period. Liquid from the regeneration of the i reactor domineralizer was evaporated and the distillate recycled for i
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pool water makeup. The evaporator concentrate was dried and the.
solid salt residue was disposed of in the same manner as other solid radioactive waste at the University. Liquid radioactive waste from
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the radioisotope laboratories at the PSBR is under the University
- I byproduct materials license and is transferred to the Health Physics Office for disposal with the waste from other campus laboratories.
Liquid waste. disposal techniques include storage for decay, release to the sanitary sewer as per 10 CFR 20, and solidification for shipment to licensed disposal sites.
Gaseous The only gaseous effluent is Ar-41, which is released from dissolved air in the reactor pool water, dry irradiation tubes, and air leakage from the pneumatic sample transfer systems.
The amount of .Ar-41 released from the reactor pool is very dependent upon the operating power level and the length of time at power. The release per MWH is highest for extended high power runs j and lowest for intermittent low power runs. The concentration of )
Ar-41 in ths reactor bay and the bay exhaust was measured by the Health Physics staff during the summer of 1986. Measurements were made for conditions of low and high power runs simulating typical }
opeeating cycles.
Based on tnese measurements an annual release of between 142 mci and 430 mci of Ar-41 was calculated, resulting in an average concentration at the building exhaust between 9% and 26% of !
the mPC for unrestricted areas. These values represe.it the extremes, with the actual release being between the two values. The maximum fenceline dose using only dilution by the 1 m/s wind into
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l the lee of the building is on the order of 0.1% to 0 3% of the unrestricted. area MPC.
During the report period, bnly one dry irradiation tube was used at high enough power levels and for long enough runs to produce any significant amount of Ar-41. The calculated annual production was 170 mC1. Since this production occurred in a stagnant volume of
, air confined by close fitting shield plugs, most of the Ar-41 decayed in place before being released to the reactor bay. The reported releases from dissolved air in the reactor pool are based on measurements made, in part, when a dry irradiation tube was in use at high power levels; the Ar-41 releases from the tubes are part of rather than in addition to the release figures quoted in the '
previous paragraph.
The use of the pneumatic transfer systems was minimal during this period and any Ar-41 releases would be insignificant. i l
4 Environmental Surveys l Technical specifications requirement 6.6.1.f.
The caly environmental survey performed was the routine TLD gamma-ray dose measurement at the facility renceline and at several control points in residential areas several miles away. The 1986-1987 measurements were consistent with past measurements at the same points and indicated an annual l l dose equivalent of about 8h mrem at the fenceline, and about 68 mrem at the 1
control points. The difference can be largely attributed to a higher concentration of K-ho in the soil at the PSSR site.
Personnel Exoosures Technical specifications requirement 6.6.1.g No reactor personnel or visitors received dose equivalents in excess of 25% of the permissible limits under 10 CFR 20.
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