ML20207D613
| ML20207D613 | |
| Person / Time | |
|---|---|
| Site: | Pennsylvania State University |
| Issue date: | 06/30/1986 |
| From: | PENNSYLVANIA STATE UNIV., UNIVERSITY PARK, PA |
| To: | |
| Shared Package | |
| ML20207D547 | List: |
| References | |
| NUDOCS 8612310116 | |
| Download: ML20207D613 (7) | |
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PENN STATE BREAZEALE REACTOR ANNUAL OPERATING REPORT, FY 85-86 PSBR Technical Specifications 6.6.1 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 into three major categories:
EDUCATION utilization is primarily in the form of laboratory classes conducted for graduate, undergraduate, associate degree candidates, and numerous high school science groups. These classes vary from neutron activation analysis of an unknown sample to the calibration of a reactor control rod. In addition, an average of 2000 visitors tour the PSBR facility each year.
RESEARCH accounts for a large portion of reactor time which 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.
TRAINING programs for Reactor Operators and Reactor Supervisors are continuously offered and are tailored to mcet the needs of the participants. Individuals taking part in thesa programs fall into such categories as power plant operating personnel, graduate students, and fcreign 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 shift to accommodate reactor operator training programs or research projects.
8612310116 861216 PDR ADOCK 05000005 R PDR
Summary of Reactor Operating Experience Technical specification requirement 6.6.1.a.
-Between July 1, 1985, and June 30, 1986, the PSBR was critical for- 471 hrs'or 1.9 hrs / shift suboritical for
- 500 hrs or 2.0 hrs / shift used while shutdown for 554 hrs or 2.2 hrs / shift not available for 122 hrs or 0.5 hrs / shift Total Usage 1647 hrs or 6.6 hrs / shift
- 112 hrs of subcritical time involved fuel movement The reactor was pulsed a total of 125 times with the following reactivities; less than $2.00 65
$2.00 to $2.50 57 greater than $2.50 3 The square wave mode of operation was used 103 times to power levels between 100-500 KW.
Total energy produced during this report period was 188 MHW with a consumption of 10 gms of U-235.
Unscheduled Shutdowns Technical specification requirement 6.6.1.b.
There were 17 unplanned scrams during this period. Of these, 14 were caused by personnel error, and 3 by system failure.
Power range switching errors by students or industrial trainees account for 13 of the 14 personnel errors. The other one occurred on August 23, 1985, when a wrench was dropped from the reactor bridge into the pool above the core.
A manual scram was initiated by the reactor operator in case the wrench would land on top of the core. It fell to the pool floor, away from the core.
The causes of the remaining 3 scrams follow:
Jan. 6, 1986 - a fuel temperature scram occurred with the reactor at standby, when the instrumented fuel element thermocouple connection was removed during a loading change. A procedure was developed to prevent future occurrences.
Apr. 16, 1986 - a linear scram (110% on the 300 kw range) occurred when the mode switch was turned to the Square Wave position and the auto mode activated prematurely due to a malfunctioning relay.
Although the condition could not be repeated, as a precaution all four relays affected by the Square Wave and Auto operations were replaced.
May 1, 1986 - a manual scram was initiated when a pneumatic transfer system capsule was mistakenly thought not to have returned from the core. The capsule, stuck at the receiving terminus air cushion ball valve, was retrieved intact. The malfunctioning ball valve was repaired.
Major Maintenance With Safety Significance Technical specification requirement 6.6.1.c No major preventative or corrective maintenance operaticna 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 -Several changes were made to the control conscle
instrumentation in order to comply with the new PSBR technical specifications as contained in the R-2 License renewed effective January 27, 1986. However, these were all covered in the safety analysis contained in the March 1, 1985, license renewal application. One additional. change occurred on May 9, 1986, when a new high voltage power' supply (HVPS) was installed to provide both a.
. voltage monitor visible to the reactor operator, and scram capabilities on the loss of, or reduction of voltage to the compensated ion chambers in the Linear and Log N channels and the gamma chamber in the percent power channel. This change does not involve an unreviewed safety question, but rather provides the operator with a means of continuously monitoring the chamber voltage and of checking the scram capabilities.
Procedures - All procedures were reviewed and revised as needed to comply with the new technical specifications. Since the changes were numerous, no attempt will be made to list them. However, a current copy of all facility procedures will be made available on request. ;
New Tests and Experiments - There were no new tests or experiments performed that were not within the scope of those described in the prevailing safety analysis report.
s 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 reactor demineralizer was evaporated and the distillate recycled for 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 the radioisotope laboratories at the PSBR is under the University 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 and lowest for intermittent low power runs. The concentration of Ar-41 in the 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 operating cycles. Based on these 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 betucen 9% and 26% of the mPC for unrestricted areas. These values represent the extremes, with the actual release being between the two values. The maximum fenceline dose using only dilution by a 1 m/s wind into the lee of the building is on the order of 0.1% to 0.3% of the unrestricted area MPC.
i During the report period, only two dry irradiation tubes were used at high enough power levels and for long enough runs to produce any significant amount of Ar-41. The calculated annual production was 276 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 levele; 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.
Environmental Surveys Technical specifications requirements 6.6.1.f.
The only environmental survey performed was the routine TLD gamma-ray dose measurement at the facility fenceline and at several control points in residential areas several ciles away. The 1985-1986 measurements were consistent with past measurements at the same points and indicated an annual dose equivalent of about 83 mrem at the fenceline, and about 69 mrem at the control points. The difference can be largely attributed to a higher concentration of K-40 in the soil at the PSBR site.
Personnel Exposures 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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