ML20197J477
| ML20197J477 | |
| Person / Time | |
|---|---|
| Site: | Pennsylvania State University |
| Issue date: | 12/04/1998 |
| From: | Sears C PENNSYLVANIA STATE UNIV., UNIVERSITY PARK, PA |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20197J481 | List: |
| References | |
| NUDOCS 9812150102 | |
| Download: ML20197J477 (7) | |
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PENNSTATE Radiation Science and Engineering Center (814) 865-6351 College of Engineering The Pennsylvania State Unisersity Breazeale Nuclear Reactor Building University Park. PA 16802 2301 Annual Operating Report,FY 97-98 PSBR Technical Specifications 6.6.1 License R-2, Docket No. 50-5 December 4,1998 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D. C. 20555
Dear Sir:
Enclosed please find the Annual Operating Report for the Penn State Breazeale Reactor (PSBR). This report covers the period from July 1,1997 through June 30,1998, as required by technical specifications requirement 6.6.1. Also included are any changes applicable to 10 CFR 50.59.
A copy of the Fony-Third Annual Progress Report of the Penn State Radiation Science and Engineering Center is included as supplementary information.
Sincerely yours, O-C. Frederick Sears Director, Radiation Science and Engineering Center Enclosures
/
cc. R. A. Erickson D. N. Wormley L. C. Burton A. J. Baratta E. J. Boeldt e M. Mendonca .
T. Dragoun _fl" 9812150102 981204 PDR ADOCK 05000005 R PDR I
ColleFe of Engineering An Fyual Opportunity University
PENN STATE BREAZEALE REACTOR Annual Operating Report, FY 97-98 PSBR Technical Specifications 6.6.1 License 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 two major categories:
EDUCATION utilization is primarily in the form of laboratory classes conducted for graduate and undergraduate students 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/ SERVICE 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.
The PSBR facility operates on an 8 AM - 5 PM shift, five days a week, with = occasional 8 AM - 8 PM or 8 AM - 12 Midnight shift to acconunodate laboratory cocrses or research/ service projects.
Summary of Reactor Operating Experience - Tech Specs requirement 6.6.1.a.
Between July 1,1997 and June 30,1998, the PSBR was critical for 755 hours0.00874 days <br />0.21 hours <br />0.00125 weeks <br />2.872775e-4 months <br /> or 2.6 hrs / shift suberitical for 517 hours0.00598 days <br />0.144 hours <br />8.54828e-4 weeks <br />1.967185e-4 months <br /> or 1.8 hrs / shift used while shutdown for 581 hours0.00672 days <br />0.161 hours <br />9.606481e-4 weeks <br />2.210705e-4 months <br /> or 2.0 hrs / shift not available 109 hours0.00126 days <br />0.0303 hours <br />1.802249e-4 weeks <br />4.14745e-5 months <br /> or 0.4 hrs / shift Total usage 1962 hours0.0227 days <br />0.545 hours <br />0.00324 weeks <br />7.46541e-4 months <br /> or 6.8 hrs / shift The reactor was pulsed a total of 56 times with the following reactivities:
< $2.00 16
$2.00 to $2.50 39
> $2.50 1
>= $3.C0 0 The square wave mode of operation was used 40 times to power levels between 100 and 500 KW.
Total energy produced during this report period was 327 MWH with a consumption of 17 grams of U-235.
l l
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PSBR Annual Operating Report, FY 97-98 l l
Unscheduled Shutdowns - Tech Specs requirement 6.6.1.b.
The 6 unplanned shutdowns during the July 1,1997 to June 30,1998 period are described below.
September 16,1997 - DCC-X Power Range (Fuel Temperature) Trip i Reactor trip from 50 watts. Fuel 2 thermocouple was accidently disconnected while setting up an I experiment. The disconnect produced a full scale reading on DCC-X as designed. Reactor tripped on DCC-X Fuel Temp 2 High. The RSS hard-wired system did not trip since Fuel Temp 2 was not selected. All systems functioned as designed. The experiment was suspended until corrective actions and training were completed. Thermocouple wiring was upgraded as well as color coded and a permanent connection for the spare thermocouple was installed. Training was j conducted regarding the event; emphasis was placed on separation of operations and the I experimenter. l January 14,1998 - DCC-X Power Range (Gamma) Trip j The reactor was being operated at 750 kw in automatic mode for a 38 minute irradiation. The l power range channel indication fed by a gamma ionization chamber gradually increased over the trradiation period to an indicated nominal value of 795 kw with a plus or minus 5 kw variation at 37-1/2 minutes. The gamma signal buildup was gradual and was not noted by the operator until just before the trip. DCC-X Power Range tripped at 800 kw (the current setpoint) as designed.
Actual reactor power was maintained prior to the trip at 750 kw as indicated by both the wide range (neutron driven) and the fuel temperatures. The basic cause of the trip was a mis-match in I calibration between the neutron channel and gamma channel for long term operation which did l not account for normal build-up of the gamma signal. The overpower setpoints had recently been l lowered from 110% as a conservative action while some calibration questions were being I addn:ssed. The power range (gamma) was properly calibrated to the wide range (neutron) with account taken for build-up of delayed gamma. Training was conducted on expected behavior of the various indications of reactor power and on operator alertness.
January 21,1998 - Manual Trip The reactor was being operated for sample irradiations utilizing the rabbit sy:; tem. A spurious rabbit system radiation alarm was received; per procedure the operator tripped the reactor. The SRO was observing the radiation monitor at the time and was able to confirm the alarm was an instantaneous spike which immediately cleared before the trip. Surveys were conducted to verify the alarm was spurious and the system was examined. The spike was believed to be due to component aging. Repairs to the monitor were made and the monitor will be upgraded / replaced in the future.
March 9,1998 - DCC-X Wide Range (Neutron) Trip A reactor trip occurrred from 800 kw. An operator trainee was balancing rods while operating at 750 kw. The transient rod had not been prepositioned and a withdrawal was made which could not be adequately compensated for by the regulating system due to the very low core excess reactivity existing at the time. DCC-X tripped at 800 kw. Over power setpoints had been set to a conservative value of 800 kw in December,1997. Staffincluding trainees were counseled as to proper prepositioning and rod balancing techniques.
April 28,1998 - Remote Manual Trip The reactor was being operated at 700 kw in the open pool. Work was in progress in the Neutron Beam Lab. A worker placed a box in front of the remote manual trip and pushed the box towards the wall depressing the trip button. Circular extensions were placed around all remote trip buttons to preclude similar accidental depression.
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PSBR Annual Operating Report, FY 97-98 June 3,1998 - Manual Trip The reactor was being operated at 500 kw to perfomi sample irradiations utilizing the rabbit system. A spurious rabbit system radiation alarm was received; per procedure the operator tripped the reactor. No sample was being irradiated at the time. The rabbit system radiation alarm system is being upgraded. See January 21,1998 trip.
Major Maintenance With Safety Significance - Tech Specs requirement 6.6.1.c. l 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 - Tech Specs requirement 6.6.1.d. l Facility Changes -
During the June 10 to June 13,1998 fuel inspection outage, a new Wide Range Channel fission detector located vertically in a new detector rack at the rear of the reactor core became the Tech Spec required linear power detector. The previously used fission detector of the same design was removed from its horizontal position above the back three fuel rows of fuel during the outage. i This change was made to allow unimpeded access to the fuel elements previously blocked by the detector's location, to allow a better neutronic view of the core by the detector for an improved linear response, and to minimize the impact on instrument response from power shifts due to l experiment interface at the front of the core. Prior to the fuel inspection outage, the new vertical detector was benchmarked against the old horizontal detector to verify its linearity. This change did not increase the probablility of occurrence or the consequences of an accident or malfunction of equipment important to safety as evaluated in the SAR.
During the June 10 to June 13,1998 fuel inspection outage, the plasma displays on the wide range and power range instrumentation drawers and the quad isolator printed circuit card in the power range drawer were replaced, and a zener diode package was installed in each drawer. The plasma displays were replaced with liquid crystal displays in both drawers. To incorporate this change,in each drawer: 1) the driver / bistable trip cards A1, A2, and A3 were replaced with new bistable trip cards, 2) the high voltage power supply that powered the plasma displays was removed, and 3) the display backplane which contains the plasma displays was replaced with a new backplane specifically designed with liquid crystal displays. The plasma displays and driver cards had been high maintenance items with periodic failures and the power supply and driver
- cards generated a significant amount of heat that could have affected the performance of the amplifiers in the drawers. The isolator replacement in the power range drawer required the removal of the isolator card (All) and the isolator power supply card (A12) and replacing the isolator card with a new quad isolator. The power supply for the new quad isolator is located in the isolator card. The isolator card is designed to eliminate any possible feedback between the instrumentation channels and or outside interference. A zener diode package was installed in both drawers in parallel with a capacitor bank between plant ground and instrumentation common. The zener diode package is designed to clamp DC voltage potentials between plant ground and instrumentation common. This will prevent protection circuits in the thermocouple amplifier from shunting fuel temperature indications. Additionally, a non-isolated signal (fc_ log _ power) that is an output for the Wide Range Drawer was removed. (This signal was an input to the Power Range Drawer. It was used in the Power Range Drawer as an integration signal during pulsing. From an operational standpoint this modification has no effect. It does prevent hardware integration of the pulse signal in the linear amplifier, A6. This signal was unused). The new quad isolator board replaced the original isolator card which was actually a buffer amplifier and didn't provide the isolation required. The new design is a true isolation amplifier with 1500 volt RMS minimum isolation protection. In summary, these changes removed from se1vice the high maintenance items associated with the plasma displays. The changes increase the reactor safety system (RSS) reliability and reduce the probability of system 3
PSBR Annual Operating Report, FY 97-98 failure. Replacing the quad isolator increases the RSS resistance to outside interference. All of the components being changed a functionally identical to those in the original RSS. Placing the zener diode trap between the. plant ground and the instrumentation common is a design change that Gamma-Metrics (original RSS manufacturer) recommended based on a failure identified on Dec.16,1996 when fuel temperature indications were accidentally shunted during a pulse. This modification increases the RSS resistance to external inteference. These modifications do not increase the probability of occurrence or consequences of an accident evaluated in the SAR.
March 19,1998 - The lever type microswitches on the safety rod were replaced with noncontact photoelectric switches. These switches provide feedback to the Reactor Safety System, the motor controllers, and the DCC-X control computer. The lever type switches had been troublesome, and required frequent adjustment or replacement. This change did not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as evaluated in the SAR.
June 24,1998 - A bypass valve was added around valve #81 in the heat exchanger primary system. This was done to provide fine now control during the performance of thermal power calibrations. This change had no safety significance.
Procedures -
Procedures are reviewed as a minimum biennially, and on an as needed basis. Changes during the year were numerous and no attempt will be made to list them.
New Tests and Exoeriments -
None Radioactive Efnuents Released - Tech Specs requiremot 6.6.1.e.
Liquid An unplanned release of approximately 2000 gallons of reactor pool water occurred over the period of August 2 to August 11,1998. The unplanned release was leakage from a storage tank located within the fenced boundary of the reactor facility. The leakage took place as a result of undetected corrosion in the bottom portion of the storage tank. The water from the south side of the reactor pool was being held in the storage tank following its transfer there on August 2; on August 7 the bulk of the water was returned to the reactor pool with a small portion of water remaining in the tank until August 11. The activity in the released water was 469 microcuries of tritium and 11 microcuries for the other radioisotopes present. A more detailed account of this release was detailed in an information letter to the NRC dated September 11,1998.
There were no planned liquid effluent releases under the reactor license for the report period.
Pool water used to flush pool drain lines during pool water transfer is evaporated and the distillate recycled for pool water makeup. The evaporator concentrate is dried and the solid salt residue disposed ofin the same way as other solid radioactive waste at the University.
Presently, the demineralizer beds are replaced when depleted. The depleted beds are solidified for shipment to licensed disposal sites.
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.
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PSBR Annual Operating Report, FY 97-98
! Gaseous l Gaseous effluent Ar-41 is released from dissolved air in the reactor pool water, air in dry
! irradiation tubes, and air leakage to and from the carbon-dioxide purged pneumatic sample transfer system. The amount of Ar-41 released from the reactor pool is very dependent upon the l operating power level and the length of time at power. The release per MWH is highest for l
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 oflow and high power runs simulating typical operating cycles. Based on these measurements, an annual release of between 249 mci and 753 mci of Ar-41 is calculated for July 1,1997 to June 30,1998, resulting in an I
average concentration at ground level outside the reactor building that is 0.4 % to 1.2 % of the effluent concentration limit in Appendix B to 10 CFR 20.1001 - 20.2402. The concentration at ground level is estimated using only dilution by a 1 m/s wind into the lee of the 200 m2 cross section of the reactor bay.
During the report period, several irradiation tubes were used at high enough power levels and for long enough runs to produce significant amounts of Ar-41. The calculated annual production was 643 mci. Since this production occurred in a stagnant volume of air confined by close fitting shield plugs, much 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; some of the Ar-41 releases from the tubes are part of rather than in addition to the release figures quoted in the previous paragraph. Even if all of the 643 mci were treated as a separate release, the percent of the Appendix B limit given in the previous paragraph would still be no more than 2.2 %. The use of the pneumatic transfer system was minimal during this period and any Ar-41 release would be insignificant since the system operates with CO-2 as the fill gas.
Tritium release from the reactor pool is another gaseous release. The evaporation rate of the reactor pool was checked by measuring the loss of water from a flat plastic dish floating in the pool. The dish had a surface area of 0.38 ft2 and showed a loss of 139.7 grams of wa:er over a 71.9 hour1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> period giving a loss rate of 5.11 g ft-2hr-l. Based on a pool area of about 395 ft2 the annual evaporation rate would be 4680 gallons. This is of course dependent upon relative humidity, temperature of air and water, air movement, etc. For a pool3 H concentration of 62313 pCi/l (the average for July 1,1997 to June 30,1998) the tritium activity released from the ventilation system woutrl be 1104 pCi. A dilution factor of 2 x 10 8ml s-1 was used to calculate the unrestricted area concentration. This is from 200 m2(cross-section of the building) times 1 m s-1 (wind velocity). These are the values used in the safety analysis in the reactor license. A sample of air conditioner condensate showed no detectable H3 . Thus, there is probably very little 3H recycled into the pool by way of the air conditioner condensate and all evaporation can be assumed to be released.
l 3H released 1104 pC Average concentration, unrestricted area 1.8 x 10-13 Ci/ml Permissible concentration, unrestricted area 1 x 10-7 Ci/ml Percentage of permissible concentration 1.8 x 10-4 %
Calculated effective dose, unrestricted area 9 x 10-5 mrem 5
PSBR Annual Operating Report, FY 97-98 Environmental Surveys - Tech Specs requirement 6.6.1.f.
The only environmental surveys perfonned were the routine TLD gamma-ray dose measurements 4 at the facility fence line and at control points in a residential area several miles away. This !
reporting year's measurements (in millirems) tabulated below represent the July 1,1997 to June l 30,1998 period. The total readings appear higher this year; however, except for the Fence North I position, subtracting out the control measurement from the other measurements give environmental doses similar to those for past reporting years. A trend analysis of the Fence North position shows an increase beginning in July of 1995 in the annual reporting year cumulative doses. At this time, a GammaCell 220 dry cell irradiator was obtained and located in the Cobalt-60 facility, not far from the Fence North TLD. In September of 1998, a shielding wall was added to reduce exposures outside of the cobalt-60 bay. The Fence North TLD will
. also be moved from an interior fence location to an outer fence location to give a more representative measurement at the outer fenceline.
i 3rd Otr '97 4th Otr '97 1st Otr '98 2nd Otr '98 Intal Fence North 27.2 31.9 39.6 28.7 127.4 Fence West 25.7 21.1 33 20.4 100.2 Fence East 20.2 25.5 35 21.6 102.3 Fence South 20.3 27.4 31.5 18.8 98.0 Control 25.2 20.2 23.6 16.6 85.6 Personnel Exposures - Tech Specs requirement 6.6.1.g.
No reactor personnel or visitors received an effective dose equivalent in excess of 10% of the permissible limits und:r 10 CFR 20.
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