ML093631137
| ML093631137 | |
| Person / Time | |
|---|---|
| Site: | Pennsylvania State University |
| Issue date: | 12/17/2009 |
| From: | Unlu K Pennsylvania State Univ |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML093631137 (7) | |
Text
PENNSTATE KENAN IJNL-, Ph.D. Phone: (814) 865-6351 Director, Radiation Science and Engineering Center Fax:. (814) 863-4840 Professor, Department of Mechanical and Nuclear Engineering E-mail: k-unluRpsu.edu The Pennsylvania State University University Park, PA 16802-2304 RADLATIONSCENCE ENGINEERING &
CENTER Ainual Operating Report, FY 08-09 PSBR Technical Specifications 6.6.1 License R-2, Docket No. 50-5 December 17, 2009 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D. C. 20555
Dear Sir or Madame:
Enclosed please find the Annual Operating Report for the Penn State Breazeale Reactor (PSBR) at the Radiation Science and Engineering Center. This report covers the period from July 1, 2008 through June 30,.2009, as required by technical specifications requirement 6.6.1. Also included are any changes applicable to 10 CFR 50.59.
Sincerely yours, Kenan Unlii, Ph.D.
Director, Radiation Science and Engineering Center
Enclosures:
Annual Operating Report, FY 08-09 cc: E. J. Pell D. N. Wormley A. A. Atchley J. S. Brenizer E. J. Boeldt W. Kennedy - NRC College of Engineering An Equal Opportunity University
PENN STATE BREAZEALE REACTOR Annual Operating Report, FY 08-09 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 three 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 2500 visitors tour the PSBR facility each year.
RESEARCH accounts for a significant portion of reactor time which involves Radionuclear Applications, Neutron Radiography, and multiple research programs by faculty and graduate students throughout the University.
SERVICE use provides vital techniques for industries in support of the national economy. For example, radio-isotopes produced at the facility enable the critical petro-chemical industry to run at full capacity and the facility serves an important function in quality control of materials used to store the nation's spent nuclear fuel and in the electronics industry.
The PSBR facility operates on an 8 AM - 5 PM shift, five days a week, with early morning, evening, and weekend shifts to accommodate laboratory courses, public education and research or service projects as needed.
Summary of Reactor Operating Experience - Technical Specification 6.6.1 .a.
Between July 1, 2008 and June 30, 2009, the PSBR was critical for 806 hours0.00933 days <br />0.224 hours <br />0.00133 weeks <br />3.06683e-4 months <br /> or 2.6 hrs/shift subcritical for 312 hours0.00361 days <br />0.0867 hours <br />5.15873e-4 weeks <br />1.18716e-4 months <br /> or 1.0 hrs/shift used while shutdown for 562 hours0.0065 days <br />0.156 hours <br />9.292328e-4 weeks <br />2.13841e-4 months <br /> or 1.8 hrs/shift not available 39 hours4.513889e-4 days <br />0.0108 hours <br />6.448413e-5 weeks <br />1.48395e-5 months <br /> or 0.1 hrs/shift Total usage 1719 hours0.0199 days <br />0.478 hours <br />0.00284 weeks <br />6.540795e-4 months <br /> or 5.5 hrs/shift The reactor was pulsed a total of 160 times with the following reactivities:
< $2.00 19
$2.00 to $2.50 117
> $2.50 24 Page 1 of 6
Annual Operating Report, FY 08-09 The square wave mode of operation was used 20 times to power levels between 100 and 500 KW.
Total energy produced during this report period was 570 MWH with a consumption of 29 grams of U-235.
Unscheduled Shutdowns - Technical Specification 6.6.1 .b.
During the reporting period, 5 unscheduled reactor shutdowns (including 3 SCRAMs) occurred as follows:
On 11/19/08, during the conduct of a normal startup (reactor not yet critical),
DCC-X (control computer) requested an "Interlock Validation Failure" SCRAM.
The Reactor Safety System (RSS) functioned as designed and all rods bottomed as expected. The symptoms were not reproducible during troubleshooting and, after circuit analysis, both components that could cause the indications (rod control pushbutton and a relay), were replaced.
On 11/25/08, an unplanned normal reactor shutdown was initiated when off-site power was lost. Building power was restored after approximately 15 minutes and reactor operations were resumed.
On 02/25/09, during the conduct of a normal startup (reactor still shutdown by Technical Specification), DCC-X (control computer) requested an "Interlock Validation Failure" SCRAM. The Reactor Safety System (RSS) functioned as designed and all rods bottomed as expected. The cause was a sticking micro-switch in the "new" rod control pushbutton replaced after the 11/19/08 SCRAM.
A spare micro-switch in the same pushbutton assembly was placed in service. A review of component history does not indicate a failure rate on these pushbuttons that warrant increased replacement intervals or redesign. Failure is a reliability issue not a safety issue and a run to failure program remains for this component.
On 4/8/09, an unplanned normal reactor shutdown was initiated when a Radiation High alarm on the pneumatic sample transfer system monitor activated. An incorrect alarm setting on the monitor was corrected and operations were resumed.
On 6/25/2009, while operating at 3 kW, a reactor Watchdog SCRAM occurred when DCC-X (control computer) rebooted. A Watchdog SCRAM occurs when DCC-X does not continuously reset the Watchdog relay in the Reactor Safety ..
System. The Reactor Safety System functioned as designed and all rods bottomed as expected. A definitive cause for the event was not determined. But troubleshooting led to replacement of multiple components in the reactor Uninterruptible Power Supply (UPS).
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Annual Operating Report, FY 08-09 Major Maintenance With Safety Significance - Technical Specification 6.6.1 .c.
None Major Changes Reportable Under 10 CFR 50.59 - Technical Specification 6.6.1 .d.
No facility or procedure changes were reportable under 10 CFR 50.59.
Facility Changes of Interest To improve purification system performance and reduce radioactive waste, new pool water surface skimmers were installed in the reactor pool in early 2009. The change did not affect system piping and does not affect the LOCA analysis.
To increase core reactivity and reduce gamma levels during beam port operations, 10 graphite TRIGA elements were placed on the periphery of the core. Core reactivity change was less than 50 cents, rod worth changes were minimal and no other changes in core performance was observed.
To correct degraded performance of the primary cooling system, the primary cooling pump and heat exchanger isolation valves were replaced. The modification required cutting and welding of coolant piping outside the isolable boundary of the system. The changes do not affect the LOCA analysis.
Procedures Several single use procedures were developed as needed to support the core changes and system modifications. Additionally, procedures are normally reviewed biennially, and on an as needed basis. Numerous minor changes and updates were made to maintain procedures during the year and they will not be listed.
New Tests and Experiments None Page 3 of 6
Annual Operating Report, FY 08-09 Radioactive Effluents Released - Technical Specification 6.6.1 .e.
Liquid - Less than 20% of the allowed or recommended concentrations There were no planned liquid effluent releases under the reactor license for the report period.
Liquid radioactive waste from the radioisotope laboratories at the PSBR is under the University byproduct materials license and is transferred to the Radiation Protection Office for disposal with the waste from other campus laboratories.
Liquid waste disposal techniques include storage for decay, release to the sanitary sewer is per 10 CFR 20, and solidification for shipment to licensed disposal sites.
Gaseous - Less than 20% of the allowed or recommended concentrations Gaseous effluent Ar-41 is released from dissolved air in the reactor pool water, air in dry irradiation tubes, air in neutron beam ports, 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.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 Radiation Protection 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 and current MWH, an annual release of between 432 mCi and 1311 mCi of Ar-41 is calculated for July 1, 2008 to June 30, 2009, resulting in an average concentration at ground level outside the reactor building that is 0.7 % to 2.1 % of the effluent concentration limit in Appendix B to 10 CFR 20. The concentration at ground level is estimated using only dilution by a 1 meter/sec wind into the lee of the 200 m 2 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 319 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. However, even if all of the 319 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 2.6 %.
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Annual Operating Report, FY 08-09 Production and release of Ar-41 from reactor neutron beam ports was minimal.
Beam port #7 has only three small collimation tubes (each 1 cm2 area) exiting the port and any Ar-41 production in these small tubes is negligible. Beam port #4 has an aluminum cap installed inside the outer end of the beam tube to prevent air movement into or out of the tube as the beam port door is opened or closed. The estimated Ar-41 production in beam port #4 for all beam port operations is 56 mCi. With the aforementioned aluminum cap in place, it is assumed that this Ar-41 decayed in place. Radiation Protection Office air measurements have found no presence of Ar-41 during beam port #4 reactor operations with the beam port cap in place.
The use of the pneumatic transfer system (rabbit) was minimal during this period and any Ar-41 release would be insignificant since the system operates with CO-2 as the fill gas. A small amount of Ar-41 is released from each rabbit capsules. A 2 minute irradiation @900kW will produce .0026 mCi. In the 2008-09 reporting period 50 rabbit capsules were irradiated at a variety of power/time combinations (typically less than 900kW). The resulting 0.13 mCi of Ar-41 are not a significant contributor.
Tritium release from the reactor pool is another gaseous release. The pool loss rate due to evaporation and pump leak-off averages - 13000 gallon/year (4.9 E4 1/year). 'For a pool Tritium concentration of 20471 pCi/1 (the average for July 1, 2008 to June 30, 2009), the tritium activity released from the ventilation system would be 1000 gCi. A dilution factor of 2 x 108 ml/sec was used to calculate the unrestricted area concentration. (The concentration at ground level 2
is estimated using a dilution by a 1 meter/sec wind into the lee of the 200 m cross-section of the reactor bay). These are the values used in the safety analysis in the reactor license.
Tritium released -1000 tCi Average concentration, unrestricted area 1.6 x 10-15 jICi/ml Permissible concentration, unrestricted area 1 x 10-7 gCi/ml Percentage of permissible concentration <.1 %
Calculated effective dose, unrestricted area <10- 4 mRem Page 5 of 6
Annual Operating Report, FY 08-09 Environmental Surveys - Technical Specification 6.6.1 .f.
The only environmental surveys performed were the routine TLD gamma-ray dose measurements at the facility fence line and at control points in two residential areas several miles away. .This reporting year's .net measurements (in millirems) are tabulated below represent the July :1, 2008 to June 30, 2009 period.
3rd Qtr '08 4th Qtr '08 1st Qtr '09 2nd Qtr '09 Total Fence North 12.9 11.4 8.5 8.9 41.7 Fence South 9.3 11.2 9.4 8.2 38.1 Fence East 14.8 8.1 7.3 9.6 39.8 Fence West 10.1 8.3 4.8 8.6 31.8 Control- 10.9 10.7 9.1 11.9 42.6 Pleasant Gap Control- 1.2 35.4 -1.8 0.8 35.6 State College I There is no meaningful increase in exposure at the facility fenceline due to licensed operations for the current fiscal year.
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