ML083510752

Penn State Breazeale Reactor, Annual Operating Report Fy 07-08 Technical Specifications 6.6.1

ML083510752
Person / Time
Site:	Pennsylvania State University
Issue date:	12/05/2008
From:	Unlu K Pennsylvania State Univ
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML083510752 (7)

Text

PENNSTATE KENAN UNLO, 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-unlu(@,psu.edu The Pennsylvania State University University Park, PA 16802-2304 Annual Operating Report, FY 07-08 PSBR Technical Specifications 6.6.1 License R-2, Docket No. 50-5 December 5, 2008 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, 2007 through June 30, 2008, as required by technical specifications requirement 6.6.1. Also included are any changes applicable to 10 CFR 50.59.

Sincerely yours, Kenan Onlil, Ph.D.

Director, Radiation Science and Engineering Center

Enclosures:

Annual Operating Report, FY 07-08 cc: E. J. Pell D. N. Wormley A. A. Atchley D. Sathianathan E. J. Boeldt W. Kennedy - NRC M. Voth - NRC 4o2~u College of Engineering An Equal Opportunity University

PENN STATE BREAZEALE REACTOR Annual Operating Report, FY 07-08 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, the calibration of a reactor control rod, experiments with radiation detectors and applications/demonstration of various nuclear methods. In addition, an average of 2500 visitors tour the PSBR facility each year.

RESEARCH accounts for a large portion of reactor time which involves Radionuclear Applications, Neutron Radiography/Imaging, multiple research programs by faculty and graduate students throughout the University.

SERVICE activities consisted with utilization of Radiation Science and Engineering Center Facilities for mostly industrial sector users. These activities involve various irradiation, testing and analysis of materials.

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 - Tech Specs requirement 6.6.1 .a.

Between July 1, 2007 and June 30, 2008, the PSBR was:

critical for 902 hours0.0104 days <br />0.251 hours <br />0.00149 weeks <br />3.43211e-4 months <br /> or 2.9 hrs/shift subcritical for 447 hours0.00517 days <br />0.124 hours <br />7.390873e-4 weeks <br />1.700835e-4 months <br /> or 1.4 hrs/shift used while shutdown for 809 hours0.00936 days <br />0.225 hours <br />0.00134 weeks <br />3.078245e-4 months <br /> or 2.6 hrs/shift not available for 318 hours0.00368 days <br />0.0883 hours <br />5.257936e-4 weeks <br />1.20999e-4 months <br /> or 1.0 hrs/shift Total usage 2476 hours0.0287 days <br />0.688 hours <br />0.00409 weeks <br />9.42118e-4 months <br /> or 7.9 hrs/shift Page 1 of 6

Annual Operating Report, FY 07-08 The reactor was pulsed a total of 173 times with the following reactivities:

< $2.00 6

$2.00 to $2.50 126

> $2.50 41 The square wave mode of operation was used 31 times to power levels between 100 and 500 KW.

Total energy produced during this report period was 675 MWH with a consumption of 35 grams of U-235.

Unscheduled Shutdowns - Tech Specs requirement 6.6.1 .b.

During the reporting period, one unscheduled reactor shutdown occurred on August 7 th, 2007. The control computer (DCC-X) requested a safety system SCRAM during a normal startup (< $1 shutdown but not yet critical) when AC power was momentarily lost to both reactor bay exhaust fans due to maintenance activities on fire alarm system and emergency lighting diesel. The DCC-X SCRAM request feature prevents reactor operation without bay fans (a Tech. Spec. Limiting Condition for Operation).

Major Maintenance With Safety Significance - Tech Specs requirement 6.6.1.c.

On August 2 0 th, 2007, Fuel Element #115 was permanently removed from service. The 12 weight percent (w/o) element was observed to bind slightly in the grid plate when removed for annual fuel inspection. Slight binding in the fuel inspection apparatus was also noted. No visible anomalies were observed and no casual analysis was conducted.

Bowing and swelling of 12 w/o elements has been experienced in the past and is believed due to the previous practice of operating 12 w/o elements in the B-ring (center) of small core configurations.

On October 9 th, 2008, the reactor entered into an unscheduled maintenance outage to repair reactor pool through-wall leakage. The PA Department of Environmental Protection and the NRC were notified of the leak. The repair activities included concrete repair (Belzona TM 4111 Magma-Quartz) and the application of a Polyurea (InstaCoteTM ML-2) spray-on liner. The reactor returned to service on November 21. (See NRC Inspection report No. 50-5/2007-203).

On June 2 nd, 2008, Instrumented Element #I-11 was permanently removed from service.

The 12 weight percent (w/o) element was observed to bind slightly in the grid plate when removed for annual fuel inspection. Slight binding in the fuel inspection apparatus was also noted. No visible anomalies were observed and no casual analysis was conducted.

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Annual Operating Report, FY 07-08 Bowing and swelling of 12 w/o elements has been experienced in the past and is believed due to the previous practice of operating 12 w/o elements in the B-ring (center) of small core configurations.

Major Changes Reportable Under 10 CFR 50.59 - Tech Specs requirement 6.6.1 .d.

No facility or procedure changes were reportable under 10 CFR 50.59.

Facility Changes of Interest As noted above, during the leak outage the reactor pool wall was modified with the addition of (BelzonaTM 4111 Magma-Quartz) concrete repair material and the application of a Polyurea (InstaCoteTM ML-2) spray-on liner. The liner is chemically inert and was tested by a nuclear utility to 200MRad with little degradation. The liner is expected to be a significant improvement over the epoxy paint coatings that had historically protected the concrete. The modification had no impact on the LOCA design basis (drain line failure).

Also during the leak outage, minor modifications to the primary coolant header were made with a change from hard mounting to load bearing "feet" (to minimize penetrations in the liner), the addition of a small drain on the base of the header and a siphon break vent on the header high point. These modifications reduce the likelihood of future pool leaks and should break a siphon of the reactor pool if it occurs through the coolant system. The modification had no impact on the LOCA design basis (drain line failure) and the primary coolant system has no specific reactor safety function.

Also during the leak outage, the purification system inlet pipe was modified to allow purification operation with lowered pool water level. This improved water chemistry during latter stages of the outage. This modification is not used during normal operations, does not increase the likelihood of a purification system leak and had no impact on the LOCA design basis.

Procedures Several single use procedures were developed as needed to support the pool leak repair and associated 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 ExDeriments None Page 3 of 6

i Annual Operating Report, FY 07-08 Radioactive Effluents Released - Tech Specs requirement 6.6.1 .e.

Liquid - Less than 25% of the allowed or recommended concentrations There were no planned liquid effluent releases under the reactor license for the report period. The through-wall pool leakage in October and November released approximately 12,000 gallons of water containing very low levels of tritium (28,000 pico-curies/liter -

2.8E-5 micro-curies/ml) into the backfill around the reactor pool. Off-site sampling of ground water wells has not shown any detectable tritium and the leak was repaired as described above. The concentration of this water was less than 25% of the allowable effluent limits and the total release is less than 1.3 mCi.

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 25% 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 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 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, an annual release of between 512 mCi and 1,553 mCi of Ar-41 is calculated for July 1, 2007 to June 30, 2008, resulting in an average concentration at ground level outside the reactor building that is 0.8 % to 2.5 % 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 I m/s 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 509 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 509 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 3.5 %.

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Annual Operating Report, FY 07-08 Production and release of Ar-41 from reactor neutron beam ports was minimal. Beam port #7 has only three small collimation tubes (each 1 cm 2 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 44 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 2007-08 reporting period 492 rabbit capsules were irradiated at a variety of power/time combinations (typically less than 900kW). The resulting 13 mCi of Ar-41 are not a significant contributor.

Tritium release from the reactor pool is another gaseous release. The evaporation rate of the reactor pool was checked previously 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 water 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-2 hr-1 . Based on a pool area of about 395 ft 2 the annual evaporation rate would be 4,680 gallons. This is of course dependent upon relative humidity, temperature of air and water, air movement, etc. For a pool 3H concentration of 23,251 pCi/l (the average for July 1,; 2007 to June 30, 2008) the tritium activity released from the ventilation system would be 412 1tCi. A dilution factor of 2 x 10i ml s1 was used to calculate the unrestricted area concentration. This is from 200 m 2 (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 a previous year showed no detectable 3 H. 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.

3H released 412 itC Average concentration, unrestricted area 6.5 x 10 [tCi/ml Permissible concentration, unrestricted area 1.0 X 107 ýtCi/ml Percentage of permissible concentration 6.5 x 10-5%

Calculated effective dose, unrestricted area <4 x 10_5 mRem Page 5 of 6

• bJ

• f Annual Operating Report, FY 07-08 Environmental Surveys - Tech Specs requirement 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, 2007 to June 30, 2008 period.

3rd Qtr '07 4th Qtr '07 1st Qtr '08 2nd Qtr '08 Total Fence North 14.6 11.3 8.5 10.3 44.7 Fence South 16.3 -.6 10.3 9.4 35.4 Fence East 17.9 12.4 8.5 11.1 49.9 Fence West 14.5 11.8 7.2 8.5 42.0 Control 13.5 16.6 6.6 9.6 46.3 Control 3.8

• 1.64 2.2 7.64

• Control missing/lost There is no meaningful increase in exposure at the facility fenceline due to licensed operations for the current fiscal year.

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