ML20149M302

Forwards Annual Operating Rept for Penn State Breazeale Reactor Which Covers Period from 950701-960630,per TS 6.6.1. Changes Applicable to 10CFR50.59,encl

ML20149M302
Person / Time
Site:	Pennsylvania State University
Issue date:	12/06/1996
From:	Witzig W PENNSYLVANIA STATE UNIV., UNIVERSITY PARK, PA
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
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ML20149M304	List: ML20149M302 ML20149M305
References
NUDOCS 9612160071
Download: ML20149M302 (7)

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PENNSTATE INTEROFFICE CORRESPONDENCE Radiation science and Engineering Center Annual Operating Report, FY 95-96 PSBR Technical Specifications 6.6.1 Lictase R-2, Docket No. 50-5 December 6,1996

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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). 'Ihis report covers the period from July 1,1995 through June 30,1996, as required by technical specifications requirement 6.6.1. Also included are any changes applicable to 10 CFR 50.59.

A copy of the Forty-First Annual Pmgn:ss Report of the Penn State Radiation Science and Engineering Center is included as supplementary informr. tion.

Sin rely yourg g, W ,is Y 'W f' arren F. Witzig Director, Radiation Science and Engineering Center Enclosures cc: Region I Administrator 1 U. S. Nuclear Regulatory Commission D. A. Shirley D. N. Womiley M. M. Reischman E. H. Klevans R. W. Granlund 9612160071 DR 961206 i

ADOCK 05000005 PDR 1 1233Idl

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PENN STATE BREAZEALE REACTOR  ;

Annual Operating Report, FY 95-%

PSBR Technical Specifications 6.,6.1 License R-2, Docket No. 50-5 i i

! Reactor Utilization j

%e Penn State Breazeale Reactor (PSBR) is a TRIGA Mark IH facility capable of 1 MW i

e steady state operation, and 2000 MW peak power pulsing operation. Utilization of the reactor and its '

associated facilities falls into three major categones:

EDUCATION utilization is primarily in the form oflaboratory classes conduced for

! graduate and undergraduate students and numerous high school science grouys. Dese -

f classes vary from neutron activation analysis of an unknown sample to the ca.ibration of a >

reactor contml 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 l

Applications, Neutron Radiograpy, a mynad 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 offered and  !

i are tailored to meet t ac needs of the participants. Individuals taking part in these programs fall into such categories as power plant operaung personnel, PSBR staff, and foreign traineen 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 laboratory courses or research i projects.  :

Summary of Reactor Operating Experience l Technical Snecifications reauimment 6.6.1.a.  ;

Between July 1,1995 and June 30,1996, the PSBR was critical for 591 hours0.00684 days <br />0.164 hours <br />9.771825e-4 weeks <br />2.248755e-4 months <br /> or 2.3 hrs / shift i suberitical for 423 hours0.0049 days <br />0.118 hours <br />6.994048e-4 weeks <br />1.609515e-4 months <br /> or 1.6 hrs / shift ,

l used while shutdown for 406 hours0.0047 days <br />0.113 hours <br />6.712963e-4 weeks <br />1.54483e-4 months <br /> or 1.5 hrs / shift not available 158 hours0.00183 days <br />0.0439 hours <br />2.612434e-4 weeks <br />6.0119e-5 months <br /> or 0.6 hrs / shift Total usage 1578 hours0.0183 days <br />0.438 hours <br />0.00261 weeks <br />6.00429e-4 months <br /> or 6.0 hrs / shift Tne reactor was pulsed a total of 96 times with the following reactivities:

< $2.00 49

$2.00 to $2.50 43

> $2.50 4

>= $3.00 0 he square wave mode of operation was used 93 times to power levels between 100 and 500 KW.

Total energy produced during this report period was 245 MWH with a consumption of 13 grams of U-235. 1 i

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Unscheduled Shutdowns

Technical Snecifications reauirement 6.6.1.b.

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ne 4 unplanned shutdowns during the July 1,1995 to June 30,1996 period are described I j- below. I

[ August 22,1995 - With the reactor operating at 5 watts, a reactor trip and evacuation alann j

! occuned when a senior reactor operator pushed a " push to set" button on the west air monitor. i i

his button is designed to indicate the local alarm setting on the chart reconter on the air i monitor but also sends a signal to the console. When the console sensed an alarm condition, l

the reactor trip and evacuation occuned. The operator thought he was only checking the local

alarm and did not understand that a signal would be sent to the console. Training on this issue I was done at an August 23,1996 staff meeting and a cover plate was installed over the " push  !

to set" button. Access to this button is only needed during monitor calibration. i

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October 5,1995 - With the reactor operating at 5 watts, a reactor trip occurred because of an  !

interlock validation failure while the shim rod was being moved to level power following a rod i l bump during a rod calibration experiment. A test plan was developed to look at all hardwase l 1- and software signals generated when the control rods are moved. It is believed that the

interlock validation failure occurred because the delay between the validation of the hardwired i

safety system and the DCC-X contml system software was greater than the one second  !

l allowed in the software timer block. During testing the validation occurs well within one I j second, and it has not been determined why on occasion the one second timing interval could j be exceeded. His interlock validation is not a Tech Spec requirement but is an additional  !

check by the software.

l February 2,1996 - While the reactor was operating at 750 kw, an Office of Physical Plant (OPP) workman assembling scaffolding in the neutron beam lab bumped the Remote l 1 Emergency Scram Button on the south wall. This resulted in a DCC-X control computer j i-stactor tnp. Since this occurrence, OPP has begun a system to have specific assigned j individuals service the PSBR building, decreasing the number of OPP personnel who frequent t the reactor building. Increased orientation is given to these individuals by the Supervisor of

! Facility Services.

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[ February 20,1996 - DCC-X reactor trip with East and West Fans Off console message while i i the reactor was operating at 13 KW. Office of Physical Plant (OPP) personnel turned off a bitaker with the mtent of turning off power for the cobalt-60 bay but instead tumed off the

power to the reactor bay including power for the bay exhaust fans. His situation was ,

l reviewed with OPP supervisors who agreed that no more work would be done until proper i

! labeling for the distribution and disconnect switches was checked. It was also noted tiat some

! work is needed to bring the distribution system up to desired code standards. Meanwhile the .

operating staff was told that labeling for the power distribution panels may be inconect. He 1

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reactor staff checked all main buildmg breakers for proper labeling on March 28,1996. ne

breaker panels throughout the building were found to be labeled consistent with the main

breaker panels in the lunch room area. The problem on February 20 was because the OPP i worker interpreted MPD to mean main power disconnect panel instead of main power i distribution panel; the staff feeling is that he should not have made this mistake. A work order ,

I with OPP has been started to bring the power distribution for the Cobalt-60 facility wiring u > l 1 to code standards; at the completion of that work, more prominent labels will be placed on t ie l main breaker panels in the lunch room area and a distribution diagram for the building will be  !

l placed there. l f.

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3 Major Maintenance With Safety Significance Technical Soecifications reauirement 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 i Technical Snecifications reauirement 6.6.1.d. '

l Facility Changes - On September 26,1995, a new version of a Recycle Program furnished by the console software manufacturer, AECL, was installed to provide a means of printing out all .

the current tuning values in each software loop. The previous version of this pmgram did not i pmvide enough significant figures to provide the information cecessary.

On February 2,1996, the local Area Network (LAN) software was modified to ennect a tag error that was noted during an evacuation drill. While the tags were correct on the window l that reads current radiation levels, the tags for the Cobalt-60 Bay and Reactor Neutron Beam  !

Lab radiation alarms had been reversed on another alarm window. This error was traced to a  ;

control file utilized in the LAN software. j On April 22,1996, modifications were completed to give north gate open and south gate open i alarms to the console (nonh and south gates are used to isolate the parking lot area behind the neutron beam lab when it is being used). Also, if the reactor is against a beam port, a reactor operate inhibit condition will now exist if the north and/or south gate is open and a beam port i door is open and a beam port door is open and/or a beam port plag is removed. j Procedures - All ?rocedures are reviewed as a minimum biennially, and on an as needed basis.

Changes during tie year were numerous and no attempt will be made to list them. A current copy of all facility procedures will be made available on request.

I New Tests and Expciments - On April 3,1995, a safety evaluation was conducted for an l experiment entitled "Detem6 tion of the Nuclear-Induced Electrical Conductivity in H 3 e for ,

Magnetohydrodynamic Energy Cocversion". Preliminary flux measurements were made in  !

the experimental stainless steel test chamler in September of 1996, with pulsing experiments planned for December 1996. The purpose of the aperiment is to obtain experimental data needed to continue the concept of nuclear-driven magneichydrodynamic (MHD) energy conversion. The experiments consist of a series of electrical coriductivity measurements in a quasi-static volume of pure 3He for a gas temperature of 300-1500K, a gas density (ret to std. atm. density) of 1 x 10-4 atm, and a neutron flux of 1 x 10-lx1016cm-2s-I. 'Ihe goal of these experiments is to provide a baseline data set for comparison to the model and to bener determine the value of the concept of nuclear-driven MHD energy conversion.

I i The safety evaluation concluded that there are no reactor components affected by this )

experiment. No reactor safety functions are affected and there are no potential effects on i reactor safety functions from this experiment. The experiment does not provide for a sudden rupture mechanism and there is no credible mechanism to cause an equipment malfunction that would increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the SAR. 'Ihere is no j credible possibility for an accident or malfunction of a different type than any evaluated previously in the SAR; the 3H is continuously being filtered out of the gas flowing thmugh the apparatus so a large release is not credible and there is no credible mechanism for fire or explosion, nor does the experiment create a possibility for an accident or malfunction of a 1

j different type than any previously evaluated in the SAR.

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It was concluded that no margin of safety is reduced by doing this experiment, and the experiment can be performed within the allowable limits for expen,ments in the Technical Specifications.

l On April 7,1996, a safety evaluation was conducted for an " Enhanced Flow Experiment".

The safe use of a secondary control rod as a control input for control experiments has been well established. The logical progression in control experiments is multiple input control.

The purpose of this experiment is to develop a method to use coolant flow as a second input to  !

the TRIGA reactor. A shroud will be placed around the core to control the amount of side flow allowed to the core. Restricting the side flow will allow a second control input but it will also put added limitations on the maximum power operation. The two components affected

• are the power level detector and the reactor core coo ing.' The power level detector is past of the safety system and has a power level trip associated with it. He core cooling removes the

heat from the reactor elements to allow operation at power. As flow is changed the flux profile may change acmss the core. A changing flux profile may cause a change in the relationship between the actual versus indicated power. In addition a change in the coolant temperature profile may change the calibration of the power instruments. Control rod positions shall be compared with and without the shmud at the same power level to give  ;

evidence of the affect on calibration. The proof of principle expenments will all be performed '

at a power of less than or equal to 200 kw. Normalized power for the hottest element in the core and the maximum power production per element will be less than analyzed in the SAR.

De pmposed experiment does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment to safety as previously evaluated in the SAR. By reducing the maximum allowable power to less than 200 kw during the initial experiments the assumptions and conclusions of this analysis and the SimuLine TRIGA model can be verified.

Any relaxation of the initial 200 kw power limit would only 'ue considered after the initial experiments am performed and the results taken to the Penn State Reactor Safeguards Committee for approval. The proposed procedure calls for constant review as the experiment progresses so that results more severe than predicted will be prevented by terminating the experiment. The experiment does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the SAR. The side flow control shutters shall be prevented from closing passively by mechanical means. Hey will be spring loaded to the

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open position and damnad so that they can not be closed without overt action. During this prehminary experiment they will be mechanically locked into a given position prior to operation. A loss of cooling accident is not credible. Since the reactor power will be limited to less than or equal to 200 kw for the initial experiments when the shroud is in place, the margin of safety will.not be reduced. This experiment will be conducted in the 1996-97 fiscal year.

Radioactive Effluents Released Technical Specifications reauirement 6.6.1.e.

Liquid here were no liquid effluent releases under the reactor license for the report period. Liquid l from the regeneration of the reactor demineralizer is evaporated and the distillate recycled for pool water makeup. The evaporator concentrate is dried and the solid salt residue is disposed ofin the same vray as other solid mdioactive waste at the University.

Liquid radioactive waste from the radioisotope laboratories at the PSBR is under the Umversity byproduct materials license and is transferred to the Health Physics Office for i

disposal with the waste from other campus laboratories. Liquid waste disposal techniques l

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include storage for decay, release to the sanitary sewer as per F CFR 20, and solidification  :

for shipment to licensed disposal sites. l Gaseous Gaseous effluent Ar-41, is released from dissolved air in the reactor pool water, dry irradiation tubes, and air leakage from the pneumatic sample transfer systems-De amount of Ar-41 released from the reactor pool is very dependent upon the operating I mwer level and the length of time at power. The release per MWH is lughest for extended 1 tigh power runs and lowest for intermittent low power runs. The concentrapon of Ar-41 in the reactor bay and the bay exhaust was measured by the Health Physics staif during the summer of 1986. Measurements were made for condidons of low and high power mas i simulating typical operating cycles. Based on these measurements, an annual release of I between 186 mci and 564 mci of Ar-41 is calculated for July 1,1995 to June 30,1996, resulting in an average concentration at ground level outside the reactor building that is 0.3 %

to 0.9 % 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 m2cross 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 6.i5 cant amounts of Ar-41. The calculated annual production was 55 mci. 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 system was minimal during this period and any Ar-41 release would be msignificant since the system operates with CO.2 as the fill gas.

Tritium release from the reactor pool is another gaseous release. He evaporation rate of the reactor pool was checked recently 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-2hr-1. 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 pool 3H concentration of 25,720 pCi/l (the average for July 1995 to June 1996) the tritium activity released from the ventilation system would be 455 Ci. A dilution factor of 2 x 108 mi s-1 was used to calculate the unrestricted area concentration. This is from 200 m2 (cross-section of the building) times I 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 iH. i Thus, there is probably very little 3Hrecycled into the pool by way of the air conditioner condensate and all evaporation can be assumed to be released.

3H released 455 C Average concentration, unrestricted area 7.2 x 10-14 Ci/ml Permissible concentration, unrestricted area 1 x 10-7 Ci/ml Percentage of permissible concentration 7.2 x 10-5 %

Calculated effective dose, unrestricted area 3.6 x 10-5 miem

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l Environmental Surveys

• Technical Soecifications reauirement 6.6.1.f. l

'Ihe only erwironmental surveys performed were the routine TLD gamma-ray dose measurements at the facility fenceline and at control points in residential areas several miles away. This reporting year's measurements (in millirems) tabulated below represent the July  !

1,1995 to June 30,1996 period. A comparison of the North, West, East, and South  :

fenceline measurements with the control measurements at Houserville (1 mile away) show the j differences to be similar to those in the past 3rd Otr '95 4th Otr '95 1 st Otr '96 2nd Otr '96 T.atal 1

Fence North 24.3 23.5 26.3 24.0 98.1 Fence West 19.2 18.4 18.1 18.5 74.2  ;

Fence East 24.0 18.5 19.5 20.3 82.3 i Fence South 17.6 20.2 20.1 18.3 76.2 i Control-Houserville 16.5 16.1 16.2 16.7 65.5 1 1

Personnel Exposures Technical Soecifications reauirement 6.61.e.  !

No reactor personnel or visitors received an effective dose equivalent in excess of 10% of the l permissible limits under 10 CFR 20.  ;

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