ML20197A440
| ML20197A440 | |
| Person / Time | |
|---|---|
| Site: | Pennsylvania State University |
| Issue date: | 06/30/1997 |
| From: | PENNSYLVANIA STATE UNIV., UNIVERSITY PARK, PA |
| To: | |
| Shared Package | |
| ML20197A447 | List: |
| References | |
| NUDOCS 9712230060 | |
| Download: ML20197A440 (6) | |
Text
._ . _ _ __ _ _ . -_ __ __
l i
PENN STATE BREAZEALE REACTOR i
Annual Operating FY % 97 PSBR Technical ficatloas 6.6.1 '
License R 2, Doc et No. 50 5 Reactor Utilisation ,
lhe Penn State Bremwa Reactor (PSBR) ls a TRIGA Mark 111 facility capable of 1 MW steady state operation, and 2000 MW peak power pulsing operation. Utilizatlon of the reactor and its associated facilities falls into two major categories:
EDUCATION utillratkm is primarily in the forrn of laboratory classes conducted for graduate and undergraduate students and numerous high school sclerce 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.
RESEARCll/ 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 Univers ity, and various applications by the indet trial sector.
The PSBR facility ogrates on an 8 AM 5 PM shift, five days a week, with an occasional 8 AM 8 PM or 3 AM 12 Midnight shift to accommodate laboratory courses or research projects.
Summary of Reactor Operatl . ! Experience Tech Specs requirement 6.6.1.a.
Between July 1,1996 and June 30,1997, the PSBR was critical for 440 hours0.00509 days <br />0.122 hours <br />7.275132e-4 weeks <br />1.6742e-4 months <br /> or 1.7 hrs / shift subcritical for 348 hours0.00403 days <br />0.0967 hours <br />5.753968e-4 weeks <br />1.32414e-4 months <br /> or 1.3 hrs / shift used while shutdown for 381 hours0.00441 days <br />0.106 hours <br />6.299603e-4 weeks <br />1.449705e-4 months <br /> or 1.4 hrs / shift not available 381 hours0.00441 days <br />0.106 hours <br />6.299603e-4 weeks <br />1.449705e-4 months <br /> or 1.4 hrs / shift Total usage 1550 hours0.0179 days <br />0.431 hours <br />0.00256 weeks <br />5.89775e-4 months <br /> or 5.8 hrs / shift The reactor was pulsed a total of 76 times with the following reactivitiu:
< $2.00 17
$2.00 to $2.50 15
> $2.50 44
>= $3.00 0 The square wave mode of operation was used 35 times to power levels between 100 and 500 KW.
Total energy produced during this report period was 182 MWil with a cos..umption of 9 grams nf U 235.
5 t>k
, .v.. ,,,.n-,, - ~ - -
i a j Unactiedeled Stemidowns . Tech Specs requirement 6.6.1.b. ]
- ne 5 unplanned shutdowns during the July 1,1996 to June 30,1997 period are descdbed below October 16,1996. Ihring a reactor startup, the reactor operator tripped the reacw when an expedmonter repanse that a sample he had reponed retumed to the laboratory terminus (based on
. sound)during a previous esperiment was ret in the terminus when he went to remove it. De j
- sequence of events was as follows. At 1132, the last in a series of 16 samples was 1 retumed to the terminus by the experimenter. At 1133, the reacw was taken to stand and at !
1137 the reactor was moved to the other end of the reactor pool for stanup for another l expedment. At 1140, the reactor stanup, the reactor was tripped by the o wrator when the l esperimenter repoemd the t sample run 8 minutes earlier was not in the term aus; this trip t was as are since the sample could still have been in the reactor core. ne system was ;
e and the sample was found several feet fmm the laboratory terminus. The experimenter was ing on sound to know that the sample had retumed to the terminus and was leaving :
the sample in the terminus for several minutes for decay before removal for countity. De 50P.
9, Pneumatic Transfer System Pmcedure, did not specifically call for visual ins ton to verify l that the sample was back. Dose calculations showed that a person standing u r the area where ,
the sample was located wouki have tWi*.'ed 0.1 mrem in 30 minutes. Fo lowing the ;
occurrence, a see thmugh windovN Nbd in the rabbit terminus to allow visual vedfication i i
of samp's return.le retum. The SOP-9 pmLH nt inodified to require the visual ve sample October 30,1996 While the reactor was oper,eting at 7501.0, the operator initiated a reactor trip I whert he received a Rabbit i Radiation level High Alarm. This alarm disables the fan and opens a bypass valve to relieve system pressure through a charcoal filter and an absolute filter. At the time of the radiation alarm and reactor trip, the last in a series of 18 samples was still in the core.
De SRO reset the alarm on the reactor hay alarm panel and at this point the fan came back on and :
the sample unexpectedly returned to the laboratory terminus, it is an SOP 9, Pneumatic Transfer System Cperation, pmcedure seguirement to not nm the fan without Health Physics permission l upon the receipt of a high radiation alarm to prevent the possibility of radiation being spread ;
i through the system due to a broken sample encapsulation (no contamination lem resultedin
, this case since the sam >le capsule was found intact). The RO had correctly pped the reactor i
but erred in that he hac not secured the rabbit system fan. SOP 9 was extens vely re written to pmvide better instructions for reactor operators and experimenters. The source of the radiation ;
alarm was believed to be an Argon-41 build up in the system. ;
- December 3,1996 - While the reactor was operating at 50 watts, the reactor operator inadvenently pressed a console calibrate bution while putting a log book stamp awa calibrate button being pressed caused a "do not operate" condition De reactor and a re, safety system is designed to operate this way since in calibration mode the actual detector signal j is bypassed and an induced signal is insened into the safety system. %c root cause was operator i ermr, !
December 16,1996 During the founeenth in a series of pulses for an experimenter, a normal peak power was recorded but Fuel Temp 1 and Fuel Temp 2 read only 33,7 degrees C and 32.5 .
degrees C respectively instead of an expected ap smnimate 490 degrees C. His was reponed to l the NRC as a Tech Spec violation on December L7,1997 since the fuel element temperature limiting safety system setting was inoperative during pre pulse steady state operation and during the subsequent pulse. An investigation revealed that when the experimenter exceeded the design !
specs of his data acquisition system, a ground loop caused the reactor safety system fuel element thermocouples to shon, thus disabling them just prior to and during a reactor pulse. The reactor >
safety system dkl not have true isolat on fmm the experimental setup. De reactor was not i
.i
_m_. m_. m . . _ _ _ _. _ aw?
3 l
i operated from December 16,1996 to January 24,1997 while this event was investigated and completely understood. As a result, no reactor signals will go into or out of the safety system operations boundary (other than to the DCC X monitoring computer which is a part of the reactor control system) unless adequate isolation is provided between the exlwrimental setup and the reactor safety system. Further details of this event can be found in tie 14-day letter dated December 20,1996 and an addendum to that letter dated on January 23,1997, both submitted to ,
the NRC.
February 6,1997 - While the reactor was operating at 750 kw against beam port #4, the reactor operator tripped the reactor as per pmcedure when he received a neutron beam laboratory beam gate open alarm. The gate opened when a reactor staff member pushed agalmt the rate with a survey meter while checking radiation levels. The gate had been closed (alarm conoition had cleared) but had not securely leiched. Adjustments were made to the door alignment and the door was fitted with a deadbolt k*k so a key must now be used to close the door, thus indicating a ,
definitive closed position to the operator.
Major Maintenance With Safety Significance . Tech Specs requirement 6.6.1.c.
No major mventative or corrective maintenance operations with safety significance have been performed during this repott period.
Major Changes Reportable Under 10 CFR 50.59. Tech Specs requirement 6.6.1.d.
Facility Changes-On February 13,1997, the lead t.nield door on beam port #4 was modiried for auto operation by a pneumatic cylinder. Previously the door was ojuned before the reactor was moved against tle beam pt and closed after the reactor was moved away from the beam port. With this modif cation the door is only open during the collection of experimental data, thus minimiring dose to experimenters, unnecessary activation of experimental equipment and degradation of the real time neutmn radiography camera and as!.ociated equipment.
During April of 1997, a new D20 tank was installed to replace an existing D2O tank. The south side of the pool was drained during installation. 'Ihis new tank was designed to enhance the neutron beam quality. It was positioned tangentially to the core in an effort to impmvc the neutron to gamma ratio.
On June 10,1997, the bridge west radiation monitor and bridge hi range monitor were moved about six inches higher on the bridge to climinate interference with west movement of tie reactor bridge tower against a new Fast Neutron Irradiator. Measurements indicate a decrease of 0.5 mR/hr as measured by the bridge west monitor at 750 kw. This is not a significant change since a normal reading on this monnor at 750 kw is about 15 mR/hr and the alert and alarm set points are 50 mR/hr and 200 mR/hr, respectively.
On June 18,1997, the alert and alarm set points on the neutron beam laboratory radiation monitor were permanently changed to 10 mR/hr for the alert and 20 mR/hr for th- alarm. The set-points had historically been 3 mR/hr for the alert and 6 mR/hr for the alarm. These adjustments followed the installation of a new D2O tank that has greatly enhanced the neutron beam in the neutron beam laboratory.
Pmoedures -
All 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. A current copy of all facility pmcedures will be made available on request.
4 New Tests arglEineriments -
D2O Tank On January 6,1997, a safety evaluation was done for a new D2O tank . This new tank was designed to operate tangentially to the core to increase the neutmn to gamma ratio. %e new sma let tank places less D20 between the core and the team lab for an increased neutron team intensity.
The reactivity effect of a leak of D20 replacing the water in the core is negative. The tank was leak tested and care was taken durin ; installation to minimize and monitor stress on the beam port penetration. Materials were selectet to minimize any future radioactive waste disposal problems for the tank. A method of monitoring the level of D20 in the tank was provided. The location of the core neutrcm detector away from the core to tank interface minimizes any effect on the detector calibration. The collimator tube shleiding was designed to withstand an impact of 4 tons, greatly exceeding the lifting capacit) of the 3 ton overhead crane in the reactor bay.
Calculations showed the tank and suppon structure were not buoyant with the tank empty.
Ilecause of known interactions letween D20 and new aluminum, the tank was equipped with a pressure gauge and relief valve. No pressure build ups were noted following installation.
De safety evaluation concluded that there are no reactor components affected by this D2O tank.
No reactor safety functions are affected and there are no potential effects on reactor safety functions from this tank. 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 ec ul ament imponant to safety as previously evaluated in the SAR. There is no credible pass bl .ity for an accident or malfunction of a different type than any evaluated previously in the SAR nor does the tank create a possibility for an accident or malfunction of a different type than any previously evaluated in the SAR. The use of the tank does not reduce the margin of safety as defined in any basis for any Technical Specification. No change in Technical Specifications or I.icense were required.
Fast Neutron irradiator (FNI)
On Janursty 13,1997 (with a final revision on June 9,1997), a safety evaluation was done for the installation and use of the FNI. The FN! was designed to te used for the fast neutron irradiation of silicon wafers. The FNI accommodates larger wafers (8 inch) than a previous Fast Flux Tube (FIT) used safely since 1984 that could accommodate up to only 5 inch wafers. De FIT uses an annular design for both the dry irradiation tube and the surrounding lead, boron and cadmium shielding. The FNI uses an annular dry irradiation tube, but the lead and boron shielding are rectangular by design to pmvide a flat coupling face between the FNI and the reactor core face to minimize the water moderator effect. Provisions are made to allow for the expulsion of water from the enclosed aluminum cowling that surrounds the lead and boron shield to eliminate any water moderation within the shield. The tube is designed with a negative buoyancy under all conditions. Materials were selected to minimize eventual radioactive waste disposal problems.
The support structure was designed and constructed to support the weight of the shielding material and dry tute, to distribute the weight over a sufficient area of the floor, and to restrict its horizontal movement. Flooding of the irradiation tube or the shield cowling will not cause a significant reactivity effect; any effect there would te would be masked by the effect of the bomn in the shield and this was verified by experiment after FNI installation. The reactivity effect of the boron shield was judged to te well within any limits on ex yeriments as defined in the Technical Specifications and was measured and found accepta ale as part of the FN1 certification process following installation. An analysis of the effect of the restricting of the core flow area on the side of the core against the FNI was analysed as not being significant; this was verified by
5 fuel temperature measurements after installation of the FNI Argon-41 production isjudged to be acceptable but will te monitored. Administrative controls to assure the shield plug is in wien the tule is used have proven adequate whh the exist!ng FIT.
Radioactive Effluents Released . Tech Specs requirement 6.6.1.e.
Lhuld There were no liquid efnurat releases under the reactor license for the report period. Liquid from the regencrution 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 of in the same way as other solid radioactive waste at the University. Presently, the demineralizer beds are not n(mnally regenerated but 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 llealth 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 solidificauon for shipment to licensed disposal sites.
GAKQUI Gaseous efnuent Ar 41 is released from dissolved air in the reactor pool water, air in dry irradiation tutes, 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 MWil 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 liealth 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 138 mci and 418 mci of Ar-41 is calculated for July 1,19% to June 30,1997, resulting in an average concentration at ground level outside the reactor building that is 0.3 % to 0.8 % of the effluent concentration limit in Appendix H 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 99 mCl. 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 react ~ bay. The re ported releases from dissolved air in the reactor pool are based on measurements made, in part, waen 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 prevlous paragraph. 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 fmm 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 ft 2 nd a 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 ft2 the annual evaporation rate would be 4680 gallons. This is of course dejendent upon relative humidity, temperature of air and water, air movement, etc. For a pool 311 concentration of 63,190 pCi/l (the average for July 1,1996 to June 30,1997) the tritium activity released from the
_ - - _ ___r _- ___ _m _m.__-__m_.___ _ -
_ . _ - . . _ _ m- m._. __. _ _._m_ -. _ _ . . _ _ _ . _ . _ _ _ _ _ _ __ _ _ _ __
6 ventiladon system would be 1119 Ci. A dilution factor of 2 x 10 8ml s-1 was used to calculate ]
the unrestricted area concentration. His is fmm 200 m2 (cross.section of the building) times 1 l analysis in the reactorlicense. A -;
m sample s 1of(wind velocity).-
air conditioner nese condensate showed arenothe values used in the safeyH. Dus, there(
detectable little 3H recycled into the pool by way of the air conditioner condensate and all evaporation can be assumed to be released. i 3H released 1119 kC f
- Average concent.ation, unrestricted area 1.8 x 1013 pCihnt l Pennissible conczntration, unrestricted area 1 x 10-7 Cihnl l Percentage of permissible concentration 1.8 x 10-4 %
Calculated effective done, unre:tricted area 9 x 10 5 mrem ;
. i Envi.onsmental Surveys Tech Specs requirement 6.6.1.f. l De only environmental surveys performed were the routine TLD gamma ray dose measurements - l st the facility fence line and at control points in residential areas several miles away. His 1
- re xwting year's measurements (in mill irems) tabulated below represent the July 1,1996 to June 3(,1997 period. A comparison of the North, West, East, and South fence line measurements ,
with the control measurements at Houserville (1 mile away) show the differences to be slightly higher but similar to those in the past 3rd Otr '% 4th Otr '% lst Otr '97 2nd Otr '97 IQ1&l l Fence North 27.8 29.7 27.5 28.8 113.8 Fence West 17.7 19.9 17.6 30.2 85.4
. Fence East - 19.0- 21.1 21.2 26.9 88.2 l
- Fen <w South 19.7 21.9 22.7 23.7 88.0 Control Houserville 15.8 16.4 17.4 20.7 70.3 i Personnel Exposures Tech Specs requirement 6.6.1.g.
No reactor rsonnel or visitors received an effective dose equivalent in excess of 10% of the permissible imits under 10 CFR 20.
f i
i a
f
' . - , t-w& a+. .m- Tt M -.m4==-de --se-- r---a--
y.-,m ;y-- v erre-%- $ 1 #"'E' 'gFT71rY La e rWV -Mr.-DMrm we wy-% '-t' 59 T 79 y t rie-y-yq w -1r--TFg'v-wr--^-tyy=yy' g- 't'--
"-