ML062430261

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Initial Examination Report No. 50-133/OL-06-01, University of Arizona
ML062430261
Person / Time
Site: 05000113
Issue date: 09/12/2006
From: Johnny Eads
NRC/NRR/ADRA/DPR/PRTB
To: Tolbert L
Univ of Arizona
Witt K, NRC/NRR/ADRA/DPR/PRT, 415-4075
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OL-06-01
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September 12, 2006Dr. Leslie TolbertVice President for Research University of Arizona Tucson, AZ 85721-0066

SUBJECT:

INITIAL EXAMINATION REPORT NO. 50-113/OL-06-01, UNIVERSITY OFARIZONA

Dear Dr. Tolbert:

During the week of August 15, 2006, the NRC administered an operator licensing examinationat your University of Arizona Research Reactor. The examination was conducted according to NUREG-1478, "Non-Power Reactor Operator Licensing Examiner Standards," Revision 1. Examination questions and preliminary findings were discussed with those members of your staff identified in the enclosed report at the conclusion of the examination. In accordance with 10 CFR 2.390 of the Commission's regulations, a copy of this letter and theenclosures will be available electronically for public inspection in the NRC Public DocumentRoom or from the Publicly Available Records (PARS) component of NRC's AgencywideDocuments Access and Management System (ADAMS). ADAMS is accessible from the NRC Web site at (the Public Electronic Reading Room) http://www.nrc.gov/reading-rm/adams.html. The NRC is forwarding the individual grades to you in a separate letter which will not bereleased publicly. Should you have any questions concerning this examination, please contact Mr. Kevin Witt at (301) 415-4075 or via internet e-mail kmw@nrc.gov

.Sincerely, /RA/

Johnny Eads, ChiefResearch and Test Reactors Branch B Division of Policy and Rulemaking Office of Nuclear Reactor RegulationDocket No. 50-113

Enclosures:

1. Initial Examination Report No. 50-113/OL-06-012. Facility comments with NRC resolution
3. Examination and answer keycc w/encls:Please see next page University of ArizonaDocket No. 50-113cc w/encl:

Office of the MayorP.O. Box 27210 Tucson, AZ 85726-7210Arizona Radiation Regulatory Agency4814 S. 40th Street Phoenix, AZ 85040University of ArizonaNuclear Research Laboratory ATTN: Dr. John Williams, Reactor DirectorBldg. 20, Rm 200 Tucson, AZ 85721-0020University of ArizonaNuclear Research Laboratory ATTN: Robert Offerle, Reactor Supervisor Bldg. 20, Rm. 200 Tucson, AZ 85721-0020University of ArizonaATTN: Dr. Caroline M. Garcia Assistant Director, Arizona Research Labs Gould-Simpson Bldg. 1011 P.O. Box 210077 Tucson, AZ 85721-0077University of ArizonaATTN: Daniel Silvain, Radiation Safety Officer 1640 North Vine Tucson, AZ 85721-0020Test, Research and TrainingReactor Newsletter 202 Nuclear Sciences Center University of Florida Gainesville, FL 32611 September 12, 2006Dr. Leslie Tolbert Vice President for Research University of Arizona Tucson, AZ 85721-0066

SUBJECT:

INITIAL EXAMINATION REPORT NO. 50-113/OL-06-01, UNIVERSITY OFARIZONA

Dear Dr. Tolbert:

During the week of August 15, 2006, the NRC administered an operator licensing examinationat your University of Arizona Research Reactor. The examination was conducted according to NUREG-1478, "Non-Power Reactor Operator Licensing Examiner Standards," Revision 1. Examination questions and preliminary findings were discussed with those members of your staff identified in the enclosed report at the conclusion of the examination.In accordance with 10 CFR 2.390 of the Commission's regulations, a copy of this letter and theenclosures will be available electronically for public inspection in the NRC Public DocumentRoom or from the Publicly Available Records (PARS) component of NRC's AgencywideDocuments Access and Management System (ADAMS). ADAMS is accessible from the NRC Web site at (the Public Electronic Reading Room) http://www.nrc.gov/reading-rm/adams.html. The NRC is forwarding the individual grades to you in a separate letter which will not bereleased publicly. Should you have any questions concerning this examination, please contact Mr. Kevin Witt at (301) 415-4075 or via internet e-mail kmw@nrc.gov

.Sincerely, /RA/Johnny Eads, ChiefResearch and Test Reactors Branch B Division of Policy and Rulemaking Office of Nuclear Reactor RegulationDocket No. 50-113

Enclosures:

1. Initial Examination Report No. 50-113/OL-06-012. Facility comments with NRC resolution
3. Examination and answer keycc w/encls:Please see next pageDISTRIBUTION w/ encls.:PUBLIC PRTB r/fJeadsDHughesFacility File (EBarnhill) O-6 F-2ADAMS ACCESSION #: ML062430261TEMPLATE #:NRR-074OFFICEPRTB:CE IOLB:LAPRTB:SCNAMEKWitt:tls*EBarnhill*JEads:tls*DATE 8/31/2006 9/12/2006 9/12/2006OFFICIAL RECORD COPY U. S. NUCLEAR REGULATORY COMMISSIONOPERATOR LICENSING INITIAL EXAMINATION REPORTREPORT NO.:50-113/OL-06-01FACILITY DOCKET NO.:50-113 FACILITY LICENSE NO.:R-52 FACILITY:University of Arizona Research Reactor EXAMINATION DATE:August 15, 2006SUBMITTED BY:____/RA/______________

8/31/06Kevin Witt, Chief ExaminerDate

SUMMARY

During the week of August 15, 2006, the NRC administered an operator licensing examination to one Senior Reactor Operator Instant license candidate. The candidate passed the writtenand operating examinations.REPORT DETAILS1.Examiners: Kevin Witt, Chief Examiner2.Results:RO PASS/FAILSRO PASS/FAILTOTAL PASS/FAILWrittenN/A1/01/0Operating TestsN/A1/01/0 OverallN/A1/01/03.Exit Meeting:Kevin M. Witt, NRC Chief ExaminerRobert Offerle, Reactor Supervisor John G. Williams, Facility DirectorThe NRC thanked the facility staff for their cooperation during the administration of theexaminations. The NRC did not note any generic weaknesses on the part of thecandidates.ENCLOSURE 1 Facility Comments with NRC ResolutionQuestion B.013What is the lowest level of University of Arizona management who can authorize irradiation ofthe demountable fuel assembly in excess of 500 watt-minutes in one day?a.On-shift Reactor Operatorb.On-shift Senior Reactor Operator c.Reactor Laboratory Director d.Reactor CommitteeAnswer:B.013d

Reference:

UARR 177 Procedures for Use of the Demountable Fuel Element , §10, p.2Facilty CommentThis is not a useful memory test. Our operators are trained to follow written procedures, not tomemorize their minutia. This particular operation has not been performed in the last sixteen years, at least.NRC ResolutionComment accepted. This question has been deleted from the examination and will not factorinto the candidates' grades. This question will be modified before it is used again.Question C.011What is the MAXIMUM amount of time after a tank constant has been calculated, that it stillmay be used to calibrate the UARR Reactor, without being corrected for changes in pool water

depth?a. 5 daysb.10 days c.14 days d.30 daysAnswer:C.011b

Reference:

UARR 125 Procedure for Power Calibration of the University of ArizonaResearch ReactorFacilty CommentThere is no reason for an operator to remember this detail. Operators are trained to followwritten procedures.NRC ResolutionComment accepted. This question has been deleted from the examination and will not factorinto the candidates' grades. This question will be modified before it is used again.ENCLOSURE 2 UNIVERSITY OF ARIZONAWRITTEN EXAM w/ ANSWER KEYOPERATOR LICENSING EXAMINATIONAugust 15, 2006ENCLOSURE 3 Section A:

L Theory, Thermodynamics & Facility Operating CharacteristicsPage 1 of 19(***** Category A continued on next page *****)QUESTION:A.001[1.0 point]{1.0}With the reactor on a constant period, which transient requires the LONGEST time to occur? A reactor power change of:

a.5% power -- going from 1% to 6% power b.10% power -- going from 10% to 20% power c.15% power -- going from 20% to 35% power d.20% power -- going from 40% to 60% power QUESTION:A.002[1.0 point]{2.0}Which ONE of the following statements describes the difference between Differential andIntegral (IRW) rod worth curves?

a.DRW relates the worth of the rod per increment of movement to rod position. IRWrelates the total reactivity added by the rod to the rod position.

b.DRW relates the time rate of reactivity change to rod position. IRW relates the totalreactivity in the core to the time rate of reactivity change.

c.IRW relates the worth of the rod per increment of movement to rod position. DRWrelates the total reactivity added by the rod to the rod position.d.IRW is the slope of the DRW at a given rod position.QUESTIONA.003[1.0 point]{3.0}Which ONE of the following statements describes why installed neutron sources are used in reactor cores?a.To provide neutrons to initiate the chain reaction.b.To increase the count rate by an amount equal to the source contribution.c.To increase the count rate by an 1/M (M = Subcritical Multiplication Factor).d.To provide a neutron level high enough to be monitored by instrumentation.QUESTION:A.004[1.0 point]{4.0}Which ONE of the following elements has the highest thermal neutron absorptioncross-section? a.Uranium 235 b.Samarium 149 c.Boron 10 d.Xenon 135 Section A:

L Theory, Thermodynamics & Facility Operating CharacteristicsPage 2 of 19(***** Category A continued on next page *****)QUESTIONA.005[1.0 point]{5.0}At the beginning of a reactor startup Keff is 0.90 with a count rate of 30 cps. Power is increasedto a new, steady-state of 60 cps. The new Keff is:a.0.91b.0.925c.0.95d.0.974QUESTION:A.006[1.0 point]{6.0}The reactor is operating at 90 Kwatts (90%) and the scram setpoint is set at 110%. What willbe the resulting peak power if an experiment inserted into the reactor causes a 100 millisecondperiod and the scram delay is 0.1 second? Neglect Fuel Temperature Coeffecient effects. a.245 Kwattsb.280 Kwattsc.300 Kwattsd.320 KwattsQUESTION:A.007[1.0 point]{7.0}Which one of the following characteristics of a material would result in the most efficientthermalization of neutrons?

a.LOW atomic mass number and HIGH scattering cross section.

b.HIGH atomic mass number and LOW scattering cross section.

c.LOW neutron absorption and LOW scattering cross section.

d.LOW neutron absorption and HIGH atomic mass number.QUESTION:A.008[1.0 point]{8.0}Which ONE of the following is the time period in which the maximum amount of Xe 135 will bepresent in the core?a.7 to 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> after a scram from 100%.b.7 to 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> after a startup to 100% power.c.3 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a power increase from 50% to 100%.d.3 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a power decrease from 100% to 50%.

Section A:

L Theory, Thermodynamics & Facility Operating CharacteristicsPage 3 of 19(***** Category A continued on next page *****)QUESTION:A.009[1.0 point]{9.0}Which ONE of the following statements describes the Nuclear Instrumentation response for arod withdrawal while the reactor is subcritical? (Assume the reactor remains subcritical)a.Count rate will not change until criticality is reached.b.Count rate will rapidly increase (prompt jump) to a new stable value.c.Count rate will rapidly increase (prompt jump), then gradually decrease to the initialvalue. d.Count rate will rapidly increase (prompt jump), then gradually increase to a new stablevalue.QUESTION:A.010[1.0 point]{10.0}The prompt temperature coefficient of reactivity is -12.1 x 10

-5 K/K/C. When a control rodwith an average rod worth of 0.1% K/K/inch is withdrawn 12 inches, reactor power increasesand becomes stable at a higher level. Assuming the moderator temperature is constant, the fuel temperature has:a.increased by about 100Cb.decreased by about 100Cc.increased by about 10Cd.decreased by about 10CQUESTION:A.011[1.0 point]{11.0}Which factor in the six factor formula is represented by the ratio:number of neutrons that reach thermal energy number of neutrons that start to slow downa.fast non-leakage probability (L f)b.resonance escape probability (p)c.reproduction factor ()d.thermal utilization factor (f)

Section A:

L Theory, Thermodynamics & Facility Operating CharacteristicsPage 4 of 19(***** Category A continued on next page *****)QUESTION:A.012[1.0 point]{12.0}Which ONE of the following is true concerning the differences between prompt and delayedneutrons?a.Prompt neutrons are released during fast fissions while delayed neutrons are releasedduring thermal fissions.b.Prompt neutrons are released during the fission process while delayed neutrons arereleased during the decay process.c.Prompt neutrons are the dominating factor in determining the reactor period whiledelayed neutrons have little effect on the reactor period.d.Prompt neutrons account for less than 1% of the neutron population while delayedneutrons account for approximately 99% of the neutron population.QUESTION:A.013[1.0 point]{13.0}Which condition below describes a critical reactor?a.K = 1, K/K = 1b.K = 1, K/K = 0c.K = 0, K/K = 1d.K = 0, K/K = 0QUESTION:A.014[1.0 point]{14.0}In a reactor at full power, the thermal neutron flux (Ø) is 2.5 x 10 12 neutrons/cm 2/sec., and themacroscopic fission cross-section f is 0.1 cm

-1. The fission rate is:a.2.5 x 10 11 fissions/cm/sec.b.2.5 x 10 13 fissions/cm/sec.c.2.5 x 10 11 fissions/cm 3/sec.d.2.5 x 10 13 fissions/cm 3/sec.QUESTION:A.015[1.0 point]{15.0}

Keff differs from K in that Keff takes into account:a.leakage from the coreb.neutrons from fast fissionc.the effect of poisonsd.delayed neutrons Section A:

L Theory, Thermodynamics & Facility Operating CharacteristicsPage 5 of 19(***** Category A continued on next page *****)QUESTION:A.016[1.0 point]{16.0}Which ONE of the following parameter changes will require a control rod INSERTION tomaintain reactor power constant following the change?a.Samarium buildupb.Fuel Temperature Decreasesc.Xenon buildupd.U 235 concentration decrease (Fuel Burnup)QUESTION:A.017[1.0 point]{17.0}The period meter has just been replaced. For the first startup the Reactor Supervisor asked you to check the indication on the meter with actual period. During the startup you establishconditions which result in a power increase from 1 watt to 1 kilowatt in 1 minute 43 seconds.

What should the reactor period meter read?a.50b.30c.22d.15QUESTION:A.018[1.0 point]{18.0}A reactor has been operating at full power for one week when a scram occurs. Twelve hourslater, the reactor is brought critical and quickly raised to full power. Considering xenon effects only, to maintain a constant power level for the next few hours, control rods must be:a.insertedb.maintained at the present positionc.withdrawnd.withdrawn, then inserted to the original positionQUESTION:A.019[1.0 point]{19.0}Every fission of Uranium-235 produces an average of:a.2.00 neutronsb.2.07 neutronsc.2.42 neutronsd.2.87 neutrons Section A:

L Theory, Thermodynamics & Facility Operating CharacteristicsPage 6 of 19(***** End of Category A *****)QUESTION:A.020[1.0 point]{20.0}Which ONE of the following is the reason for the -80 second period following a reactor scram?a.The negative reactivity added during a scram is greater than beta-effective. b.The amount of negative reactivity added is greater than the Shutdown Margin.c.The half-life of the longest-lived group of delayed neutron precursors is approximately55 seconds.d.The fuel temperature coefficient adds positive reactivity as the fuel cools down, thusretarding the rate at which power drops.

Section B: Normal / Emergency Procedures & Radiological ControlsPage 7 of 19(***** Category B continued on next page *****)QUESTION:B.001[1.0 point]{1.0}A Limiting Safety System Setting has been exceeded. Which fuel elements are required to bemeasured for distortion?a.B- and C- ringsb.B- and D- ringsc.C- and D- ringsd.D- and E- ringsQUESTIONB.002[1.0 point]{2.0}An experiment is being removed from the pool after irradiation. The Health Physicist has you stop at four feet below the surface of the pool. A portable instrument indicates 5 mr/hr over background radiation. All of the reading is due to the experiment and the "tenth thickness" forwater is equal to 24 inches. WHICH ONE of the following is the expected dose at one foot after removal of the experiment from the pool?a.400 mr/hrb.1600 mr/hrc.2000 mr/hrd.8000 mr/hrQUESTIONB.003[1.0 point]{3.0}An individual receives 100 mRem of Beta (), 25 mRem of gamma (), and 5 mRem of neutronradiation. What is his/her total dose?a.275 mRemb.205 mRemc.175 mRemd.130 mRemQUESTIONB.004[1.0 point]{4.0}In accordance with 10 CFR 20, (no emergency exists) an individual in a restricted area may receive in excess of 1.25 rem/qtr when three conditions are met. Which ONE of the following is NOT a condition for exceeding the limit?a.An updated NRC form 4 is on record.b.Dose for the quarter will not exceed 3 REM.c.Cumulative dose rate will not exceed 5 (N-18) Rem.d.Personnel dosimeters must be read at double normal frequency.

Section B: Normal / Emergency Procedures & Radiological ControlsPage 8 of 19(***** Category B continued on next page *****)QUESTIONB.005[1.0 point]{5.0}Who has responsibility for establishing reentry requirements following an evacuation from thereactor laboratory?a.The Duty Health Physicistb.The Facility Directorc.The Radiation Control Directord.The Emergency DirectorQUESTIONB.006[1.0 point]{6.0}What is the maximum allowable dose which the facility director can authorize for a volunteer to receive to save the life of someone injured and trapped in the reactor compartment?a.125 Remb.100 Remc. 75 Remd. 50 RemQUESTIONB.007[1.0 point]{7.0}Which ONE of the following statements is the definition of a "Channel Test"?a.A combination of sensors, electronic circuits, and output devices connected by theappropriate network in order to measure and display the value of a parameter.b.The adjustment of the channel such that its output corresponds with acceptableaccuracy to known values of the parameter which the channel measures.c.A qualitative verification of acceptable performance by observation of channel behavior.

d.The introduction of a signal into the channel to verify that it is operable.QUESTIONB.008[1.0 point]{8.0}What is the lowest level of University of Arizona management required to be present duringmaintenance of the reactor control and safety system.a.Reactor Operatorb.Senior Reactor Operatorc.Reactor Supervisord.Facility Director Section B: Normal / Emergency Procedures & Radiological ControlsPage 9 of 19(***** Category B continued on next page *****)QUESTIONB.009[1.0 point]{9.0}Which ONE of the following statements correctly describes the relationship between the SafetyLimit (SL) and the Limiting Safety System Setting.a.The SL is a maximum operationally limiting value that prevents the LSSS from beingreached during normal runs.b.The LSSS is a maximum operationally limiting value that prevents the SL from beingreached during normal runs.c.The SL is a parameter that ensures the integrity of the fuel cladding. The LSSS initiatesprotective action to preclude reaching the SL.d.The LSSS is a parameter that ensures the integrity of the fuel cladding. The SL initiatesprotective action to preclude reaching the SL.QUESTIONB.010[1.0 point]{10.0}Which of the following experiments is NOT required to be doubly encapsulated?a.An experiment containing 23 milligrams of explosive materials. b.Fueled experiments containing 10 milligrams of liquid fissionable material.c.An experiment containing 20 milligrams of a material highly reactive with water.d.A fueled experiments with 1 millicurie total inventory of iodine isotopes (131 through135).QUESTIONB.011[1.0 point]{11.0}For spent fuel to be considered "self-shielding", the radiation level at 3 feet in air with nointervening shielding must be at LEAST 100 REM/hr. Assuming the average energy of radiation emitted by the spent fuel is 1 Mev, Select from the following the minimum activity required to meet the "self-shielding" limit for a fuel element.a. 10 curiesb. 15 curiesc.100 curiesd.150 curies Section B: Normal / Emergency Procedures & Radiological ControlsPage 10 of 19(***** Category B continued on next page *****)QUESTIONB.012[1.0 point]{12.0}A pool water loss exceeding _____________ which can not be stopped by the siphon breaks or by manually turning a valve or isolating the demineralizer constitutes "Notification of an Unusual Event?"a. 10 gallons/weekb. 10 gallons/hourc.100 gallons/hourd.100 gallons/minuteQUESTIONB.013[1.0 point]{13.0} Question deleted due to facility commentsWhat is the lowest level of University of Arizona management who can authorize irradiation of the demountable fuel assembly in excess of 500 watt-minutes in one day?a.On-shift Reactor Operatorb.On-shift Senior Reactor Operatorc.Reactor Laboratory Directord.Reactor CommitteeQUESTIONB.014[1.0 point]{14.0}Which ONE of the following is the correct Technical Specification Basis for limiting the reactivityworth of secured experiments to less than $1.00a.The sudden insertion or removal of an experiment of $3.00 or less will not cause the fueltemperature to increase by more than 75C when the reactor is operating at full power.b.The sudden insertion or removal of an experiment of $3.00 or less will not cause the fueltemperature to increase by more than 415C when the reactor is operating at full power.c.The sudden insertion or removal of an experiment of $3.00 or less will not cause the fueltemperature to exceed 450C when the reactor is operating at full power.d.The sudden insertion or removal of an experiment of $3.00 or less will not cause the fueltemperature to exceed 1000C when the reactor is operating at full power.

Section B: Normal / Emergency Procedures & Radiological ControlsPage 11 of 19(***** Category B continued on next page *****)QUESTION:B.015[1.0 point]{15.0}In accordance with the Technical Specifications, which ONE of the following defines an"Instrument Channel Check?"a.The introduction of a signal into a channel for verification that it is operable.

b.The qualitative verification of acceptable performance by observation of channelbehavior. c.A combination of sensors, electronic circuits and output devices which measure anddisplay the value of a parameter.d.The adjustment of a channel such that its output corresponds with acceptable accuracyto known values of the parameter which the channel measures.QUESTION:B.016[1.0 point]{16.0}In accordance with the Power Level Calibration Procedure, after power level is determined:a.The positions of the neutron detectors are adjusted to give the proper indicationb.The high voltages to the neutron detectors are adjusted to give the proper indicationc.The pointers on the power meters and recorders are adjusted to give the properindicationd.The compensating voltage of the compensated ion chamber is adjusted to give theproper indicationQUESTION:B.017[1.0 point]{17.0}To maintain an active Reactor Operator license, the functions of a reactor operator must be actively performed for at least:a.one hour per monthb.four hours per calendar quarterc.sixteen hours per yeard.three hours per calendar quarterQUESTION:B.018[1.0 point]{18.0}"Special Nuclear Material" is defined to be:a.Uranium-233, Uranium-235, or Uranium-238b.Plutonium, Uranium-238, or Thoriumc.Uranium-233, Uranium-235 or Thorium d.Uranium-233, Plutonium or enriched Uranium Section B: Normal / Emergency Procedures & Radiological ControlsPage 12 of 19(***** End of Category B *****)QUESTION:B.019[1.0 point]{19.0}Safety System channels which are required to be operable in all modes of operation are:a.reactor power level, reactor period, pool water levelb.reactor power level, manual scram, power failurec.reactor period, pool water level, manual scramd.power failure, pool water level, manual scramQUESTION:B.020[1.0 point]{20.0}A tour group of seven persons escorted by a reactor operator is about to enter the facility duringnormal operation. The minimum number of self-reading dosimeters that must be issued to the tour group is:a.0b.1c.2d.3 Section C: Facility and Radiation Monitoring SystemsPage 13 of 19(***** Category C continued on next page *****)QUESTIONC.001[1.0 point]{1.0}You've been assigned the task of starting up the reactor, and maintaining power at 100 kW. Atthe time you reach 100 kW the pool temperature is 27C. How long can you maintain power at100 kW before you must shutdown?a.21/4 hoursb.3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />sc.33/4 hoursd.41/2 hoursQUESTIONC.002[1.0 point]{2.0}Which ONE of the following conditions will prevent withdrawing the transient r od? a.Power level is 10 kW.b.A reactor period of 15 secondsc.Startup Count rate reads 2 counts/sec.d.Shock absorber anvil not fully inserted (pulse mode).QUESTIONC.003[2.0, (0.25 each)](4.0)Identify each lettered component on figure C - 1 using the correct component name from column B. (Items listed in column B may be used once, more than once, or not at all. Only one answer may occupy each space in column A.)

COLUMN ACOLUMN Ba._____1.Pneumatic Transfer Systemb._____2.Pulse Detector c._____3.Transient Rod d._____4.Regulating Rode._____5.Shim Rodf._____6.Left Safety Channel Detector g._____7.Right Safety Channel Detector h._____8.Neutron Source9.Wide Range Fission Chamber Detector 10.Linear Channel Detector Section C: Facility and Radiation Monitoring SystemsPage 14 of 19(***** Category C continued on next page *****)QUESTIONC.004[1.0 point]{5.0}Which ONE of following describes the automatic action(s) which occur on a high level trip of theparticulate air monitor?a.Shifts the stack fan from low to high speed.b.Shifts the stack fan exhaust for discharge through the "absolute filter."c.Switches off the stack fan and window fan and initiates a reactor room purge.d.Switches off the window-mounted exhaust fan and starts the stack exhaust fan.QUESTIONC.005[1.0 point]{6.0}Which of the following is the correct method used to detect neutrons in the Linear Power Channel detector?a.U 235 lining on inside of tube.b.B 10 lining on inside of tube.c.BF 3 gas.d.Be 11 lining on inside of tube.QUESTIONC.006[1.0 point]{7.0}Given the following indication for the shim rod: ROD/MAG UP LIGHT - OFF,ROD/MAG DOWN LIGHT - ON, CYL/CONT LIGHT ON, what is the condition of the shim rod and its drive?a.rod fully inserted, drive fully inb.rod fully in, drive intermediatec.rod between fully in and out, drive intermediated.can not be determined by lights given.QUESTIONC.007[1.0 point]{8.0}How is water or condensation removed from the rotary specimen rack (Lazy Susan)?a.The pool is periodically drained.b. An inert gas is inserted into the rack to blow out the condensation.c.An electric heater is placed in an insulated specimen tube, which is inserted into therack.d.Water absorbing material is placed into a perforated specimen tube, which is insertedinto the rack.

Section C: Facility and Radiation Monitoring SystemsPage 15 of 19(***** Category C continued on next page *****)QUESTIONC.008[1.0 point]{9.0}Using the figure C - 2, which valve lineup is correct for returning a rabbit from the reactor?a.Valves 1 and 2 open, 3 and 4 shut.b.Valves 1 and 3 open, 2 and 4 shut.c.Valves 2 and 4 open, 1 and 3 shut.d.Valves 2 and 3 open, 1 and 4 shut.QUESTION:C.009[1.0 point]{10.0}The TRIGA fuel elements consist of:a.70% enriched uranium with stainless steel cladb.20% enriched uranium with stainless steel cladc.70% enriched uranium with aluminum cladd.20% enriched uranium with aluminum cladQUESTIONC.010[1.0 point]{11.0}Period information is supplied from the:a.Wide Range Log Channelb.Linear Channelc.Right Safety Channeld.Left Safety ChannelQUESTIONC.011[1.0 point]{12.0} Question deleted due to facility commentsWhat is the MAXIMUM amount of time after a tank constant has been calculated, that it stillmay be used to calibrate the UARR Reactor, without being corrected for changes in pool water

depth?a. 5 daysb.10 daysc.14 daysd.30 days Section C: Facility and Radiation Monitoring SystemsPage 16 of 19(***** Category C continued on next page *****)QUESTIONC.012[1.0 point]{13.0}Which of the following gases is one you would expect to detect following a fuel element failure?a.Ar 41b.Kr 88c.H 3d.N 16QUESTIONC.013[1.0 point]{14.0}Which one of the following conditions would NOT require the reactor to be shutdown and thedemineralizer turned off?a.Pool conductivity exceeds T.S. limit.b.Release of fission products from a fuel element.c.Pool activity exceeds the water activity monitor setpoint.d.Dropping an irradiated water soluble sample into the pool.QUESTIONC.014[1.0 point]{15.0}What is the total worth of all three control rods?a.$2.84b.$6.78c.$9.48d.$11.55QUESTIONC.015[1.0 point]{16.0}During a $2.00 pulse, what would you expect as a maximum core temperature?a.156Cb.193Cc.235Cd.277C Section C: Facility and Radiation Monitoring SystemsPage 17 of 19(***** Category C continued on next page *****)QUESTION:C.016[1.0 point]{17.0}In the automatic mode, the controlling signal is:a.reactor power as measured by the wide range log channelb.reactor power as measured by the right safety channelc.reactor period as measured by the left safety channeld.reactor power as measured by the linear channelQUESTION:C.017[1.0 point]{18.0}Which ONE condition below will NOT result in a scram?a.Power level of 110 kWb.Pool water level of 13 feetc.Reactor period of 2 secondsd.Safety channel switched to "calibrate"QUESTION:C.018[1.0 point]{19.0}Upon the receipt of a scram signal, the regulating blade:a.automatically drives into the coreb.magnet and drive both fall into the corec.magnet is de-energized, and the blade falls into the cored.remains where it is, and must be manually driven into the coreQUESTION:C.019[1.0 point]{20.0}Significant amounts of Argon-41 are produced in the pool water, the pneumatic transfer system, and the:a.neutron thermalizerb.fast irradiation facilityc.rotary specimen rackd.neutron radiography beam tube Section C: Facility and Radiation Monitoring SystemsPage 18 of 19(***** Category C continued on next page *****)Figure C-1 Section C: Facility and Radiation Monitoring SystemsPage 19 of 19(***** End of Category C *****)Figure C-2 Section C: Facility and Radiation Monitoring SystemsPage 20 of 19(***** End of Category C *****)

Section A:

L Theory, Thermodynamics & Facility Operating CharacteristicsANSWERSAnswer:A.001a.

Reference:

Lamarsh, Introduction to Nuclear Engineering, 1975, Page 249Answer:A.002a.

Reference:

Lamarsh, Introduction to Nuclear Engineering, 1975, Page 270Answer:A.003d.

Reference:

Glasstone, S. and Sesonske, A, Nuclear Reactor Engineering, Kreiger Publishing, Malabar, Florida, 1991, Section 5.286Answer:A.004d.

Reference:

Lamarsh, Introduction to Nuclear Engineering, 1975, App. IIAnswer:A.005c.

Reference:

Intro to Nuc Eng, John R. Lamarsh © 1983, § p.Answer:A.006c.

Reference:

Intro to Nuc Eng, John R. Lamarsh © 1983, § 7.1, Example 7.6, p. 289.Standard Power equationAnswer:A.007a.

Reference:

Intro to Nuc Eng, John R. Lamarsh © 1983, § 3.5, pp. 59-60.Answer:A.008a.

Reference:

Intro to Nuc Eng, John R. Lamarsh © 1983, § 7.4 Figure 7.13, p. 322.Answer:A.009d.

Reference:

Intro to Nuc Eng, John R. Lamarsh © 1983, § 7.1, pp. 286-258.Answer:A.010a.

Reference:

Intro to Nuc Eng, John R. Lamarsh © 1983, § 7.3, p. 308-312.Reactivity added by control rod = +(0.001K/K/inch) (12 inches) =

0.01K/K.Fuel temperature change = - Reactivity added by rod ÷ fuel temp coeff.

Fuel temp. change = (-0.012K/K) ÷ (-1.21 x 10

-4K/K/C) = 99.2CAnswer:A.011b.

Reference:

Intro to Nuc Eng, John R. Lamarsh © 1983, § 6.5 p. 239.Answer:A.012b.

Reference:

Intro to Nuc Eng, John R. Lamarsh © 1983, § 3.7 pp. 73 - 75.Answer:A.013b.

Reference:

Intro to Nuc Eng, John R. Lamarsh © 1983, § 7.1, p. 282.Answer:A.014c.

Reference:

R = Ø f = (2.5 x 10

12) x 0.1 = 2.5 x 10 11Answer:A.015a.

Reference:

Intro to Nuc Eng, John R. Lamarsh © 1983, § 6.5, p. 239.Answer:A.016b.

Reference:

Intro to Nuc Eng, John R. Lamarsh © 1983, § 7.3 & 7.4, pp. 312-314 &316-327.

Section A:

L Theory, Thermodynamics & Facility Operating CharacteristicsANSWERSAnswer:A.017d.

Reference:

P = Poet/TAnswer:A.018a.

Reference:

Lamarsh, Introduction to Nuclear Engineering, 1975, Page 289Answer:A.019c.

Reference:

Lamarsh, Introduction to Nuclear Engineering, 1975, Page 68Answer:A.020c.

Reference:

Lamarsh, Introduction to Nuclear Engineering, 1975, Page 255 Section B: Normal / Emergency Procedures & Radiological Controls ANSWERSAnswer:B.001a.

Reference:

UARR Technical Specification 4.1.d p. 17Answer:B.002d

Reference:

Radiological Health Handbook Revised January 1970.Answer:B.003d.

Reference:

10 CFR 20.4(c)Answer:B.004.d

Reference:

10 CFR 20.101.b(1), (2) and (3)Answer:B.005c.

Reference:

UARR Emergency Plan, § 3.4, page 11.Answer:B.006b.

Reference:

UARR Emergency Plan § 3.5 Authorization of Radiation Exposures inExcess of 10CFR Limits, page 11.Answer:B.007d.

Reference:

UARR Technical Specifications § 1.0 DEFINITIONS , p. 1.Answer:B.008a.

Reference:

UARR Technical Specifications § 4.5(c) Maintenance, p. 21. Also, UARROperating Procedure UARR100 § 2.3(2) p. 7.Answer:B.009c.

Reference:

UARR Technical Specifications § 1.0 DefinitionsAnswer:B.010d.

Reference:

UARR Technical Specifications, § 3.7, Experiments , p. 16Answer:B.011d.

Reference:

Radiological Health Handbook Revised January 1970.Answer:B.012c.

Reference:

UARR 114 Procedure for Responding to suspected Primary CoolantLeaks § 3, Answer:B.013d. Question deleted

Reference:

UARR 117 Procedures for Use of the Demountable Fuel Element , § 10, p. 2Answer:B.014b.

Reference:

UARR Tech Specs § 3.1 bases. Answer:B.015b.

Reference:

UARR Technical Specifications, Section 1.0Answer:B.016 a.

Reference:

UARR Procedure 125Answer:B.017b.

Reference:

10 CFR 55.53 Section B: Normal / Emergency Procedures & Radiological Controls ANSWERSAnswer:B.018d.

Reference:

UARR Procedure 124Answer:B.019d.

Reference:

UARR Technical Specifications, Section 3.5Answer:B.020d.

Reference:

UARR Procedure 100 Section C: Facility and Radiation Monitoring SystemsANSWERSAnswer:C.001d.Maximum Temperature: = 45C.Heatup rate: = 4C/Hr.Present Temperature: = 27C.Therefore: (45C - 27C) ÷ 4C/Hr=18/4 Hrs. = 41/2 hours

Reference:

UARR Safety Analysis Report, p. 29.Answer:C.002a.

Reference:

Univ. of Arizona Research L, Technical Specifications § 3.5 page 13.Answer:C.003a. = 2;b. = 3;c. = 6;d. = 4;e, = 9;f, = 5;g. = 8;h. = 1

Reference:

UARR Safety Analysis Report, p. 15. UARR Core positiondiagram. Answer:C.004d.

Reference:

UARR TRIGA L Description, Ventilation SystemAnswer:C.005b.

Reference:

UARR Reference T-63 Reactor Console pgs 5 - 8. Answer:C.006a.

Reference:

UARR Safety Analysis Report Figure 6.3.Answer:C.007d.

Reference:

TRIGA Mark I L Mechanical Maintenance and Operating Manual.Answer:C.008c

Reference:

UARR TRIGA L Description, Irradiation Facilities.Answer:C.009b.

Reference:

UARR Safety Analysis Report, Page 11Answer:C.010a.

Reference:

UARR Safety Analysis Report, Page 39Answer:C.011b. Question deleted

Reference:

UUAR125 Procedure for Power Calibration of the University of ArizonaResearch L Triga LAnswer:C.012b.

Reference:

UARR SAR § Release of Fission Products from a fuel element.

p. 57Answer:C.013a.

Reference:

UARR100, p. 10,Answer:C.014c.

Reference:

UARR SAR, Table 3.1Answer:C.015c.

Reference:

UARR SAR Table 3.2.Answer:C.016d.

Reference:

UARR Safety Analysis Report, Page 39 Section C: Facility and Radiation Monitoring SystemsANSWERSAnswer:C.017c.

Reference:

UARR Safety Analysis Report, Page 45Answer:C.018c.

Reference:

UARR Safety Analysis Report, Page 44Answer:C.019c.

Reference:

UARR Safety Analysis Report, Page 62

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