ML062430261

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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ML061230272	List: ML062430261 ML062430274
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OL-06-01
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September 12, 2006 Dr. Leslie Tolbert Vice President for Research University of Arizona Tucson, AZ 85721-0066

SUBJECT:

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

Dear Dr. Tolbert:

During the week of August 15, 2006, the NRC administered an operator licensing examination at 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 the enclosures will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's Agencywide Documents 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 be released 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, Chief Research and Test Reactors Branch B Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Docket No. 50-113

Enclosures:

1. Initial Examination Report No. 50-113/OL-06-01

2. Facility comments with NRC resolution

3. Examination and answer key cc w/encls:

Please see next page

University of Arizona Docket No. 50-113 cc w/encl:

Office of the Mayor P.O. Box 27210 Tucson, AZ 85726-7210 Arizona Radiation Regulatory Agency 4814 S. 40th Street Phoenix, AZ 85040 University of Arizona Nuclear Research Laboratory ATTN: Dr. John Williams, Reactor Director Bldg. 20, Rm 200 Tucson, AZ 85721-0020 University of Arizona Nuclear Research Laboratory ATTN: Robert Offerle, Reactor Supervisor Bldg. 20, Rm. 200 Tucson, AZ 85721-0020 University of Arizona ATTN: Dr. Caroline M. Garcia Assistant Director, Arizona Research Labs Gould-Simpson Bldg. 1011 P.O. Box 210077 Tucson, AZ 85721-0077 University of Arizona ATTN: Daniel Silvain, Radiation Safety Officer 1640 North Vine Tucson, AZ 85721-0020 Test, Research and Training Reactor Newsletter 202 Nuclear Sciences Center University of Florida Gainesville, FL 32611

September 12, 2006 Dr. Leslie Tolbert Vice President for Research University of Arizona Tucson, AZ 85721-0066

SUBJECT:

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

Dear Dr. Tolbert:

During the week of August 15, 2006, the NRC administered an operator licensing examination at 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 the enclosures will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's Agencywide Documents 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 be released 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, Chief Research and Test Reactors Branch B Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Docket No. 50-113

Enclosures:

1. Initial Examination Report No. 50-113/OL-06-01

2. Facility comments with NRC resolution

3. Examination and answer key cc w/encls:

Please see next page DISTRIBUTION w/ encls.:

PUBLIC PRTB r/f Jeads DHughes Facility File (EBarnhill) O-6 F-2 ADAMS ACCESSION #: ML062430261 TEMPLATE #:NRR-074 OFFICE PRTB:CE IOLB:LA PRTB:SC NAME KWitt:tls*

EBarnhill*

JEads:tls*

DATE 8/31/2006 9/12/2006 9/12/2006 OFFICIAL RECORD COPY

U. S. NUCLEAR REGULATORY COMMISSION OPERATOR LICENSING INITIAL EXAMINATION REPORT REPORT NO.:

50-113/OL-06-01 FACILITY DOCKET NO.:

50-113 FACILITY LICENSE NO.:

R-52 FACILITY:

University of Arizona Research Reactor EXAMINATION DATE:

August 15, 2006 SUBMITTED BY:

____/RA/______________

8/31/06 Kevin Witt, Chief Examiner Date

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 written and operating examinations.

REPORT DETAILS 1.

Examiners:

Kevin Witt, Chief Examiner 2.

Results:

RO PASS/FAIL SRO PASS/FAIL TOTAL PASS/FAIL Written N/A 1/0 1/0 Operating Tests N/A 1/0 1/0 Overall N/A 1/0 1/0 3.

Exit Meeting:

Kevin M. Witt, NRC Chief Examiner Robert Offerle, Reactor Supervisor John G. Williams, Facility Director The NRC thanked the facility staff for their cooperation during the administration of the examinations. The NRC did not note any generic weaknesses on the part of the candidates.

ENCLOSURE 1

Facility Comments with NRC Resolution Question B.013 What 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 Operator b.

On-shift Senior Reactor Operator c.

Reactor Laboratory Director d.

Reactor Committee Answer:

B.013 d

Reference:

UARR 177 Procedures for Use of the Demountable Fuel Element, §10, p.2 Facilty Comment This is not a useful memory test. Our operators are trained to follow written procedures, not to memorize their minutia. This particular operation has not been performed in the last sixteen years, at least.

NRC Resolution Comment accepted. This question has been deleted from the examination and will not factor into the candidates grades. This question will be modified before it is used again.

Question C.011 What is the MAXIMUM amount of time after a tank constant has been calculated, that it still may be used to calibrate the UARR Reactor, without being corrected for changes in pool water depth?

a.

5 days b.

10 days c.

14 days d.

30 days Answer:

C.011 b

Reference:

UARR 125 Procedure for Power Calibration of the University of Arizona Research Reactor Facilty Comment There is no reason for an operator to remember this detail. Operators are trained to follow written procedures.

NRC Resolution Comment accepted. This question has been deleted from the examination and will not factor into the candidates grades. This question will be modified before it is used again.

ENCLOSURE 2

UNIVERSITY OF ARIZONA WRITTEN EXAM w/ ANSWER KEY OPERATOR LICENSING EXAMINATION August 15, 2006 ENCLOSURE 3

Section A: L Theory, Thermodynamics & Facility Operating Characteristics Page 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 and Integral (IRW) rod worth curves?

a.

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

d.

IRW is the slope of the DRW at a given rod position.

QUESTION A.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 absorption cross-section?

a.

Uranium 235 b.

Samarium 149 c.

Boron 10 d.

Xenon 135

Section A: L Theory, Thermodynamics & Facility Operating Characteristics Page 2 of 19

(***** Category A continued on next page *****)

QUESTION A.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 increased to a new, steady-state of 60 cps. The new Keff is:

a.

0.91 b.

0.925 c.

0.95 d.

0.974 QUESTION:

A.006

[1.0 point]

{6.0}

The reactor is operating at 90 Kwatts (90%) and the scram setpoint is set at 110%. What will be the resulting peak power if an experiment inserted into the reactor causes a 100 millisecond period and the scram delay is 0.1 second? Neglect Fuel Temperature Coeffecient effects.

a.

245 Kwatts b.

280 Kwatts c.

300 Kwatts d.

320 Kwatts QUESTION:

A.007

[1.0 point]

{7.0}

Which one of the following characteristics of a material would result in the most efficient thermalization 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 Xe135 will be present 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 Characteristics Page 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 a rod 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 initial value.

d.

Count rate will rapidly increase (prompt jump), then gradually increase to a new stable value.

QUESTION:

A.010

[1.0 point]

{10.0}

The prompt temperature coefficient of reactivity is -12.1 x 10-5 K/K/EC. When a control rod with an average rod worth of 0.1% K/K/inch is withdrawn 12 inches, reactor power increases and becomes stable at a higher level. Assuming the moderator temperature is constant, the fuel temperature has:

a.

increased by about 100EC b.

decreased by about 100EC c.

increased by about 10EC d.

decreased by about 10EC QUESTION:

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 down a.

fast non-leakage probability (Lf) b.

resonance escape probability (p) c.

reproduction factor ()

d.

thermal utilization factor (f)

Section A: L Theory, Thermodynamics & Facility Operating Characteristics Page 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 delayed neutrons?

a.

Prompt neutrons are released during fast fissions while delayed neutrons are released during thermal fissions.

b.

Prompt neutrons are released during the fission process while delayed neutrons are released during the decay process.

c.

Prompt neutrons are the dominating factor in determining the reactor period while delayed neutrons have little effect on the reactor period.

d.

Prompt neutrons account for less than 1% of the neutron population while delayed neutrons 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 = 1 b.

K = 1, K/K = 0 c.

K = 0, K/K = 1 d.

K = 0, K/K = 0 QUESTION:

A.014

[1.0 point]

{14.0}

In a reactor at full power, the thermal neutron flux (Ø) is 2.5 x 1012 neutrons/cm2/sec., and the macroscopic fission cross-section f is 0.1 cm-1. The fission rate is:

a.

2.5 x 1011 fissions/cm/sec.

b.

2.5 x 1013 fissions/cm/sec.

c.

2.5 x 1011 fissions/cm3/sec.

d.

2.5 x 1013 fissions/cm3/sec.

QUESTION:

A.015

[1.0 point]

{15.0}

Keff differs from K4 in that Keff takes into account:

a.

leakage from the core b.

neutrons from fast fission c.

the effect of poisons d.

delayed neutrons

Section A: L Theory, Thermodynamics & Facility Operating Characteristics Page 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 to maintain reactor power constant following the change?

a.

Samarium buildup b.

Fuel Temperature Decreases c.

Xenon buildup d.

U235 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 establish conditions 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.

50 b.

30 c.

22 d.

15 QUESTION:

A.018

[1.0 point]

{18.0}

A reactor has been operating at full power for one week when a scram occurs. Twelve hours later, 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.

inserted b.

maintained at the present position c.

withdrawn d.

withdrawn, then inserted to the original position QUESTION:

A.019

[1.0 point]

{19.0}

Every fission of Uranium-235 produces an average of:

a.

2.00 neutrons b.

2.07 neutrons c.

2.42 neutrons d.

2.87 neutrons

Section A: L Theory, Thermodynamics & Facility Operating Characteristics Page 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 approximately 55 seconds.

d.

The fuel temperature coefficient adds positive reactivity as the fuel cools down, thus retarding the rate at which power drops.

Section B: Normal / Emergency Procedures & Radiological Controls Page 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 be measured for distortion?

a.

B-and C-rings b.

B-and D-rings c.

C-and D-rings d.

D-and E-rings QUESTION B.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" for water 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/hr b.

1600 mr/hr c.

2000 mr/hr d.

8000 mr/hr QUESTION B.003

[1.0 point]

{3.0}

An individual receives 100 mRem of Beta (), 25 mRem of gamma (), and 5 mRem of neutron radiation. What is his/her total dose?

a.

275 mRem b.

205 mRem c.

175 mRem d.

130 mRem QUESTION B.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 Controls Page 8 of 19

(***** Category B continued on next page *****)

QUESTION B.005

[1.0 point]

{5.0}

Who has responsibility for establishing reentry requirements following an evacuation from the reactor laboratory?

a.

The Duty Health Physicist b.

The Facility Director c.

The Radiation Control Director d.

The Emergency Director QUESTION B.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 Rem b.

100 Rem c.

75 Rem d.

50 Rem QUESTION B.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 the appropriate network in order to measure and display the value of a parameter.

b.

The adjustment of the channel such that its output corresponds with acceptable accuracy 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.

QUESTION B.008

[1.0 point]

{8.0}

What is the lowest level of University of Arizona management required to be present during maintenance of the reactor control and safety system.

a.

Reactor Operator b.

Senior Reactor Operator c.

Reactor Supervisor d.

Facility Director

Section B: Normal / Emergency Procedures & Radiological Controls Page 9 of 19

(***** Category B continued on next page *****)

QUESTION B.009

[1.0 point]

{9.0}

Which ONE of the following statements correctly describes the relationship between the Safety Limit (SL) and the Limiting Safety System Setting.

a.

The SL is a maximum operationally limiting value that prevents the LSSS from being reached during normal runs.

b.

The LSSS is a maximum operationally limiting value that prevents the SL from being reached during normal runs.

c.

The SL is a parameter that ensures the integrity of the fuel cladding. The LSSS initiates protective action to preclude reaching the SL.

d.

The LSSS is a parameter that ensures the integrity of the fuel cladding. The SL initiates protective action to preclude reaching the SL.

QUESTION B.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 through 135).

QUESTION B.011

[1.0 point]

{11.0}

For spent fuel to be considered "self-shielding", the radiation level at 3 feet in air with no intervening 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 curies b.

15 curies c.

100 curies d.

150 curies

Section B: Normal / Emergency Procedures & Radiological Controls Page 10 of 19

(***** Category B continued on next page *****)

QUESTION B.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/week b.

10 gallons/hour c.

100 gallons/hour d.

100 gallons/minute QUESTION B.013

[1.0 point]

{13.0} Question deleted due to facility comments What 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 Operator b.

On-shift Senior Reactor Operator c.

Reactor Laboratory Director d.

Reactor Committee QUESTION B.014

[1.0 point]

{14.0}

Which ONE of the following is the correct Technical Specification Basis for limiting the reactivity worth of secured experiments to less than $1.00 a.

The sudden insertion or removal of an experiment of $3.00 or less will not cause the fuel temperature to increase by more than 75EC 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 fuel temperature to increase by more than 415EC 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 fuel temperature to exceed 450EC 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 fuel temperature to exceed 1000EC when the reactor is operating at full power.

Section B: Normal / Emergency Procedures & Radiological Controls Page 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 channel behavior.

c.

A combination of sensors, electronic circuits and output devices which measure and display the value of a parameter.

d.

The adjustment of a channel such that its output corresponds with acceptable accuracy to 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 indication b.

The high voltages to the neutron detectors are adjusted to give the proper indication c.

The pointers on the power meters and recorders are adjusted to give the proper indication d.

The compensating voltage of the compensated ion chamber is adjusted to give the proper indication QUESTION:

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 month b.

four hours per calendar quarter c.

sixteen hours per year d.

three hours per calendar quarter QUESTION:

B.018

[1.0 point]

{18.0}

"Special Nuclear Material" is defined to be:

a.

Uranium-233, Uranium-235, or Uranium-238 b.

Plutonium, Uranium-238, or Thorium c.

Uranium-233, Uranium-235 or Thorium d.

Uranium-233, Plutonium or enriched Uranium

Section B: Normal / Emergency Procedures & Radiological Controls Page 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 level b.

reactor power level, manual scram, power failure c.

reactor period, pool water level, manual scram d.

power failure, pool water level, manual scram QUESTION:

B.020

[1.0 point]

{20.0}

A tour group of seven persons escorted by a reactor operator is about to enter the facility during normal operation. The minimum number of self-reading dosimeters that must be issued to the tour group is:

a.

0 b.

1 c.

2 d.

3

Section C: Facility and Radiation Monitoring Systems Page 13 of 19

(***** Category C continued on next page *****)

QUESTION C.001

[1.0 point]

{1.0}

You've been assigned the task of starting up the reactor, and maintaining power at 100 kW. At the time you reach 100 kW the pool temperature is 27EC. How long can you maintain power at 100 kW before you must shutdown?

a.

21/4 hours b.

3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> c.

33/4 hours d.

41/2 hours QUESTION C.002

[1.0 point]

{2.0}

Which ONE of the following conditions will prevent withdrawing the transient rod?

a.

Power level is 10 kW.

b.

A reactor period of 15 seconds c.

Startup Count rate reads 2 counts/sec.

d.

Shock absorber anvil not fully inserted (pulse mode).

QUESTION C.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 A COLUMN B a.

1.

Pneumatic Transfer System b.

2.

Pulse Detector c.

3.

Transient Rod d.

4.

Regulating Rod e.

5.

Shim Rod f.

6.

Left Safety Channel Detector g.

7.

Right Safety Channel Detector h.

8.

Neutron Source 9.

Wide Range Fission Chamber Detector

10. Linear Channel Detector

Section C: Facility and Radiation Monitoring Systems Page 14 of 19

(***** Category C continued on next page *****)

QUESTION C.004

[1.0 point]

{5.0}

Which ONE of following describes the automatic action(s) which occur on a high level trip of the particulate 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.

QUESTION C.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.

U235 lining on inside of tube.

b.

B10 lining on inside of tube.

c.

BF3 gas.

d.

Be11 lining on inside of tube.

QUESTION C.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 in b.

rod fully in, drive intermediate c.

rod between fully in and out, drive intermediate d.

can not be determined by lights given.

QUESTION C.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 the rack.

d.

Water absorbing material is placed into a perforated specimen tube, which is inserted into the rack.

Section C: Facility and Radiation Monitoring Systems Page 15 of 19

(***** Category C continued on next page *****)

QUESTION C.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 clad b.

20% enriched uranium with stainless steel clad c.

70% enriched uranium with aluminum clad d.

20% enriched uranium with aluminum clad QUESTION C.010

[1.0 point]

{11.0}

Period information is supplied from the:

a.

Wide Range Log Channel b.

Linear Channel c.

Right Safety Channel d.

Left Safety Channel QUESTION C.011

[1.0 point]

{12.0} Question deleted due to facility comments What is the MAXIMUM amount of time after a tank constant has been calculated, that it still may be used to calibrate the UARR Reactor, without being corrected for changes in pool water depth?

a.

5 days b.

10 days c.

14 days d.

30 days

Section C: Facility and Radiation Monitoring Systems Page 16 of 19

(***** Category C continued on next page *****)

QUESTION C.012

[1.0 point]

{13.0}

Which of the following gases is one you would expect to detect following a fuel element failure?

a.

Ar41 b.

Kr88 c.

H3 d.

N16 QUESTION C.013

[1.0 point]

{14.0}

Which one of the following conditions would NOT require the reactor to be shutdown and the demineralizer 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.

QUESTION C.014 [1.0 point]

{15.0}

What is the total worth of all three control rods?

a.

$2.84 b.

$6.78 c.

$9.48 d.

$11.55 QUESTION C.015

[1.0 point]

{16.0}

During a $2.00 pulse, what would you expect as a maximum core temperature?

a.

156EC b.

193EC c.

235EC d.

277EC

Section C: Facility and Radiation Monitoring Systems Page 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 channel b.

reactor power as measured by the right safety channel c.

reactor period as measured by the left safety channel d.

reactor power as measured by the linear channel QUESTION:

C.017

[1.0 point]

{18.0}

Which ONE condition below will NOT result in a scram?

a.

Power level of 110 kW b.

Pool water level of 13 feet c.

Reactor period of 2 seconds d.

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 core b.

magnet and drive both fall into the core c.

magnet is de-energized, and the blade falls into the core d.

remains where it is, and must be manually driven into the core QUESTION:

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 thermalizer b.

fast irradiation facility c.

rotary specimen rack d.

neutron radiography beam tube

Section C: Facility and Radiation Monitoring Systems Page 18 of 19

(***** Category C continued on next page *****)

Figure C-1

Section C: Facility and Radiation Monitoring Systems Page 19 of 19

(***** End of Category C *****)

Figure C-2

Section C: Facility and Radiation Monitoring Systems Page 20 of 19

(***** End of Category C *****)

Section A: L Theory, Thermodynamics & Facility Operating Characteristics ANSWERS Answer:

A.001 a.

Reference:

Lamarsh, Introduction to Nuclear Engineering, 1975, Page 249 Answer:

A.002 a.

Reference:

Lamarsh, Introduction to Nuclear Engineering, 1975, Page 270 Answer:

A.003 d.

Reference:

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

A.004 d.

Reference:

Lamarsh, Introduction to Nuclear Engineering, 1975, App. II Answer:

A.005 c.

Reference:

Intro to Nuc Eng, John R. Lamarsh © 1983, § p.

Answer:

A.006 c.

Reference:

Intro to Nuc Eng, John R. Lamarsh © 1983, § 7.1, Example 7.6, p. 289.

Standard Power equation Answer:

A.007 a.

Reference:

Intro to Nuc Eng, John R. Lamarsh © 1983, § 3.5, pp. 59-60.

Answer:

A.008 a.

Reference:

Intro to Nuc Eng, John R. Lamarsh © 1983, § 7.4 Figure 7.13, p. 322.

Answer:

A.009 d.

Reference:

Intro to Nuc Eng, John R. Lamarsh © 1983, § 7.1, pp. 286-258.

Answer:

A.010 a.

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/EC) = 99.2EC Answer:

A.011 b.

Reference:

Intro to Nuc Eng, John R. Lamarsh © 1983, § 6.5 p. 239.

Answer:

A.012 b.

Reference:

Intro to Nuc Eng, John R. Lamarsh © 1983, § 3.7 pp. 73 - 75.

Answer:

A.013 b.

Reference:

Intro to Nuc Eng, John R. Lamarsh © 1983, § 7.1, p. 282.

Answer:

A.014 c.

Reference:

R = Ø f = (2.5 x 1012) x 0.1 = 2.5 x 1011 Answer:

A.015 a.

Reference:

Intro to Nuc Eng, John R. Lamarsh © 1983, § 6.5, p. 239.

Answer:

A.016 b.

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 Characteristics ANSWERS Answer:

A.017 d.

Reference:

P = Poet/T Answer:

A.018 a.

Reference:

Lamarsh, Introduction to Nuclear Engineering, 1975, Page 289 Answer:

A.019 c.

Reference:

Lamarsh, Introduction to Nuclear Engineering, 1975, Page 68 Answer:

A.020 c.

Reference:

Lamarsh, Introduction to Nuclear Engineering, 1975, Page 255

Section B: Normal / Emergency Procedures & Radiological Controls ANSWERS Answer:

B.001 a.

Reference:

UARR Technical Specification 4.1.d p. 17 Answer:

B.002 d

Reference:

Radiological Health Handbook Revised January 1970.

Answer:

B.003 d.

Reference:

10 CFR 20.4(c)

Answer:

B.004.d

Reference:

10 CFR 20.101.b(1), (2) and (3)

Answer:

B.005 c.

Reference:

UARR Emergency Plan, § 3.4, page 11.

Answer:

B.006 b.

Reference:

UARR Emergency Plan § 3.5 Authorization of Radiation Exposures in Excess of 10CFR Limits, page 11.

Answer:

B.007 d.

Reference:

UARR Technical Specifications § 1.0 DEFINITIONS, p. 1.

Answer:

B.008 a.

Reference:

UARR Technical Specifications § 4.5(c) Maintenance, p. 21. Also, UARR Operating Procedure UARR100 § 2.3(2) p. 7.

Answer:

B.009 c.

Reference UARR Technical Specifications § 1.0 Definitions Answer:

B.010 d.

Reference:

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

B.011 d.

Reference:

Radiological Health Handbook Revised January 1970.

Answer:

B.012 c.

Reference:

UARR 114 Procedure for Responding to suspected Primary Coolant Leaks § 3, Answer:

B.013 d. Question deleted

Reference:

UARR 117 Procedures for Use of the Demountable Fuel Element, § 10,

p. 2 Answer:

B.014 b.

Reference:

UARR Tech Specs § 3.1 bases.

Answer:

B.015 b.

Reference:

UARR Technical Specifications, Section 1.0 Answer:

B.016 a.

Reference:

UARR Procedure 125 Answer:

B.017 b.

Reference:

10 CFR 55.53

Section B: Normal / Emergency Procedures & Radiological Controls ANSWERS Answer:

B.018 d.

Reference:

UARR Procedure 124 Answer:

B.019 d.

Reference:

UARR Technical Specifications, Section 3.5 Answer:

B.020 d.

Reference:

UARR Procedure 100

Section C: Facility and Radiation Monitoring Systems ANSWERS Answer:

C.001 d.

Maximum Temperature: = 45EC.

Heatup rate: = 4EC/Hr.

Present Temperature: = 27EC.

Therefore: (45EC - 27EC) ÷ 4EC/Hr

=

18/4 Hrs. = 41/2 hours

Reference:

UARR Safety Analysis Report, p. 29.

Answer:

C.002 a.

Reference:

Univ. of Arizona Research L, Technical Specifications § 3.5 page 13.

Answer:

C.003 a. = 2; b. = 3; c. = 6; d. = 4; e, = 9; f, = 5; g. = 8; h. = 1

Reference:

UARR Safety Analysis Report, p. 15. UARR Core position diagram.

Answer:

C.004 d.

Reference:

UARR TRIGA L Description, Ventilation System Answer:

C.005 b.

Reference:

UARR Reference T-63 Reactor Console pgs 5 - 8.

Answer:

C.006 a.

Reference:

UARR Safety Analysis Report Figure 6.3.

Answer:

C.007 d.

Reference:

TRIGA Mark I L Mechanical Maintenance and Operating Manual.

Answer:

C.008 c

Reference:

UARR TRIGA L Description, Irradiation Facilities.

Answer:

C.009 b.

Reference:

UARR Safety Analysis Report, Page 11 Answer:

C.010 a.

Reference:

UARR Safety Analysis Report, Page 39 Answer:

C.011 b. Question deleted

Reference:

UUAR125 Procedure for Power Calibration of the University of Arizona Research L Triga L Answer:

C.012 b.

Reference:

UARR SAR § Release of Fission Products from a fuel element. p. 57 Answer:

C.013 a.

Reference:

UARR100, p. 10, Answer:

C.014 c.

Reference:

UARR SAR, Table 3.1 Answer:

C.015 c.

Reference:

UARR SAR Table 3.2.

Answer:

C.016 d.

Reference:

UARR Safety Analysis Report, Page 39

Section C: Facility and Radiation Monitoring Systems ANSWERS Answer:

C.017 c.

Reference:

UARR Safety Analysis Report, Page 45 Answer:

C.018 c.

Reference:

UARR Safety Analysis Report, Page 44 Answer:

C.019 c.

Reference:

UARR Safety Analysis Report, Page 62