Text

Initial Examination Report No. 50-252/OL-05-01, University of New Mexico, April 18, 2005
	ML051400079
Person / Time
Site:	University of New Mexico
Issue date:	05/19/2005
From:	Madden P; NRC/NRR/DRIP/RNRP
To:	Busch R; Univ of New Mexico
	Eresian W, NRC/NRR/DRIP/RNRP, 415-1833
Shared Package
ML042430459	List: ML042430439; ML051300302; ML051400079;
References
	50-252/OL-05-01 50-252/OL-05-01
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May 19, 2005 Dr. Robert Busch, Chief Reactor Supervisor University of New Mexico 209 Farris Engineering Bldg.

Albuquerque, NM 87131-0001

SUBJECT:

INITIAL EXAMINATION REPORT NO. 50-252/OL-05-01, UNIVERSITY OF NEW MEXICO

Dear Dr. Busch:

During the week of April 18, 2005, the NRC administered initial examinations to employees of your facility who had applied for a license to operate your University of New Mexico reactor.

The examination was conducted in accordance with NUREG-1478, "Non-Power Reactor Operator Licensing Examiner Standards," Revision 1. At the conclusion of the examination, the examination questions and preliminary findings were discussed with those members of your staff identified in the enclosed report.

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 NRCs document 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. Warren Eresian at 301-415-1833 or internet e-mail wje@nrc.gov.

Sincerely,

/RA/

Patrick M. Madden, Section Chief Research and Test Reactors Section New, Research and Test Reactors Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Docket No. 50-252

Enclosures:

1. Initial Examination Report No. 50-252/OL-05-01

2. Examination and answer key cc w/encls: Please see next page

University of New Mexico Docket No. 50-252 cc:

City Manager City of Albuquerque City Hall Albuquerque, NM 87101 Dr. Robert Busch, Chief Reactor Supervisor Chemical and Nuclear Engineering Department University of New Mexico 209 Farris Engineering Department Albuquerque, NM 87131-1341 Dr. Norman Roderick, Reactor Administrator University of New Mexico Albuquerque, NM 87131-1341 Mr. James De Zetter, Radiation Safety Officer Radiation Control Program Director, State of New Mexico University of New Mexico Albuquerque, NM 87131-1341 TRTR Newsletter University of Florida Department of Nuclear Engineering Sciences 202 Nuclear Sciences Center Gainesville, FL 32611

May 19, 2005 Dr. Robert Busch, Chief Reactor Supervisor University of New Mexico 209 Farris Engineering Bldg.

Albuquerque, NM 87131-0001

SUBJECT:

INITIAL EXAMINATION REPORT NO. 50-252/OL-05-01, UNIVERSITY OF NEW MEXICO

Dear Dr. Busch:

During the week of April 18, 2005, the NRC administered initial examinations to employees of your facility who had applied for a license to operate your University of New Mexico reactor.

The examination was conducted in accordance with NUREG-1478, "Non-Power Reactor Operator Licensing Examiner Standards," Revision 1. At the conclusion of the examination, the examination questions and preliminary findings were discussed with those members of your staff identified in the enclosed report.

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 NRCs document 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. Warren Eresian at 301-415-1833 or internet e-mail wje@nrc.gov.

Sincerely,

/RA/

Patrick M. Madden, Section Chief Research and Test Reactors Section New, Research and Test Reactors Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Docket No. 50-252

Enclosures:

1. Initial Examination Report No. 50-252/OL-05-01

2. Examination and answer key cc w/encls: Please see next page DISTRIBUTION:

PUBLIC RNRP\R&TR r/f Facility File (EBarnhill)

MMendonca, PM WEresian PMadden EXAMINATION PACKAGE ACCESSION NO.: ML042430459 REPORT ACCESSION NO.: ML051400079 TEMPLATE NO.: NRR-074 OFFICE RNRP:CE IROB:LA RNRP:SC NAME WEresian EBarnhill PMadden DATE 05/ 17 /2005 05/ 18 /2005 05/ 19 /2005 C = COVER E = COVER & ENCLOSURE N = NO COPY OFFICIAL RECORD COPY

U. S. NUCLEAR REGULATORY COMMISSION OPERATOR LICENSING INITIAL EXAMINATION REPORT REPORT NO.: 50-252/OL-05-01 FACILITY DOCKET NO.: 50-252 FACILITY LICENSE NO.: R-102 FACILITY: University of New Mexico EXAMINATION DATES: April 18-20, 2005 EXAMINERS: Warren Eresian, Chief Examiner SUBMITTED BY: /RA/ 05/ 17 /2005 Warren Eresian, Chief Examiner Date

SUMMARY

During the week of April 18, 2005, the NRC administered operator licensing examinations to six Reactor Operator candidates. All candidates passed the operating examination. Three candidates failed the written examination. Of the three failures, two candidates failed Category B only, while the third candidate failed Categories A and B.

ENCLOSURE 1

REPORT DETAILS

1. Examiners: Warren Eresian, Chief Examiner

2. Results:

RO PASS/FAIL SRO PASS/FAIL TOTAL PASS/FAIL Written 3/3 N/A 3/3 Operating Tests 6/0 N/A 6/0 Overall 3/3 N/A 3/3

3. Exit Meeting:

Warren Eresian, NRC Chief Examiner Robert Busch, Chief Reactor Supervisor The NRC thanked the facility staff for their cooperation during the examinations. No generic concerns were noted. The facility reviewed the written examination and as a result Category A, Question 17 was deleted because of no correct answer; Category B, Question 15 was deleted because of no correct answer; Category C, Question 10 contained two correct answers.

U. S. NUCLEAR REGULATORY COMMISSION RESEARCH REACTOR LICENSE EXAMINATION FACILITY: University of New Mexico REACTOR TYPE: AGN-201 DATE ADMINISTERED: 04/18/2005 REGION: 4 CANDIDATE:___________________________

INSTRUCTIONS TO CANDIDATE:

Answers are to be written on the answer sheets provided. Attach all answer sheets to the examination.

Points for each question are indicated in parentheses for each question. A score of 70 percent in each category is required to pass the examination.

Examinations will be picked up 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after the examination starts.

% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 19 40 ______ ______ A. REACTOR THEORY, THERMODYNAMICS, AND FACILITY OPERATING CHARACTERISTICS 14 30 ______ ______ B. NORMAL AND EMERGENCY PROCEDURES AND RADIOLOGICAL CONTROLS 15 30 ______ ______ C. FACILITY AND RADIATION MONITORING SYSTEMS 48 100 ______ ______

FINAL GRADE %

All work done on this examination is my own. I have neither given nor received aid.

Candidate's Signature ENCLOSURE 2

NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:

1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.

2. After the examination has been completed, you must sign the statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completing the examination.

This must be done after you complete the examination.

3. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.

4. Use black ink or dark pencil only to facilitate legible reproductions.

5. Print your name in the blank provided in the upper right-hand corner of the examination cover sheet.

6. Print your name in the upper right-hand corner of the answer sheets.

7. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK. NOTE: partial credit will NOT be given on multiple choice questions.

8. If the intent of a question is unclear, ask questions of the examiner only.

9. When turning in your examination, assemble the completed examination with examination questions, examination aids and answer sheets. In addition, turn in all scrap paper.

10. When you are done and have turned in your examination, leave the examination area as defined by the examiner. If you are found in this area while the examination is still in progress, your license may be denied or revoked.

A. REACTOR THEORY, THERMODYNAMICS & FACILITY OPERATING CHARACTERISTICS Page 3 QUESTION: 001 (1.00)

Element X is a 1/v absorber. The absorption cross section a for element X for 0.0253 eV neutrons is 100 barns. Which ONE of the following is the absorption cross section of element X for 0.0506 eV neutrons?

a. 50.0 barns

b. 70.7 barns

c. 100.0 barns

d. 200.0 barns QUESTION: 002 (1.00)

Inelastic scattering is the process whereby a neutron collides with a nucleus and:

a. recoils with the same kinetic energy it had prior to the collision.

b. recoils with a lower kinetic energy, with the nucleus emitting a gamma ray.

c. is absorbed by the nucleus, with the nucleus emitting a gamma ray.

d. recoils with a higher kinetic energy, with the nucleus emitting a gamma ray.

QUESTION: 003 (1.00)

Which ONE of the following statements explains why delayed neutrons allow control of the reactor?

a. Delayed neutrons are born at higher energies than prompt neutrons and require more collisions to reach thermal energy.

b. Delayed neutrons shorten the average time for core response to a reactivity addition.

c. Delayed neutrons increase the average generation time of the neutron population.

d. Delayed neutrons make up a higher percentage of the core's total neutron population than prompt neutrons.

(***** CATEGORY A CONTINUED ON NEXT PAGE *****)

A. REACTOR THEORY, THERMODYNAMICS & FACILITY OPERATING CHARACTERISTICS Page 4 QUESTION: 004 (1.00)

Which ONE of the following factors is affected MOST by an increase in fission product poisoning?

a. Resonance Escape Probability

b. Fast Fission Factor

c. Thermal Utilization Factor

d. Reproduction Factor QUESTION: 005 (1.00)

A step insertion of positive reactivity to a critical reactor causes a rapid increase in the neutron population known as a prompt jump. Which ONE of the following explains the cause of this occurrence?

a. rapid positive reactivity insertion due to the fuel temperature coefficient (Doppler) feedback

b. shift in the prompt neutron lifetime on up-power maneuvers

c. magnitude of the reactivity insertion exceeding the value of the average effective delayed neutron fraction

d. immediate increase in the prompt neutron population QUESTION: 006 (1.00)

Which ONE of the following describes the difference between reflectors and moderators?

a. Reflectors decrease core leakage while moderators thermalize neutrons.

b. Reflectors shield against neutrons while moderators decrease core leakage.

c. Reflectors decrease thermal leakage while moderators decrease fast leakage.

d. Reflectors thermalize neutrons while moderators decrease core leakage.

(***** CATEGORY A CONTINUED ON NEXT PAGE *****)

A. REACTOR THEORY, THERMODYNAMICS & FACILITY OPERATING CHARACTERISTICS Page 5 QUESTION: 007 (1.00)

A critical reactor is operating at a steady-state power level of 1.00 W. Reactor power is increased to a new steady-state power level of 1.05 W. Neglecting any temperature effects, what reactivity insertion is required to accomplish this?

a. 0.05 delta k/k.

b. 5.0% delta k/k.

c. 1.05% delta k/k.

d. Indeterminate, since any amount of positive reactivity could be used.

QUESTION: 008 (1.00)

Which ONE of the following factors in the six-factor formula can be varied by the reactor operator?

a. Fast fission factor.

b. Reproduction factor.

c. Fast non-leakage factor.

d. Thermal utilization factor.

QUESTION: 009 (1.00)

Which ONE of the following is MOST efficient in thermalizing neutrons important to sustaining the chain reaction?

a. Hydrogen atoms in the water shield tank molecules.

b. Hydrogen atoms in the polyethylene.

c. Aluminum atoms in the fuel cladding.

d. Carbon atoms in the reflector.

(***** CATEGORY A CONTINUED ON NEXT PAGE *****)

A. REACTOR THEORY, THERMODYNAMICS & FACILITY OPERATING CHARACTERISTICS Page 6 QUESTION: 010 (1.00)

Rod reactivity worth testing is in progress. Reactor power has been stabilized at 1 watt. The coarse rod is inserted 2 cm. and the stable power rise from 2 to 4 watts takes 2 minutes. Which ONE of the following is the approximate differential rod worth of the coarse rod?

a. 4.8 X 10-4 delta k/k per cm.

b. 2.5 X 10-4 delta k/k per cm.

c. 5.5 X 10-4 delta k/k per cm.

d. 2.4 x 10-4 delta k/k per cm.

QUESTION: 011 (1.00)

In a critical reactor, 100 fast neutrons are produced from fission and start to slow down. Twenty neutrons are captured in resonance peaks and 10 leak out of the core after they have reached thermal energy. The remaining neutrons are absorbed in fuel and other materials. Each fission produces 2.5 neutrons and 85% of the neutrons absorbed in fuel result in fission. For this reactor, the thermal utilization factor is:

a. 0.47

b. 0.62

c. 0.67

d. 1.61 QUESTION: 012 (1.00)

Which ONE of the following statements describes the difference between Differential (DRW) 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

(***** CATEGORY A CONTINUED ON NEXT PAGE *****)

A. REACTOR THEORY, THERMODYNAMICS & FACILITY OPERATING CHARACTERISTICS Page 7 QUESTION: 013 (1.00)

The reactor is initially subcritical with a Keff of 0.94. Two (2) safety rods worth a total of 2.4% delta k/k are inserted into the core. Which ONE of the following is the new K eff?

a. 0.950

b. 0.954

c. 0.962

d. 0.971 QUESTION: 014 (1.00)

A reactor is operating at criticality. Instantaneously, all of the delayed neutrons are suddenly removed from the reactor. The Keff of the reactor in this state would be approximately:

a. 1.007

b. 1.000

c. 0.000

d. 0.993 QUESTION: 015 (1.00)

Of the approximately 200 Mev of energy released per fission event, the largest amount appears in the form of:

a. Beta and gamma radiation

b. Prompt and delayed neutrons

c. Kinetic energy of the fission fragments

d. Alpha radiation

(***** CATEGORY A CONTINUED ON NEXT PAGE *****)

A. REACTOR THEORY, THERMODYNAMICS & FACILITY OPERATING CHARACTERISTICS Page 8 QUESTION: 016 (1.00)

A reactor startup is being performed with an experiment installed in the glory hole. 1/M plots are being used to predict critical rod position. The following data has been obtained:

ROD POSITION COUNT RATE All rods removed 10 CPS 1 SR inserted 15 CPS 2 SR's inserted 31 CPS FR fully inserted 44 CPS CR @ 5 cm inserted 59 CPS CR @ 10 cm inserted 91 CPS CR @ 15 cm inserted 167 CPS Based on the information provided, Which ONE of the following coarse rod positions corresponds most closely with the expected critical rod position?

a. 18 cm inserted

b. 20 cm inserted

c. 24 cm inserted

d. Criticality cannot occur with all rods inserted due to the experiment that is installed 1/M PLOT Rod Position, cm.

(***** CATEGORY A CONTINUED ON NEXT PAGE *****)

A. REACTOR THEORY, THERMODYNAMICS & FACILITY OPERATING CHARACTERISTICS Page 9 QUESTION: 017 (1.00) DELETED The Reactor Supervisor has asked you to perform a rough calculation of Shutdown Margin to ensure Technical Specification requirements are not violated. The most recent rod worth surveillance data is:

CONTROL ROD DELTA K/K WORTH SR #1 1.2 % delta k/k SR #2 1.3 % delta k/k CR 1.2 % delta k/k FR 0.3 % delta k/k Core excess reactivity 0.2 % delta k/k Based on the provided information, which ONE of the following is the present shutdown margin?

a. 2.3 % delta k/k.

b. 2.5 % delta k/k.

c. 2.9 % delta k/k.

d. 3.8 % delta k/k.

QUESTION: 018 (1.00)

Two different neutron sources are used during two reactor startups. The source used in the first startup emits ten times as many neutrons per second as the source used for the second startup. Assuming all other factors are the same, which ONE of the following states the expected result at criticality?

a. Count rate will be lower for the first startup.

b. Count rate will be higher for the first startup.

c. Rod position will be higher for the first startup (rods will be further into the core.)

d. Rod position will be lower for the first startup (rods will be further out of the core.)

(***** CATEGORY A CONTINUED ON NEXT PAGE *****)

A. REACTOR THEORY, THERMODYNAMICS & FACILITY OPERATING CHARACTERISTICS Page 10 QUESTION: 019 (1.00)

A reactor with a negative fuel temperature reactivity coefficient is critical at full power. A control rod is removed and the power decreases to a lower steady-state value. The reactivity of the reactor at the lower power level is zero because:

a. the positive reactivity due to the fuel temperature decrease balances the negative reactivity due to the control rod removal.

b. the negative reactivity due to the fuel temperature decrease balances the negative reactivity due to the control rod removal.

c. the positive reactivity due to the fuel temperature increase balances the negative reactivity due to the control rod removal.

d. the negative reactivity due to the fuel temperature increase balances the negative reactivity due to the control rod removal.

QUESTION: 020 (1.00)

The AGN-201 is designed to produce a fission rate within the thermal fuse that is approximately twice the average of the core. Which ONE of the following describes how this higher reaction rate is accomplished?

a. The polystyrene media used in the thermal fuse is a better moderator, raising the thermal flux in the fuse area.

b. The non-uniform fuel loading in the upper fuel disc increases the thermal flux in fuse area.

c. The fuel enrichment used in the thermal fuse is twice that of the balance of the core resulting in a higher fission rate in the fuse area.

d. The fuel density used in the thermal fuse is twice that of the balance of the core resulting in a higher fission rate in the fuse area.

(***** END OF CATEGORY A *****)

B. NORMAL/EMERGENCY PROCEDURES AND RADIOLOGICAL CONTROLS Page 11 QUESTION: 001 (1.00)

A radiation survey of an area reveals a general radiation reading of 1 mRem/hr. However, a small section of pipe (point source) reads 10 mRem/hr at one (1) meter. Which ONE of the following is the posting requirement for the area, in accordance with 10 CFR Part 20?

a. CAUTION - RADIATION AREA

b. CAUTION - HIGH RADIATION AREA

c. CAUTION - RADIOACTIVE MATERIAL

d. CAUTION - AIRBORNE RADIOACTIVITY AREA QUESTION: 002 (1.00)

The reactor is operating at steady-state power. Under this circumstance:

a. At least two persons must be present. One NRC-licensed operator must be present at the reactor console.

b. Two NRC-licensed operators must be present. One of the operators must be present at the reactor console.

c. One NRC-licensed operator and a Reactor Supervisor must be present at the reactor console.

d. Only one NRC-licensed operator must be present.

QUESTION: 003 (1.00)

A channel test of Nuclear Safety Channels #2 and #3 shall be performed prior to the first reactor startup of the day or prior to each reactor operation extending more than one day. This is an example of a(n):

a. safety limit.

b. limiting condition for operation.

c. surveillance requirement.

d. limiting safety system setting.

(***** CATEGORY B CONTINUED ON NEXT PAGE *****)

B. NORMAL/EMERGENCY PROCEDURES AND RADIOLOGICAL CONTROLS Page 12 QUESTION: 004 (1.00)

A survey instrument with a window probe is used to measure the beta-gamma dose rate from an irradiated experiment. The dose rate is 100 mR/hr with the window open and 60 mR/hr with the window closed. The gamma dose rate is:

a. 100 mR/hr.

b. 60 mR/hr.

c. 40 mR/hr.

d. 160 mR/hr.

QUESTION: 005 (1.00)

Match the 10CFR Part 55 requirements listed in Column A for an actively licensed operator with the correct time period from Column B. Column B answers may be used once, more than once, or not at all.

Column A Column B

a. License Expiration. 1. 1 year

b. Medical Examination 2. 2 years

c. Requalification Written Examination 3. 3 years

d. Requalification Operating Test 4. 6 years QUESTION: 006 (1.00)

Which ONE of the following is the basis for the maximum core temperature safety limit?

a. Prevent separation of the core.

b. Prevent melting of the polyethylene core material.

c. Prevent operating personnel from being exposed to high temperature.

d. Prevent spontaneous ignition of the graphite reflector.

(***** CATEGORY B CONTINUED ON NEXT PAGE *****)

B. NORMAL/EMERGENCY PROCEDURES AND RADIOLOGICAL CONTROLS Page 13 QUESTION: 007 (1.00)

Which ONE of the following is the MAXIMUM allowable excess reactivity with all control and safety rods fully inserted and including the potential reactivity worth of all experiments?

a. 0.065% delta k/k

b. 0.25% delta k/k

c. 0.65% delta k/k

d. 1.0% delta k/k QUESTION: 008 (1.00)

Which ONE of the following defines a CHANNEL CHECK?

a. Connection of output devices for the purpose of measuring the response to a process variable.

b. Adjustment such that the output responds within standards of accuracy and range to known inputs.

c. Introduction of a signal into a channel to verify it is operable.

d. A qualitative verification of acceptable performance by observation of channel behavior.

QUESTION: 009 (1.00)

In the event of any emergency, if the radiation level at the console exceeds _______ mR/hr, the operator should sound the evacuation alarm.

a. 25.

b. 50.

c. 75.

d. 100.

(***** CATEGORY B CONTINUED ON NEXT PAGE *****)

B. NORMAL/EMERGENCY PROCEDURES AND RADIOLOGICAL CONTROLS Page 14 QUESTION: 010 (1.00)

Which ONE of the following precautions must be taken to reduce the likelihood of damage to reactor components and/or radioactivity releases during an experimental failure?

a. Any experiment containing gaseous or liquid fissionable material can only be inserted into a subcritical reactor.

b. Any experiment containing corrosive materials shall be doubly encapsulated.

c. Any experiment containing explosive materials shall be doubly encapsulated.

d. The mass of any corrosive material in an experiment shall be less than two (2) grams.

QUESTION: 011 (1.00)

Which ONE of the following is the dose rate from a 20 curie cobalt source at 5 feet? (Assume Co-60 emits 2.5 Mev)

a. 0.6 rem/hr

b. 1.2 rem/hr

c. 6 rem/hr

d. 12 rem/hr QUESTION: 012 (1.00)

Which ONE of the following would be classified as a Nonroutine Operation?

a. Control rod calibration.

b. Monthly inspections.

c. Excess reactivity determination with the reactor in its design configuration.

d. Control rod drive maintenance.

(***** CATEGORY B CONTINUED ON NEXT PAGE *****)

B. NORMAL/EMERGENCY PROCEDURES AND RADIOLOGICAL CONTROLS Page 15 QUESTION: 013 (1.00)

The total scram withdrawal time of the coarse control rod and the safety rods must be less than:

a. 300 milliseconds.

b. 500 milliseconds.

c. 800 milliseconds.

d. 1000 milliseconds.

QUESTION: 014 (1.00)

During the reactor startup check-out, a radiation survey is to be performed around the reactor. Any radiation level higher than normal should be reported to the Reactor Supervisor. The normal reading for a shutdown reactor is approximately:

a. < 2 mR/hr.

b. < 4 mR/hr.

c. < 6 mR/hr.

d. < 10 mR/hr.

QUESTION: 015 (1.00) DELETED A Reactor Assistant is a person who may perform all of the following functions except:

a. scram the reactor with the Reactor Scram button.

b. assist the Reactor Operator during a startup by removing cadmium from the glory hol e.

c. perform a radiation survey during startup checkout..

d. perform a reactor inspection during startup check-out.

(***** END OF CATEGORY B *****)

C. FACILITY AND RADIATION MONITORING SYSTEMS Page 16 QUESTION: 001 (1.00)

In order to extend the operating life of the Channel 1 source-range detector:

a. the detector is partially covered by a neutron-absorbing cadmium jacket.

b. the high voltage on the detector is automatically switched off at high power.

c. the detector is moved away from the neutron flux as power increases.

d. a negative voltage is applied to the detector as power increases.

QUESTION: 002 (1.00)

Which ONE condition listed below will NOT cause the red light on the safety interlock indicator to illuminate.

a. Shield water low level.

b. No magnet current.

c. Shield tank temperature below 18EC.

d. Earthquake switch open.

QUESTION: 003 (1.00)

An aluminum baffle plate separates the fuel disks in the upper section of the core from the fuel disks in the lower section of the core. Of the total of ______ fuel disks, ______ are in the upper section and _____ are in the lower section.

a. 7; 4; 3

b. 7; 3; 4

c. 9; 6; 3

d. 9; 5; 4

(***** CATEGORY C CONTINUED ON NEXT PAGE *****)

C. FACILITY AND RADIATION MONITORING SYSTEMS Page 17 QUESTION: 004 (1.00)

Which ONE of the following is NOT a control rod system interlock?

a. Reactor startup cannot commence unless both safety rods and the coarse control rod are fully withdrawn from the core.

b. Only one safety rod can be inserted at a time.

c. The coarse control rod cannot be inserted unless both safety rods are fully inserted.

d. At any operating power below 50x10-6 watts, only the coarse control rod can be inserted.

QUESTION: 005 (1.00)

Which ONE of the following identifies the type of detector used in the Channel 2 Neutron Monitoring system?

a. GM tube.

b. Fission chamber.

c. Ionization chamber.

d. Scintillation detector QUESTION: 006 (1.00)

The U-235 fuel in the AGN is contained in fuel disks and control rods. Of the total fuel in the reactor, approximately how much is contained in the control rods?

a. 7%.

b. 15 %.

c. 22%

d. 30%.

(***** CATEGORY C CONTINUED ON NEXT PAGE *****)

C. FACILITY AND RADIATION MONITORING SYSTEMS Page 18 QUESTION: 007 (1.00)

Which ONE of the following is designed to contain fission product gases that might leak from the core?

a. Lead shielding.

b. Water shield.

c. Steel Reactor Tank.

d. Aluminum Core Tank.

QUESTION: 008 (1.00)

Which ONE of the following describes the design purpose of the space between the reactor core and the graphite reflector?

a. Ensures free fall of the bottom half of the core during the most severe transient.

b. Increases the fast neutron population in the vicinity of experiments placed in the access ports.

c. Allows for accumulation and venting of fission product gases created during reactor operation.

d. Prevents core damage during the design basis earthquake and 6 cm. displacements.

QUESTION: 009 (1.00)

All of the remote area radiation monitors (general lab, reactor top, reactor console, checkpoint three) are:

a. G-M detectors.

b. ionization chambers.

c. scintillation detectors.

d. proportional counters.

(***** CATEGORY C CONTINUED ON NEXT PAGE *****)

C. FACILITY AND RADIATION MONITORING SYSTEMS Page 19 QUESTION: 010 (1.00)

The safety rods and coarse rod have a reactivity worth that is approximately ______ times the worth of the fine rod.

a. 2

b. 4

c. 6

d. 10 QUESTION: 011 (1.00)

When the reactor is in the Standard Loading #2 as Amended configuration, the Pu-Be source is located in:

a. access port 1.

b. access port 2.

c. access port 3.

d. access port 4.

QUESTION: 012 (1.00)

In general, the upper fuel disks are different than the lower fuel disks in that the upper disks are ________ and have ______ U-235 contained in them.

a. thicker; more

b. thicker; less

c. thinner; more

d. thinner; less

(***** CATEGORY C CONTINUED ON NEXT PAGE *****)

C. FACILITY AND RADIATION MONITORING SYSTEMS Page 20 QUESTION: 013 (1.00)

Which ONE of the following scrams or trips is designed to prevent excessive radiation levels?

a. Reactor period.

b. High reactor power.

c. Low water level.

d. Low water temperature.

QUESTION: 014 (1.00)

The Low Level Interlock is controlled by power level indication from:

a. Channel 1.

b. Channel 2.

c. Channel 3.

d. Auxiliary Channel.

QUESTION: 015 (1.00)

The startup neutron source should be removed when:

a. the reactor becomes supercritical.

b. when Channel 2 reaches 1x10-12 amps.

c. when Channel 2 reaches 5x10-9 amps.

d. when Channel 2 reaches 1.3x10-7 amps.

(***** END OF CATEGORY C *****)

Lamarsh, Introduction To Nuclear Engineering, 3rd Edition, page 362.

From Equation Sheet, (P/Po) = 2 = e(120/) ; period = 173 seconds.

From Equation Sheet, period = = ( - )/ ; reactivity = = 4.8x10-4 delta k/k.

Differential rod worth = /2 cm. = 2.4x10-4 delta k/k/cm.

ANSWER: 011 (1.00)

C.

REFERENCE:

Lamarsh, Introduction to Nuclear Engineering, 3rd Edition, page 286.

A total of 70 thermal neutrons (100-20-10) are absorbed in the fuel plus other materials. Since the reactor is critical, there were 40 fissions (40x2.5 = 100). Since 85% of the absorptions result in fission, there were 40/0.85 = 47 neutrons absorbed in fuel. The thermal utilization is 47/70 = 0.67.

ANSWER: 012 (1.00)

A.

REFERENCE:

Lamarsh, Introduction to Nuclear Engineering, 3rd Edition, page 361, 362.

ANSWER: 013 (1.00)

C.

Reactor Operation and Training Manual, page 28.

ANSWER: 015 (1.00)

A.

REFERENCE:

Reactor Operation and Training Manual, page 31.

A. REACTOR THEORY, THERMODYNAMICS & FACILITY OPERATING CHARACTERISTICS ANSWER SHEET MULTIPLE CHOICE (Circle or X your choice)

If you change your answer, write your selection in the blank.

001 a b c d _____

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005 a b c d _____

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017 a b c d _____ DELETED 018 a b c d _____

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(***** END OF CATEGORY A *****)

B. NORMAL/EMERGENCY PROCEDURES AND RADIOLOGICAL CONTROLS ANSWER SHEET MULTIPLE CHOICE (Circle or X your choice)

If you change your answer, write your selection in the blank.

001 a b c d _____

002 a b c d _____

003 a b c d _____

004 a b c d _____

005 a_____b_____c_____d _____

006 a b c d _____

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015 a b c d _____ DELETED

(***** END OF CATEGORY B *****)

C. FACILITY AND RADIATION MONITORING SYSTEMS ANSWER SHEET MULTIPLE CHOICE (Circle or X your choice)

If you change your answer, write your selection in the blank.

001 a b c d _____

002 a b c d _____

003 a b c d _____

004 a b c d _____

005 a b c d _____

006 a b c d _____

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(***** END OF CATEGORY C *****)

EQUATION SHEET Q = mcT P = P0 10SUR(t)

P = P0 e(t/) = (R*/) + [(-)/]

= 0.08 seconds-1 (DR1)D12 = (DR2)D22 (DR) = (DRo)e-t (DR) = 6CiE/D2

= (K-1)/K (CR1) (1-K1) = (CR2) (1-K2) 1 Curie = 3.7x1010 dps 1 gallon water = 8.34 pounds 1 Btu = 778 ft-lbf EF = 9/5EC + 32 1 Mw = 3.41x106 BTU/hr EC = 5/9 (EF - 32)

N = S/(1-K)

ML051400079

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