ML100320262

Initial Examination Report No. 50-284/OL-10-01, Idaho State University

ML100320262
Person / Time
Site:	Idaho State University
Issue date:	02/04/2010
From:	Johnny Eads Research and Test Reactors Branch B
To:	Kunze J Idaho State University
Young P T, NRR/PRTB, 415-4094
References
Download: ML100320262 (28)

Text

February 4, 2010 Dr. Jay F. Kunze, Reactor Administrator Idaho State University College of Engineering Campus Box 8060 Pocatello, ID 83209

SUBJECT:

INITIAL EXAMINATION REPORT NO. 50-284/OL-10-01, IDAHO STATE UNIVERSITY AGN-201M REACTOR

Dear Dr. Kunze:

During the week of January 11, 2010, the NRC administered operator licensing examinations at your Idaho State University AGN-201M Reactor. The examinations were conducted according to NUREG-1478, "Operator Licensing Examiner Standards for Research and Test Reactors,"

Revision 2. 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 Title 10 of the Code of Federal Regulations Section 2.390, 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 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. Phillip T. Young at (301) 415-4094 or via internet e-mail phillip.young@nrc.gov.

Sincerely,

/RA/

Johnny H. Eads Jr., Chief Research and Test Reactors Branch B Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Docket No. 50-284

Enclosures:

1. Initial Examination Report No. 50-284/OL-10-01

2. Facility Comments on written examination

3. Written examination with facility comments incorporated cc without enclosures:

Please see next page

February 4, 2010 Dr. Jay F. Kunze, Reactor Administrator Idaho State University College of Engineering Campus Box 8060 Pocatello, ID 83209

SUBJECT:

INITIAL EXAMINATION REPORT NO. 50-284/OL-10-01, IDAHO STATE UNIVERSITY AGN-201M REACTOR

Dear Dr. Kunze:

During the week of January 11, 2010, the NRC administered operator licensing examinations at your Idaho State University AGN-201M Reactor. The examinations were conducted according to NUREG-1478, "Operator Licensing Examiner Standards for Research and Test Reactors,"

Revision 2. 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 Title 10 of the Code of Federal Regulations Section 2.390, 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 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. Phillip T. Young at (301) 415-4094 or via internet e-mail phillip.young@nrc.gov.

Sincerely,

/RA/

Johnny H. Eads Jr., Chief Research and Test Reactors Branch B Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Docket No. 50-284

Enclosures:

1. Initial Examination Report No. 50-284/OL-10-01

2. Facility Comments on written examination

3. Written examination with facility comments incorporated cc without enclosures:

Please see next page DISTRIBUTION w/ encls.:

PUBLIC PRTB r/f RidsNRRDPRPRTA RidsNRRDPRPRTB Facility File (CRevelle) O-7 F-08 ADAMS ACCESSION #: ML100320262 TEMPLATE #:NRR-074 OFFICE PRTB:CE IOLB:LA E PRTB:SC NAME PDoyle:mxc CRevelle JEads DATE 02/01/2009 02/02/2009 02/4/2009 C = COVER E = COVER & ENCLOSURE N = NO COPY

OFFICIAL RECORD COPY Idaho State University Docket No. 50-284 cc:

Idaho State University Dr. Richard T. Jacobsen, College of Engineering Dean Campus Box 8060 Pocatello, ID 83209-8060 Idaho State University Dr. Richard R. Brey Radiation Safety Officer Physics Department Box 8106 Pocatello, ID 83209-8106 Toni Hardesty, Director Idaho Dept. of Environmental Quality 1410 North Hilton Boise, ID 83606 Test, Research and Training Reactor Newsletter 202 Nuclear Sciences Center University of Florida Gainesville, FL 32611

U. S. NUCLEAR REGULATORY COMMISSION OPERATOR LICENSING INITIAL EXAMINATION REPORT REPORT NO.: 50-284/OL-10-01 FACILITY DOCKET NO.: 50-284 FACILITY LICENSE NO.: R-110 FACILITY: Idaho State University AGN-201M Reactor EXAMINATION DATES: January 11 - 14, 2010 SUBMITTED BY: __________________________ 1/27/2010 Phillip T. Young, Chief Examiner Date

SUMMARY

During the week of January 11, 2010, the NRC administered operator licensing examinations to five Reactor Operator and one Senior Reactor Operator (Instant) candidates. Three of the Reactor Operator candidates failed one or more sections of the written examination and one of the Reactor Operator candidates also failed the operating examination.

REPORT DETAILS

1. Examiners: Phillip T. Young, Chief Examiner, NRC

2. Results:

RO PASS/FAIL SRO PASS/FAIL TOTAL PASS/FAIL Written 2/3 1/0 3/3 Operating Tests 4/1 1/0 5/1 Overall 2/3 1/0 3/3

3. Exit Meeting:

Phillip T. Young, NRC Chief Examiner Dr. Jay F. Kunze, Reactor Administrator Adam Mallicoat, Reactor Supervisor Thanked the facility for their support during the examination.

Discussed the following observations; applicants were weak in overall knowledge of Technical Specification organization, weak in understanding of what causes the amber interlock lamp on the console to illuminate, weak in knowledge of how a BF3 detector operates, weak in knowledge of 10CFR20 dose limits and radiation area limits.

ENCLOSURE 1

FACILITY COMMENTS:

Mr. Young, Below I've outlined my issues and reasoning I have with some of the exam questions. Please take this information in to consideration when grading the exams.

Adam Mallicoat COMMENT:

A-12) The question assumes that the source is in the reactor without stating it. Our procedures dictate that the source is never in the reactor except for initial start up. I believe that B or D should be considered acceptable answers.

NRC Resolution:

The point of the question is understanding subcritical neutron multiplication, given the source does not remain in at Idaho State, the question is misleading with no correct answer.

This question is deleted from the examination COMMENT:

B-6) We believe none of the answers are correct, as written. At a minimum there needs to be an RO at the console, a Certified Observer in the reactor room, and a SRO within 30 minutes of the facility. An Authorized Observer isn't necessarily a Certified Observer.

NRC Resolution:

Agree with facility comment, the question is deleted COMMENT:

C-1) This wording is technically correct but in practice we've refereed to the safety chassis as the interlock bus. We agree that C is the only appropriate answer.

NRC Resolution:

In the future we will refer to the safety chassis as the safety chassis/interlock bus.

COMMENT:

C-3) After discussion we feel that B should also be considered a correct answer. SR-1 and SR-2 are considered to be safety rods and CCR and FCR are control rods. FCR cannot be inserted unless SR-1 and SR-2 are fully inserted.

NRC Resolution:

Agree with facility comment, will accept either a or b as correct answers.

ENCLOSURE 2

FACILITY COMMENTS: continued COMMENT:

C-10) It doesn't specify the magnitudes that would cause a reactor scram. So the reactor can have a period, radiation level, and water level without scraming the reactor. In the future it would be more clear to express the options as too short a period, too high a rad. level, or two low a water level. We feel that either A and B should also be accepted or the question should be omitted.

NRC Resolution:

Radiation level does not cause a scram thus is the only correct answer.

U. S. NUCLEAR REGULATORY COMMISSION NON-POWER INITIAL REACTOR LICENSE EXAMINATION FACILITY: Idaho State University AGN-201M Reactor REACTOR TYPE: AGN-201M DATE ADMINISTERED: 1/12/2010 REGION: 4 CANDIDATE: _________________________________

INSTRUCTIONS TO CANDIDATE:

Answers are to be written on the answer sheet provided. Attach the answer sheets to the examination. Points for each question are indicated in brackets for each question. A 70% in each section is required to pass the examination. Examinations will be picked up three (3) hours after the examination starts.

% of Category % of Candidates Category Value Total Score Value Category 20.00 33.3 A. Reactor Theory, Thermodynamics and Facility Operating Characteristics 15.00 33.3 B. Normal and Emergency Operating Procedures and Radiological Controls 10.00 33.3 C. Facility and Radiation Monitoring Systems 45.00 100.0 TOTALS All work done on this examination is my own. I have neither given nor received aid.

Candidate's Signature ENCLOSURE 3

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 neither received nor 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 and each answer sheet.

6. Mark your answers on the answer sheet provided. USE ONLY THE PAPER PROVIDED AND DO NOT WRITE ON THE BACK SIDE OF THE PAGE.

7. The point value for each question is indicated in [brackets] after the question.

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. Ensure all information you wish to have evaluated as part of your answer is on your answer sheet. Scrap paper will be disposed of immediately following the examination.

11. To pass the examination you must achieve a grade of 70 percent or greater in each category.

12. There is a time limit of three (3) hours for completion of the examination.

13. When you have completed and turned in you examination, leave the examination area. If you are observed in this area while the examination is still in progress, your license may be denied or revoked.

EQUATION SHEET

Pmax =

( )2 eff = 0.1sec 1 Q& = m& c P T = m& H =UAT (2 l )

t P = P0 e S S SCR = l* =1x10 4 sec 1 K eff eff + &

SUR = 26 .06

( ) (

CR1 1 K eff1 = CR2 1 K eff 2 ) CR1 ( 1 ) = CR2 ( 2 )

(1 ) M=

1 CR

= 2 P = P0 10SUR(t )

P= P0 1 K eff CR1 1 K eff1 1 K eff l*

M= SDM = =

1 K eff 2 K eff l* 0.693 K eff 2 K eff1

+ T1 =

eff + & 2 K eff1 K eff 2 K eff 1

= DR = DR0 e t 2 DR1 d1 = DR2 d 2 2

K eff 6 Ci E (n ) ( 2 )2 = (1 )2 DR =

R2 Peak2 Peak1 DR - Rem, Ci - curies, E - Mev, R - feet 1 Curie = 3.7 x 1010 dis/sec 1 kg = 2.21 lbm 1 Horsepower = 2.54 x 103 BTU/hr 1 Mw = 3.41 x 106 BTU/hr 1 BTU = 778 ft-lbf °F = 9/5 °C + 32 1 gal (H2O) 8 lbm °C = 5/9 (°F - 32) cP = 1.0 BTU/hr/lbm/°F cp = 1 cal/sec/gm/°C

Section A - Reactor Theory, Thermodynamics and Facility Operating Characteristics Question A.001 [1.0 point] (1.0)

Which of the following is the largest effect on the reactivity worth of a control rod?

a. Overall reactor power.

b. Speed of the control rod.

c. Axial and radial flux shape.

d. Delayed neutron fraction value.

Answer: A.001 c.

Reference:

NUCLEAR REACTOR THEORY; LAMARSH Question A.002 [1.0 point] (2.0)

Which ONE of the following describes the difference between a moderator and reflector?

a. A reflector increases the neutron production factor and a moderator increases the fast fission factor.

b. A reflector decreases the thermal utilization factor and a moderator increases the fast fission factor.

c. A reflector decreases the neutron production factor and a moderator decreases the fast non-leakage factor.

d. A reflector increases the fast non-leakage factor and a moderator increases the thermal utilization factor.

Answer: A.002 d.

Reference:

Glasstone & Sesonke, Nuclear Reactor Engineering, Chapter 1, Section 1.51 & 1.52 Question A.003 [1.0 point] (3.0)

The delayed neutron fraction changes over core life primarily due to the:

a. buildup of Pu240 which increases the delayed neutron fraction.

b. buildup of Pu239 which decreases the delayed neutron fraction.

c. depletion of U235 which decreases the delayed neutron fraction.

d. depletion of U238 which increases the delayed neutron fraction.

Answer: A.003 b.

Reference:

Glasstone & Sesonke, Nuclear Reactor Engineering, Chapter 5, Section 5.170, Chapter 2, Table 2.10.

Section A - Reactor Theory, Thermodynamics and Facility Operating Characteristics Question A.004 [1.0 point] (4.0)

Select the answer that describes the inherent safety feature provided by the temperature coefficient of reactivity.

a. Its negative value causes reactivity to increase as moderator temperature increases.

b. Its negative value causes reactivity to decrease as moderator temperature increases.

c. Its positive value causes reactivity to increase as moderator temperature increases.

d. Its positive value causes reactivity to decrease as moderator temperature increases.

Answer: A.004 b.

Reference:

Basic Reactor Theory Question A.005 [1.0 point] (5.0)

The reactor is initially shut down with count rate at 8 counts per second (cps) and Keff = 0.975.

Control rods are inserted, changing Keff to 0.995. Select the stable count rate would you expect.

a. 15 cps

b. 25 cps

c. 40 cps

d. 80 cps Answer: A.005 c.

Reference:

Basic Reactor Theory Question A.006 [1.0 point] (6.0)

Which one of the following is the correct reason that delayed neutrons allow human control of the reactor?

a. More delayed neutrons are produced than prompt neutrons.

b. Delayed neutrons increase the mean neutron lifetime.

c. Delayed neutrons take longer to thermalize than prompt neutrons.

d. Delayed neutrons are born at higher energies than prompt neutrons.

Answer: A.006 b.

Reference:

Standard NRC Question1

Section A - Reactor Theory, Thermodynamics and Facility Operating Characteristics Question A.007 [1.0 point] (7.0)

What is the kinetic energy range of a thermal neutron?

a. > 1 MeV

b. 100 KeV - 1 MeV

c. 1 eV - 100 KeV

d. < 1 eV Answer: A.007 d.

Reference:

Standard NRC Question Question A.008 [1.0 point] (8.0)

Which ONE of the following is the type of neutron source that is used at the Idaho State University AGN-201?

a. Radium - Beryllium

b. Plutonium - Beryllium

c. Americium - Plutonium

d. Neptunium - Antimony Answer: A.008 a.

Reference:

ISU General Information, "The AGN-201 Reactor", p 5.

Question A.009 [1.0 point] (9.0)

Which ONE of the following elements will produce the greatest energy loss per collison?

a. Oxygen

b. Graphite

c. Hydrogen

d. Uranium 238 Answer: A.009 c.

Reference:

Glasstone & Sesonske, Nuclear Reactor Engineering, Chapter 3, Section 3.66, Table 3.3, p 134.

Section A - Reactor Theory, Thermodynamics and Facility Operating Characteristics Question A.010 [1.0 point] (10.0)

Excess reactivity is the amount of reactivity:

a. associated with samples.

b. needed to achieve prompt criticality.

c. available above that which is required to keep the reactor critical.

d. available above that which is required to make the reactor subcritical.

Answer: A.010 c.

Reference:

Glasstone, Nuclear Reactor Engineering, Chapter 5, Section 5.172 Question A.011 [1.0 point] (11.0)

In the ISU AGN - 201, the largest thermal neutron microscopic cross section is:

a. Xenon-135 capture.

b. Uranium-235 fission.

c. Uranium-238 fission.

d. Graphite C-12 absorption.

Answer: A.011 a.

Reference:

Glasstone & Sesonke, Nuclear Reactor Engineering, Chapter 5, Section 5.62; Question A.012 [1.0 point] (12.0) Question deleted Which ONE of the following causes indicated power (count rate) to stabilize several hours after a reactor scram from full power? Assume normal system/component operation and no maintenance activity.

a. Xenon removal by decay at a constant rate.

b. Subcritical multiplication of source neutrons.

c. Decay of compensating voltage at low power levels.

d. Power level dropping below the minimum detectable level.

Answer: A.012 b.

Reference:

Glasstone, Nuclear Reactor Engineering, Chapter 5, Section 5.45-5.47

Section A - Reactor Theory, Thermodynamics and Facility Operating Characteristics Question A.013 [1.0 point] (13.0)

Which ONE of the following samples when placed individually into the reactor experimental facilities will have a POSITIVE reactivity affect?

a. Gold wire

b. Indium foils

c. Cadmium foils

d. Polyethylene disk Answer: A.013 d.

Reference:

ISU Experiments 3a and 4b Question A.014 [1.0 point] (14.0)

What is the definition of a cross section?

a. The probability that a neutron will be captured by the nucleus.

b. The most likely energy at which a charged particle will be captured.

c. The length a charged particle travels past the nucleus before being captured.

d. The area of the nucleus including the electron cloud.

Answer: A.014 a.

Reference:

Lamarsh, Introduction to Nuclear Engineering, 3rd Edition, page 55 Question A.015 [1.0 point] (15.0)

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.

Answer: A.015 b.

Reference:

Lamarsh, Introduction to Nuclear Engineering, 3rd Edition, page 64.

Section A - Reactor Theory, Thermodynamics and Facility Operating Characteristics Question A.016 [1.0 point] (16.0)

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. immediate increase in the prompt neutron population.

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

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

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

Answer: A.016 a.

Reference:

Lamarsh, Introduction to Nuclear Engineering, 3rd Edition, page 340.

Question A.017 [1.0 point] (17.0)

Which of the following power manipulations would take the longest to complete assuming the same period is maintained?

a. 100 mW to 400 mW

b. 400 mW to 500 mW

c. 1 W to 3.5 W

d. 3.5 W to 4.5 W Answer: A.017 a.

Reference:

Lamarsh, Introduction to Nuclear Engineering, 3rd Edition, page 346.

Section A - Reactor Theory, Thermodynamics and Facility Operating Characteristics Question A.018 [1.0 point] (18.0)

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

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

c. 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.

d. 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.

Answer: A.018 c.

Reference:

Safety Analysis Report, dated November 23, 1995, pg. 104.

Question A.019 [1.0 point] (19.0)

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 value of 60 CPS. The new Keff is:

a. 0.91

b. 0.925

c. 0.95

d. 0.975 Answer: A.019 c.

Reference:

Lamarsh, Introduction To Nuclear Engineering, 3rd Edition.

(CR2/CR1) = (1-Keff0)/(1-Keff1) (60/30) = (0.90)(1-Keff1) Keff1 = 0.95

Section A - Reactor Theory, Thermodynamics and Facility Operating Characteristics Question A.020 [1.0 point] (20.0)

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

a. Alpha radiation

b. Beta and gamma radiation

c. Prompt and delayed neutrons

d. Kinetic energy of the fission fragments Answer: A.020 d.

Reference:

Lamarsh, Introduction to Nuclear Engineering, 3rd Edition, page 88.

END OF SECTION A

Section B. - Normal & Emerg Operating Procedures & Radiological Controls Question B.001 [1 point] (1.0)

Temporary procedures which do NOT change the intent of the original procedure or involve an unreviewed safety question may be approved as a MINIMUM by the:

a. Reactor Administer.

b. Reactor Supervisor.

c. Reactor Safety Committee.

d. Dean of the College of Engineering.

Answer: B.001 b.

Reference:

ISU Technical Specifications, 6.6, page 26.

Question B.002 [1 point] (2.0)

The Technical Specification basis for the MAXIMUM core temperature limit is to prevent:

a. breakdown of the graphite reflector.

b. instrument inaccuracies due to drift.

c. release of fission products.

d. boiling of the shield water.

Answer: B.002 c.

Reference ISU Technical Specifications, 2.1 Basis, page 6.

Question B.003 [1 point] (3.0)

To prevent damage to the reactor or excessive release of radioactive materials in the event of an experiment failure, experiments containing corrosive materials shall:

a. be doubly encapsulated.

b. be limited to less than 5 grams.

c. not be inserted into the reactor or stored at the facility.

d. have a TEDE of less than 500 mrem over two hours from the beginning of the release.

Answer: B.003 a.

Reference:

ISU Technical Specifications, 3.3.a, page 11.

Section B. - Normal & Emerg Operating Procedures & Radiological Controls Question B.004 [1 point] (4.0)

According to Technical Specifications the reactor is considered Shutdown when:

a. the reactor is subcritical.

b. the reactor console key switch is in the "OFF" position.

c. no experiments worth more than o.25¢ are being moved or serviced

d. all safety and control rods are withdrawn and the key is removed from the console with the key switch in "OFF Answer: B.004 d.

Reference:

ISU Technical Specification, 1.22, page 4.

Question B.005 [1 point] (5.0)

The shutdown margin, required by Technical Specifications, with the most reactive safety or control rod fully inserted and the fine control rod fully inserted shall be at least:

a. 0.31 % k/k

b. 0.65 % k/k

c. 1.00 % k/k

d. 1.25 % k/k Answer: B.005 c.

Reference:

ISU Technical Specifications, 3.1.b, page 8.

Question B.006 [1 point] (6.0)

Which ONE of the following would satisfy the MINIMUM Technical Specification staffing requirements whenever the reactor is NOT Shutdown?

a. One authorized operator at the reactor console, a licensed RO in the reactor room.

b. One licensed SRO in the reactor control room and an authorized operator in the reactor room.

c. One authorized operator at the reactor console, a licensed RO in the reactor control room and a licensed SRO on call.

d. One licensed RO at the reactor console, an certified observer in the reactor control room and a licensed SRO on call one hour away.

Answer: B.006 b.

Reference:

ISU Technical Specifications, 6.1.11, page 23;

Section B. - Normal & Emerg Operating Procedures & Radiological Controls Question B.007 [1 point] (7.0)

For abnormal radiation levels within the Operations Boundary, which ONE of the following describes how the reactor operator initiates an emergency building evacuation?

a. Call the Radiation Safety Office.

b. Call 911 and notify the Pocatello Police and Fire Departments.

c. Trip the fire alarm on the east wall near the entrance to the Reactor Lab.

d. Using a Bull Horn making a public address announcement throughout the building.

Answer: B.007 c.

Reference:

ISU Emergency Plan, 7.3 Protective Actions, 7.3.2, page 11.

Question B.008 [1 point] (8.0)

During the preparations for a reactor startup a rod drop test is performed in accordance with O.P. #1. This test is considered satisfactory if ALL of the following criteria are met EXCEPT:

a. The readings of Channels 1, 2, and 3 return to the values they had prior to raising the rods.

b. The rods drop as indicated by the "ENGAGED" lights going out for the rods that were raised.

c. The position indicators for the fine and course control rods are within 0.05 centimeters of 0.00.

d. The drive motors automatically return the magnets to the down position and the "DOWN" and "ENGAGED" lights illuminate for the dropped rods.

Answer: B.008 c.

Reference:

ISU Operating Procedure #1, Rev. 3, Step IV.E, page 6 Question B.009 [1 point] (9.0)

The reactor room high radiation alarm:

a. will automatically scram the reactor on an alarm condition.

b. serves as the evacuation alarm for inadvertent criticality.

c. would require the reactor to be shutdown on an alarm condition.

d. is required to be operable during control rod drive inspection and maintenance.

Answer B.009 c.

Reference ISU TS 3.2 Basis, p 10.

Section B. - Normal & Emerg Operating Procedures & Radiological Controls Question B.010 [1 point] (10.0)

During a reactor startup the low level scram on Channel #1 ensures:

a. protection for a rod drop event.

b. an operating neutron monitor channel.

c. protection for a temperature excursion.

d. the minimum number of period trips are available for startup.

Answer: B.010 b.

Reference:

TS 3.2 Basis, page 10 Question B.011 [1 point] (11.0)

The primary reason for using Nitrogen gas in the rabbit tube assembly is because it is:

a. not explosive.

b. the least expense gas.

c. more resistant to corrosion than other gases.

d. more resistant to neutron activation than other gases.

Answer: B.011 d.

Reference:

Radiation Protection #2.

Question B.012 [1 point] (12.0)

A reactor sample has a disintegration rate of 2 X 1012 disintegrations per second and emits a 0.6 Mev . The expected dose rate from this sample at a distance of 10 feet would be approximately: (Assume a point source)

a. 200 mR/hr

b. 325 mR/hr

c. 2 R/hr

d. 7.5 R/hr Answer: B.012 c.

Reference:

Glasstone & Sesonke, Sect 9.41, p 525.

DR = 6CE/f*2 R/hr, =6(2 X 10*12/3.7X10*10)(0.6)/10*2, =1.9459 R/hr

Section B. - Normal & Emerg Operating Procedures & Radiological Controls Question B.013 [1 point] (13.0)

A channel test of the seismic displacement interlock is required by Technical Specifications to be performed:

a. monthly

b. quarterly

c. semiannually

d. annually Answer: B.013 d.

Reference:

ISU TS 4.2.d, page154.

Question B.014 [1 point] (14.0)

All of the following prerequisites in MP-2, "Procedure to Open the AGN-201 Core Tank", must be met in order to open the Core Tank for maintenance EXCEPT:

a. the Reactor Supervisor must be present.

b. work must stop if radiation levels exceed 50 mRem/hr.

c. the reactor must have be shutdown for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

d. no eating drinking or smoking permitted in the laboratory.

Answer: B.014 b.

Reference:

ISU MP-2 Question B.015 [1 point] (15.0)

The reason for allowing only one control rod at a time to be removed and disassembled during control rod maintenance is to:

a. prevent inadvertent criticality.

b. limit the radiation exposure to personnel.

c. prevent the inadvertent interchange of parts.

d. limit the number of maintenance operations being performed concurrently.

Answer: B.015 c.

Reference:

ISU MP-1, step 4.b, p 2.

END OF SECTION B

Section C. - Facility and Radiation Monitoring Systems Question C.001 [1 point] (1.0)

In the event of a safety chassis grid to cathode short the:

a. fine control rod would scram.

b. magnet current reversal relay would energize.

c. overcurrent relay will disconnect the tube supply voltage.

d. reset relay will energize and remove power to the magnets.

Answer: C.001 c.

Reference ISU SAR Section 4.3.2 Instrumentation System Question C.002 [1 point] (2.0)

Which ONE of the following trips/conditions is associated with the safety chassis interlock bus?

a. period trip.

b. water level.

c. manual scram.

d. low sensitrol level.

Answer: C.002 b.

Reference:

ISU SAR Section 4.3.2 Instrumentation System, Figure 4.3-8 Question C.003 [1 point] (3.0)

Which ONE of the following statements describes the control rod interlocks?

a. The safety rods cannot be inserted unless the course control rod is "ENGAGED".

b. The fine control rod cannot be inserted until the safety rods are "FULLY INSERTED".

c. The fine control rod cannot be inserted unless the course control rod is "DISENGAGED".

d. The safety rods must be fully inserted before their drive motors will operate in the "LOWER" position.

Answer: C.003 a. Due to facility comment accept either a or b as correct answers

Reference:

ISU SAR Section 4.3.2 Instrumentation System, Figure 4.3-8

Section C. - Facility and Radiation Monitoring Systems Question C.004 [1 point] (4.0)

Which ONE of the following statements describes the design/operation of the control rod drive assemblies?

a. The dashpots consist of a spring to reduce rod impact following a scram.

b. The fine control rod does not have a dashpot since it does not scram.

c. The course control rod dashpot uses magnetic force to slow the rod down before impact on a scram.

d. Dashpots are only associated with the safety rods since these rods have been raised against spring tension to assist in driving these rods down on a scram.

Answer: C.004 b.

Reference:

ISU General Information, AGN - 201 Reactor, Control Rods Question C.005 [1 point] (5.0)

Each ONE of the following would be considered an advantage of using fueled control rods over poison rods, EXCEPT:

a. larger reactor size.

b. more symmetrical flux distribution at power.

c. no critical mass assembled when shutdown.

d. simplification of calculations for a homogeneous reactor.

Answer: C.005 a.

Reference:

ISU Exam 5/3/88; AGN - 201 Characteristics.

Question C.006 [1 point] (6.0)

The shield tank is designed to provide shielding from:

a. the thermal column area.

b. high energy radiation.

c. high energy radiation.

d. fast neutron radiation.

Answer: C.006 d.

Reference:

ISU Tech. Specs, 5.1.d., page 18.

Section C. - Facility and Radiation Monitoring Systems Question C.007 [1 point] (7.0)

The shield tank water temperature interlock prevents reactor operation:

a. during periods of high thermal stress.

b. in the event of a high temperature condition.

c. during a condition that will produce high radiation levels.

d. from a reactivity addition due to a temperature decrease.

Answer: C.007 d.

Reference:

ISU Tech. Specs., 3.2 Basis, page 10.

Question C.008 [1 point] (8.0)

The shield tank water level trip will occur if water level drops below:

a. 8 inches

b. 10 inches

c. 12 inches

d. 14 inches Answer: C.008 b.

Reference:

ISU Tech. Specs 3.2.e and ISU Surveillance Procedure #4, p 1.

Question C.009 [1 point] (9.0)

The Idaho State University reactor Access Ports pass through the steel tank:

a. up to the reflector.

b. then the lead shield, up to the reflector.

c. then the lead shield, the reflector and then back out again.

d. then the lead shield, reflector, and the core and then back out again.

Answer: C.009 c.

Reference:

ISU General Information, AGN - 201 Reactor, Access Ports & Glory Hole.

Section C. - Facility and Radiation Monitoring Systems Question C.010 [1 point] (10.0)

Which ONE of the following does NOT automatically cause a reactor scram?

a. Reactor period.

b. Radiation level.

c. Water level.

d. Power failure.

Answer: C.010 b.

Reference:

ISU Safety Analysis Report, dated January 2003, Figure 4.3.3 END OF SECTION C END OF WRITTEN EXAMINATION