ML12345A143

Examination Report 50-243/OL-13-01, Oregon State University

ML12345A143
Person / Time
Site:	Oregon State University
Issue date:	12/20/2012
From:	Gregory Bowman Division of Policy and Rulemaking
To:	Reese S Oregon State University
Young P
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ML12325A282	List: ML12345A111 ML12345A143
References
50-243/OL-13-001
Download: ML12345A143 (32)

Text

December 20, 2012 Dr. Steven E. Reese, Director Oregon State University Radiation Center, A100 Corvallis, OR 97331-5903

SUBJECT:

INITIAL EXAMINATION REPORT NO. 50-243/OL-13-01, OREGON STATE UNIVERSITY

Dear Dr. Reese:

During the week of November 26, 2012, the NRC administered operator licensing examinations at your Oregon State University TRIGA 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/

Gregory T. Bowman, Chief Research and Test Reactors Oversight Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Docket No. 50-243

Enclosures:

1. Examination Report No. 50-243/OL-13-01

2. Written examination with facility comments cc w/o enclosures: See next page

Dr. Steven E. Reese, Director December 20, 2012 Oregon State University Radiation Center, A100 Corvallis, OR 97331-5903

SUBJECT:

EXAMINATION REPORT NO. 50-243/OL-13-01, OREGON STATE UNIVERSITY

Dear Dr. Reese:

During the week of November 26, 2012, the NRC administered operator licensing examinations at your Oregon State University TRIGA 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/

Gregory T. Bowman, Chief Research and Test Reactors Oversight Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Docket No. 50-243

Enclosures:

1. Examination Report No. 50-243/OL-13-01

2. Written examination with facility comments cc w/o enclosures: See next page DISTRIBUTION w/ encls.:

PUBLIC PRTB r/f RidsNRRDPRPRTA RidsNRRDPRPRTB Facility File (CRevelle) O-7 F-08 ADAMS ACCESSION #: ML12345A143 TEMPLATE #:NRR-074 OFFICE PRTB:CE IOLB:LA PROB:

NAME PYoung CRevelle GBowman DATE 12/18/2012 12/10 /2012 12/20/2012 OFFICIAL RECORD COPY

Oregon State University Docket No. 50-243 cc:

Mayor of the City of Corvallis Corvallis, OR 97331 David Stewart-Smith Oregon Office of Energy 625 Marion Street, N.E.

Salem, OR 97310 Dr. Richard Spinrad, Vice President for Research Oregon State University Administrative Services Bldg., Room A-312 100 Radiation Center Corvallis, OR 97331-5904 Dr. Michael Hartman, Reactor Administrator Oregon State University 100 Radiation Center, A-100 Corvallis, OR 97331-5903 Dr. Todd Palmer, Chairman Reactor Operations Committee Oregon State University 100 Radiation Center, A-100 Corvallis, OR 97331-5904 Test, Research, and Training Reactor Newsletter University of Florida 202 Nuclear Sciences Center Gainesville, FL 32611

U. S. NUCLEAR REGULATORY COMMISSION OPERATOR LICENSING INITIAL EXAMINATION REPORT REPORT NO.: 50-243/OL-13-01 FACILITY DOCKET NO.: 50-243 FACILITY LICENSE NO.: R-120 FACILITY: Oregon State University TRIGA Reactor EXAMINATION DATES: November 27 - 29, 2012 SUBMITTED BY: __________/RA/__________ __12/18/12_______

Phillip T. Young, Chief Examiner Date

SUMMARY

During the week of November 26, 2012 the NRC administered licensing examinations to five Reactor Operator (RO) applicants and one Senior Operator Upgrade (SROU). The applicants passed all portions of the examination.

REPORT DETAILS

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

2. Results:

RO PASS/FAIL SRO PASS/FAIL TOTAL PASS/FAIL Written 5/0 0/0 5/0 Operating Tests 5/0 1/0 6/0 Overall 5/0 1/0 6/0

3. Exit Meeting:

Phillip T. Young Steve Reese, OSU, Director Todd Keller, OSU, Reactor Administrator Gary Wachs, OSU, Reactor Supervisor Robert Schickler, OSU, SRO The examiner thanked the facility for their assistance ensuring the exam administration went smoothly and their feedback on the written examination. The examiner discussed several weak points during the Operating Test including: 1) Procedure chances, 2) Tagout and Lockout, 3)

Ventilation System Operate/Disable switch operation and consequences, 4) the applicants did not understand the source of the bubbles raising from the reactor, and 5) Operation of the secondary cooling system during reactor start-up.

ENCLOSURE 1

Oregon State University Facility Comments with NRC Resolution.

Question: A.014.

Comment: Appears to be an alpha decay scheme. NRC key answer is a gamma decay.

NRC Resolution: Accepted comment and corrected typographical error, correct response is a.

Question: B.006.

Comment: The required thickness to reduce the dose rate by a factor of 10 is 1.55 inches. This requires four one-half inch thick sheets of lead. The answer key options are either three or five. I would suggest tweaking the mu value such that the required thickness is something like 1.49 inches (three sheets).

NRC Resolution: Accepted answer c. or d. and changed question answers for future use.

Question: B.014.

Comment: Answers are correct but the reference given is Tech Spec 1.31 which is a definition (square wave). The question refers to 10 CFR 55 requirements for maintaining a license.

NRC Resolution: Accepted comment and changed question reference.

Question: B.017.

Comment: OSTROP 4 requires the reactor Supervisors signature for restart (section IX.D).

Answer c refers to the supervisors initials.

NRC Resolution: Accepted facility comment, no impact on this examination but question modified for future use.

Question: C.007.

Comment: Correct answer per OSTROP 2 is 30 inches. Key answer c, 50 is incorrect.

NRC Resolution: Accepted facility comment, answer changed to b.

Question: C.008.

Comment: Correct answer per OSTROP 4 is 15 seconds. Key answer c, 25 sec is incorrect.

NRC Resolution: Accepted facility comment, answer changed to b.

Question: C.011.

Comment: Question is worded in a somewhat unclear manner. Reactor Supervisor permission is required to operate the reactor with these items inoperable.

NRC Resolution: Facility comment NOT accepted, the question asks the applicant to identify whether the listed conditions require Reactor Supervisor permission to start-up the reactor.

Question: C.019.

Comment: Answer is correct, but we no longer have FLIP fuel at the facility.

NRC Resolution: Accepted facility comment, no impact on this examination but question noted as deleted for future use.

U. S. NUCLEAR REGULATORY COMMISSION NON-POWER REACTOR INITIAL LICENSE EXAMINATION FACILITY: OREGON STATE UNIVERSITY REACTOR TYPE: TRIGA DATE ADMINISTERED: 11/27/2012 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 parentheses for each question. A 70% overall is required to pass the examination. Examinations will be picked up three (3) hours after the examination starts.

% OF CATEGORY  % OF CANDIDATES CATEGORY CATEGORY VALUE TOTAL SCORE VALUE A. REACTOR THEORY, 20.00 33.3 _______ _______ THERMODYNAMICS, AND FACILITY OPERATING CHARACTISTICS B. NORMAL AND EMERGENCY 20.00 33.3 _______ _______ OPERATING PROCEDURES AND RADIOLOGICAL CONTROLS 20.00 C. FACILITY AND RADIATION 33.3 _______ _______

MONITORING SYSTEMS 60.00 ___________ TOTALS FINAL GRADE ALL THE WORK DONE ON THIS EXAMINATION IS MY OWN. I HAVE NEITHER GIVEN NOR RECEIVED AID.

CANDIDATE'S SIGNATURE

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. Fill in the date on the cover sheet of the examination (if necessary).

7. Print your name in the upper right-hand corner of the first page of each section of your answer sheets.

8. The point value for each question is indicated in parentheses after the question.

9. Partial credit will NOT be given.

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

11. When you are done and have turned in your examination, leave the examination area as defined by the examiner.

EQUATION SHEET

( 2 )2 = (1 )2 eff = 0.1sec 1 Q = m cP T = m H =UAT Peak2 Peak1 t

P = P0 e S S SCR = * =1x104 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

• M= SDM = =

1 K eff 2 K eff

• 0.693 K eff 2 K eff1

+ T1 =

eff + 2 K eff1 K eff 2 K eff 1

= DR = DR0 et 2 DR1 d1 = DR2 d 2 2

K eff 6 Ci E (n )

DR =

R2 DR - Rem/hr, Ci - curies, E - Mev, R - feet 1 Curie = 3.7 x 1010 dis/sec 1 kg = 2.21 lbm 1 inch=2.54 cm 3 6 1 Horsepower = 2.54 x 10 BTU/hr 1 Mw = 3.41 x 10 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 Fac. Operating Characteristics Question: A.001 [1.0 point] (1.0)

In a subcritical reactor, Keff is increased from 0.861 to 0.946. Which ONE of the following is the amount of reactivity that was added to the reactor core?

a. 0.085 K/K

b. 0.104 K/K

c. 0.161 K/K

d. 0.218 K/K Answer: A.01 b.

Reference:

Lamarsh, Introduction to Nuclear Engineering, 2nd Edition, Page 282 1 = (0.861 - 1)/0.861 = -0.161 k/k 2 = (0.946 - 1)/0.946 = -0.057 k/k

= 2 - 1 = -0.057 - (-0.161) = +0.104 k/k Question: A.002 [1.0 point] (2.0)

You enter the control room and note that ALL nuclear instrumentation show a STEADY NEUTRON LEVEL, and no rods are in motion. Which ONE of the following conditions CANNOT be true?

a. The reactor is critical.

b. The reactor is sub-critical.

c. The reactor is super-critical.

d. The neutron source has been removed from the core.

Answer: A.02 c.

Reference:

Standard NRC Question Question: A.003 [1.0 point] (3.0)

The neutron microscopic cross-section for absorption a generally

a. increases as neutron energy increases.

b. decreases as neutron energy increases.

c. increases as the mass of the target nucleus increases.

d. decreases as the mass of the target nucleus increases.

Answer: A.03 b.

Reference:

Standard NRC Question

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

Which ONE of the following is the reason that Xenon Peaks after a shutdown?

a. Iodine decays faster than Xenon decays

b. Promethium decays faster than Xenon decays

c. Xenon decays faster than Iodine decays

d. Xenon decays faster than Promethium Answer: A.004 a.

Reference:

Standard NRC Question1 Question: A.005 [1.0 point] (5.0)

To make a just critical reactor PROMPT CRITICAL, by definition you must add reactivity equal to

a. eff

b. eff

c. eff

d. Keff Answer: A.05 c.

Reference:

Standard NRC Question Question: A.006 [1.0 point] (6.0)

Which ONE of the following is the dominant factor in determining differential rod worth?

a. Rod speed

b. Total Reactor Power

c. Axial and Radial Flux

d. Delayed neutron fraction Answer: A.06 c.

Reference:

Standard NRC Theory Question

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

With the reactor on a CONSTANT period, which ONE of the following transients will take the LONGEST time to complete? A reactor increase from

a. 1 to 5% of full power.

b. 10 to 20% of full power.

c. 20 to 35% of full power.

d. 40 to 60% of full power.

Answer: A.07 a.

Reference:

time is proportional to P/P0 5/1 > 20/10 > 35/20 > 60/40 Question: A.008 [1.0 point] (8.0)

Which ONE of the following is the MAJOR source of energy released from the fission process?

a. Kinetic energy of the fission fragments

b. Kinetic energy of the fission neutrons

c. Decay of the fission fragments

d. Prompt gamma rays Answer: A.08 a.

Reference:

Standard NRC Reactor Theory Question Question: A.009 [1.0 point] (9.0)

Most nuclear text books list the delayed neutron fraction () as being 0.0065. Most research reactors however have an effective delayed neutron fraction (effective) of 0.0070 . Which ONE of the following is the reason for this difference?

a. Delayed neutrons are born at higher energies than prompt neutrons resulting in a greater worth for the neutrons.

b. Delayed neutrons are born at lower energies than prompt neutrons resulting in a greater worth for the neutrons.

c. The fuel includes U238 which via neutron absorption becomes Pu239 which has a larger for fission.

d. The fuel includes U238 which has a relatively large for fast fission.

Answer: A.09 b.

Reference:

Standard NRC Reactor Theory Question

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

A fast neutron will lose the most energy in a collision with which ONE of the following atoms?

a. H1

b. H2

c. C12

d. U238 Answer: A.10 a.

Reference:

Standard NRC Reactor Theory Question Question: A.011 [1.0 point] (11.0)

For U235, the thermal fission cross-section is 582 barns, and the capture cross-section is 99 barns.

When a thermal neutron is absorbed by U235, the probability that fission will occur is:

a. 0.146

b. 0.170

c. 0.830

d. 0.855 Answer: A.11 d.

Reference:

DOE Fundamentals Handbook, Module X, Probability = f/(f+ a) = 582/(528 + 99) = 582/681 = 0.855 Question: A.012 [1.0 point] (12.0)

As a reactor continues to operate over time, for a CONSTANT power level, the average THERMAL neutron flux

a. decreases, due to the increase in fission product poisons.

b. increases, in order to compensate for fuel depletion.

c. decreases, because the fuel is being depleted.

d. remains the same Answer: A.12 b.

Reference:

Standard NRC Question

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

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.13 d.

Reference:

Standard NRC Question Question: A.014 [1.0 point] (14.0)

The following shows part of a decay chain for the radioactive element Radon (Rn). This decay chain is a good example of ___ decay.

3.8 d

a. Alpha

b. Beta

c. Gamma

d. Neutron Answer: A.14 c. a. typo

Reference:

DOE Fundamentals Handbook Nuclear Physics and Reactor Theory Vol. 2 Question: A.015 [1.0 point] (15.0)

You are poolside at the reactor conducting a tour when someone from the group asks what the blue glow around the reactor is within the pool. Which of the following would be the most correct response?

a. It is binding energy released directly through chain reactions of the fission process

b. It is an effect where high energy, charged particles (e.g., electrons) lose and emit their energy while slowing down through the pool

c. It is an effect when high energy, charged particles (e.g., electrons) pass through the pool at a speed which is greater than the speed of light

d. It is the energy release from the interaction between a neutrino and antineutrino which is known as pair annihilation.

Answer: A.15 c.

Reference:

PSBR Training Manual, Chapter 1.8 Bremstrahlung and Cerenkov Effect

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

A nuclear reactor startup is being performed by adding equal amounts of positive reactivity and waiting for neutron population to stabilize. As the reactor approaches criticality, the numerical change in stable neutron population after each reactivity addition __________, and the time required for the neutron population to stabilize after each reactivity addition ___________.

a. increases; remains the same

b. increases; increases

c. remains the same; remains the same

d. remains the same; increases Answer: A.16 b.

Reference:

Question ID #P1766, NRC Generic Fundamentals Exam Question BankPWR2010 Question: A.017 [1.0 point] (17.0)

Which ONE of the following statements is the most correct regarding a characteristic of subcritical multiplication?

a. The number of neutrons gained per generation doubles for each succeeding generation.

b. A constant neutron population is achieved when the total number of neutrons produced in one generation is equal to the number of source neutrons in the next generation.

c. For equal reactivity additions, it takes less time for the equilibrium subcritical neutron population level to be reached as Keff approaches one.

d. Doubling the indicated power will reduce the margin to criticality by approximately one-half.

Answer: A.17 d.

CR1 1 k 2

=

CR2 1 k1 If CR2 is twice CR1, then to be equal, (1-Keff2) must be half of (1-Keff1).

Reference. DOE Handbook, Vol 2, Section 2.0

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

Given a critical nuclear reactor operating below the point of adding heat (POAH), what reactivity effects are associated with reaching the POAH?

a. There are no reactivity effects because the reactor is critical.

b. The increase in fuel temperature will begin to create a positive reactivity effect.

c. The decrease in fuel temperature will begin to create a negative reactivity effect.

d. The increase in fuel temperature will begin to create a negative reactivity effect.

Answer: A.18 d.

Reference:

DOE Fundamentals Handbook Nuclear Physics and Reactor Theory Vol. 2 Question: A.019 [1.0 point] (19.0)

Which one of the following most correctly completes the following as the reason for having an installed neutron source within the core?

A startup without an installed neutron source...

a. could result in a very short period due to the reactor going critical before neutron population built up high enough to be read on nuclear instrumentation.

b. is impossible as there would be no neutrons available to start up the reactor.

c. would be very slow due to the long time to build up neutron population from so low a level.

d. can be compensated for by adjusting the compensating voltage on the source range detector.

Answer: A.19 a.

Reference:

DOE Fundamentals Handbook Nuclear Physics and Reactor Theory Vol. 2

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

Given the associated graph, which of the following answers best describe the neutron behavior within Region II?

a. The neutron cross section is inversely proportional to the neutron velocity (1/V)

b. The neutron cross section decreases steadily with increasing neutron energy (1/E).

c. Neutrons of specific energy levels (e.g., 50 ev, 100 kev) have a greater potential for leakage from the reactor core

d. Neutrons of specific energy levels (e.g., 50 ev, 100 kev) are more likely to be readily absorbed than neutrons at other energy levels.

Answer: A.020 d.

Reference:

DOE Fundamentals Handbook Nuclear Physics and Reactor Theory Vol. 2

Section B: Normal/Emergency Procedures & Radiological Controls Question B.001 [1.0 points, 1/4 each] {1.0}

Match type of radiation (Column A) with the proper penetrating power (Column B).

Column A Column B

a. Gamma 1. Stopped by thin sheet of paper

b. Beta 2. Stopped by thin sheet of metal

c. Alpha 3. Best shielded by light (low-z) material

d. Neutron 4. Best shielded by heavy (high-z) material Answer: B.01 a. = 4; b. = 2; c. = 1; d. = 3

Reference:

Standard NRC Question Question B.002 [1.0 point] {2.0}

Which ONE of the following correctly defines a Safety Limit?

a. Limits on important process variables which are found to be necessary to reasonably protect the integrity of certain physical barriers which guard against the uncontrolled release of radioactivity.

b. The Lowest functional capability of performance levels of equipment required for safe operation of the facility.

c. Settings for automatic protective devices related to those variables having significant safety functions.

d. a measuring or protective channel in the reactor safety system.

Answer: B.02 a.

Reference:

Technical Specifications 2.1 Question B.003 [1.0 point] {3.0}

Which ONE of the following is the correct definition of a CHANNEL CHECK?

a. The combination of sensor, line, amplifier, and output devices which are connected for the purposes of measuring the value of a parameter.

b. An 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. This verification, may include comparison with independent channels measuring the same variable or other measurements of the variable.

d. The introduction of a signal into the channel for verification that it is operable.

Answer: B.03 c.

Reference:

Technical Specifications 1.4

Section B: Normal/Emergency Procedures & Radiological Controls Question B.004 [1.0 point] {4.0}

While working in an area marked "Caution, Radiation Area," you discover your dosimeter is off scale and leave the area. Assuming you had been working in the area for 45 minutes, what is the maximum dose you would have received?

a. 3.8 mr

b. 35.6 mr

c. 75.0 mr

d. 100 mr Answer: B.04 c.

Reference:

10 CFR 20.1003 Maximum dose in a radiation area is 100 mr/hr.

100 mr/hr @ 0.75 hr = 75 mr.

Question B.005 [1.0 point] {5.0}

Which ONE of the following is the definition for Annual Limit on Intake (ALI)?

a. The concentration of a radio-nuclide in air which, if inhaled by an adult worker for a year, results in a total effective dose equivalent of 100 millirem.

b. 10CFR20 derived limit, based on a Committed Effective Dose Equivalent of 5 Rems whole body or 50 Rems to any individual organ, for the amount of radioactive material inhaled or ingested in a year by an adult worker.

c. The effluent concentration of a radio-nuclide in air which, if inhaled continuously over a year, would result in a total effective dose equivalent of 50 millirem for noble gases.

d. Projected dose commitment values to individuals, that warrant protective action following a release of radioactive material.

Answer: B.05 b.

Reference:

10CFR20.1003

Section B: Normal/Emergency Procedures & Radiological Controls Question B.006 [1.0 point] {6.0}

You been assigned to decrease the dose rate from a point source by about a factor of 10. The point source emits a 1.5 MeV gamma. Your shielding consists of 2 inch thick lead sheets. How many sheets (minimum) are required? Given: the mass attenuation coefficient for lead (for 1.5 MeV gammas

= 0.051 cm2/gram and density of lead is 11.4 gram/cm3.

a. 1 sheets 2 sheets

b. 2 sheets 3 sheets

c. 3 sheets 4 sheets

d. 5 sheets 5 sheets Answer B.06 c. or d. per facility comment, also changed answers per facility comment.

Reference:

First calculate F = 0.051 cm2/g H 11.4 g/cm3 = 0.5814 cm-1.

Next calculate thickness. I = I0 e-Fx ln(1/10) = - Fx x = - [ln(1/10)]/F = ln (0.1)/0.5814 = 3.96 . 4.0 cm Finally calculate number of sheets: (4.0 cm)/(2.54cm/in) = 1.57 inches or about 4 sheets.

Question B.007 [1.0 point] {7.0}

Your Annual limit (Occupational Dose Limit for an adult) for Total Effective Dose Equivalent is Y

a. 1.25 rems

b. 5.0 rems

c. 15.0 rems

d. 50 rems Answer: B.07 b.

Reference:

10CFR20.2001.a(1)

Question B.008 [1.0 point] {8.0}

Which ONE of the following is the Technical Specification BASIS for the Limiting Condition of Operation for pool water temperature being maintained below 120°F?

a. To prevent damage to the resin in the purification system.

b. To prevent cavitation in the primary coolant pump.

c. To maintain the integrity of the Aluminum Reactor Tank.

d. to ensure correct operation of the conductivity cells in the purification system.

Answer B.08 c.

Reference:

Technical Specification 3.3

Section B: Normal/Emergency Procedures & Radiological Controls Question B.009 [1.0 point] {9.0}

Classify each of the experiments listed below as either Class A, Class B or Class C, according to OSTROP 18, Procedures for the Approval and Use of Reactor Experiments.

a. Placing an empty containment tube in the Lazy Susan to test new sample containers.

b. Placing a new experiment into a Beam Tube.

c. An experiment requiring the movement of reactor shielding.

d. An experiment containing explosives.

Answer: B.09 a. = A; b. = B; c. = B; d. = C

Reference:

OSTROP 18, Procedures for the Approval and Use of Reactor Experiments II. CLASSIFICATION, APPROVAL AND USE OF REACTOR EXPERIMENTS Question B.010 [1.0 point] {10.0}

Which ONE of the following materials would require suspension of reactor operations until approval from the Reactor Operations Committee, if that material were dropped (even in minute quantities) into the reactor tank?

a. Mercury

b. Copper or copper bearing alloys.

c. Glass

d. Type 18-8 Stainless Steel Answer: B.10 a.

Reference:

OSTROP 7, Operating Procedures for Reactor Water Systems, I.F, WARNING p. 6.

Question B.011 [1.0 point] {11.0}

Which ONE of the listed measuring channels is REQUIRED by Technical Specifications for steady-state, pulsing and square wave modes of operation?

a. Linear Power Level

b. Log Power Level

c. Nvt circuit

d. Fuel Element Temperature Answer: B.11 d.

Reference:

Technical Specification Table in 3.2.2

Section B: Normal/Emergency Procedures & Radiological Controls Question B.012 [1.0 point] {12.0}

Which ONE of the following correctly defines the Emergency Plan term Protective Action Guide(s)?

a. The person or persons appointed by the Emergency Coordinator to ensure that all personnel have evacuated the facility or a specific part of the facility.

b. a condition or conditions which call(s) for immediate action, beyond the scope of normal operating procedures, to avoid an accident or to mitigate the consequences of one.

c. Projected radiological dose or dose commitment values to individuals that warrant protective action following a release of radioactive material.

d. Specific instrument readings, or observations; radiological dose or dose rates; or specific contamination levels of airborne, waterborne, or surface- deposited radioactive materials that may be used as thresholds for establishing emergency classes and initiating appropriate emergency measures.

Answer: B.12 c.

Reference:

Emergency Plan 2.0 Definitions.

Question B.013 [1.0 point] {13.0}

Per the Emergency Plan the primary Emergency Support Center is Room A100. However, for many of the more minor emergencies, the control point may be moved up to the

a. Health Physics Office (D204)

b. Reactor Conference Room (D300)

c. Common Room containing Large Emergency Cabinet (B134)

d. Reactor Control Room (D302)

Answer: B.13 d.

Reference:

Emergency Plan 8.0 Emergency Equipment and Facilities.

Question B.014 [1.0 point, 1/4 each] {14.0}

Match the 10CFR55 requirements for maintaining an active operator license in column A with the corresponding time period from column B.

Column A Column B

a. Renew License 1 year

b. Medical Exam 2 years

c. Pass Requalification Written Examination 4 years

d. Pass Requalification Operating Test 6 years Answer: B.14 a. = 6; b. = 2; c. = 2; d. = 1

Reference:

Technical Specifications 1.31 10CFR55 per facility comment

Section B: Normal/Emergency Procedures & Radiological Controls Question B.015 [1.0 point] {15.0}

The Emergency Response Plan defines Emergency Planning Zone (EPZ) as

a. within the walls of the reactor bay (Room D104).

b. the area within a 100 meter radius of the reactor core centerline.

c. within the walls of the Reactor Building.

d. within the walls of the Radiation Center.

Answer: B.15 c.

Reference:

Emergency Response Plan § 6.0 Emergency Planning Zone.

Question B.016 [1.0 point] {16.0}

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

a. The SL is a maximum operationally limiting value that prevent exceeding the LSSS during normal operations.

b. The SL is a limit on important process variables that assures the integrity of the fuel cladding.

The LSSS initiates protective actions to preclude reaching the SL.

c. The LSSS is a limit on important process variables that assures the integrity of the fuel cladding.

The SL initiates protective action to preclude reaching the LSSS.

d. The SL is a maximum setpoint for instrumentation response. The LSSS is the minimum number of channels required to be operable.

Answer: B.16 b.

Reference:

Technical Specifications §§ 1.22 and 1.23 Safety Limits and Limiting Safety System Settings, p. 4 Question B.017 [1.0 point] {17.0}

What is the minimum level of permission to restart the reactor following an unplanned automatic scram?

a. Any Senior Reactor Operators signature in the log book.

b. The Reactor Supervisors (or his alternates) verbal agreement.

c. The Reactor Supervisors (or his alternates) initials signature in the log book.

d. The Facility Directors verbal agreement.

Answer: B.17 c.

Reference:

OSTROP 4, § VIII AUTOMATIC SCRAMS, step D, p. 15

Section B: Normal/Emergency Procedures & Radiological Controls Question B.018 [1.0 point] {18.0}

The Reactor Operator on duty is responsible for authorizing individuals to use the crane. The crane bridge is not allowed in the southern half of the bay if reactor power is greater than _______.

a. 300 watts

b. 3 Kwatts

c. 30 Kwatts

d. 300 Kwatts Answer: B.18 d.

Reference:

OSTROP 23, § III.B Question B.019 [1.0 point] {19.0}

Which ONE of the following statements concerning emergency exposure limits is correct? For lifesaving situations, a total effective dose of up to ____ is permissible without authorization, due to the implied urgency of the situation.

a. 5 rem

b. 10 rem

c. 25 rem

d. 50 rem Answer: B.19 c.

Reference:

Emergency Plan § 7.4.1 Question B.020 [1.0 point] {20.0}

According to the OSTROP 11, Fuel Element Handling Procedure, even the least radioactive fuel element in the core has an associated dose rate of _______________ at a distance of three feet in air after a week or so of decay.

a. greater than 100 R/hr

b. greater than 500 R/hr

c. greater than 1000 R/hr

d. greater than 10000 R/hr Answer: B.20 a.

Reference:

OSTROP 11, Fuel Element Handling Procedure

Section B: Normal/Emergency Procedures & Radiological Controls Section C: Facility and Radiation Monitoring Systems Question C.001 [1.0 point, each] {1.0}

For irradiation of materials involving quantities of Uranium and Thorium in standard OSTR Irradiation Facilities the irradiation must be less than or equal to:

Facility Exposure

a. Thermal column 1. 1 MWh

b. Pneumatic Transfer 2. 0.083 MWh

c. Rotating Rack 3. 30 MWh Answer: C.01 a. = 3; b. = 2; c. = 1

Reference:

Approved Experiment B.11 Question C.002 [1.0 point] {2.0}

Which of the following cannot cause the stack monitor abnormal flow annunciator:

a. A Detector failure

b. The pump being off

c. A foreign object lodged in the sample line

d. The throttle valve not being positioned correctly Answer: C.02 a.

Reference:

OSTROP 1 Question C.003 [1.0 point] {3.0}

The Continuous Air Monitor (CAM) High Activity Alarm can be expected from certain normal operations such as:

a. After a pulse

b. An increase in bulk water temperature

c. Liquid Holdup Tank level high

d. Long term operation at 1 MW Answer: C.03 a.

Reference:

OSTROP 1

Section B: Normal/Emergency Procedures & Radiological Controls Question C.004 [1.0 point] {4.0}

An increase in primary water radioactivity due to fission products will be detected first by the:

a. Stack Monitor

b. Continuous Air Monitor

c. Reactor Top Area Radiation Monitor

d. Primary System Water High Activity Alarm Answer: C.04 b.

Reference:

OSTROP 1 Question C.005 [1.0 point] {5.0}

Operating with the reactor tank water level below the low water level alarm will cause an increase in production of radiation in the reactor tank. Which of the following is the result of the increased radiation?

a. Nitrogen 16

b. Argon 41

c. Sodium 24

d. Shine from the reactor fuel Answer: C.05 b.

Reference:

OSTROP 1 and OSTROP 2 Question C.006 [1.0 point] {6.0}

When starting up the stack monitor it takes about ________ for the natural particulate background radioactivity to reach equilibrium on the particulate channel of the stack monitor

a. 10 minutes

b. 1/2 hour

c. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />

d. 11/2 hour Answer: C.06 c.

Reference:

OSTROP 2

Section C: Facility and Radiation Monitoring Systems Question C.007 [1.0 point] {7.0}

Which one of the following indications is the Liquid Holdup Tank level that requires notification of the Health Physics group so that they can initiate plans to sample and pump the tank?

a. >15 inches

b. >30 inches

c. 50 inches

d. 60 inches Answer: C.07 c. b. typo

Reference:

OSTROP 2 Question C.008 [1.0 point] {8.0}

For reactor operation in the AUTOMATIC mode, the servo system will limit the period to _______ while attaining the power level called for on the FLUX CONTROL potentiometer.

a. >3seconds

b. 15 seconds

c. 25 seconds

d. 60 seconds Answer: C.08 c. b. typo

Reference:

OSTROP 4 Question C.009 [1.0 point, each] {9.0}

Match the core configuration in Column A with the correct accent marker line placed down the right edge of applicable pages in the Reactor Operations Log Book noted in Column B.

Column A Column B

a. normal configuration 1. yellow

b. sample-holding dummy fuel element assembly in position B-1 2. pink

c. cadmium-lined in-core irradiation tube in position B-1 3. none Answer: C.09 a. = 3; b. = 1; c. = 2.

Reference:

OSTROP 5

Section B: Normal/Emergency Procedures & Radiological Controls Question C.010 [1.0 point] {10.0}

In order to keep occupational radiation doses within Oregon States ALARA program objectives, there are restrictions on personnel occupancy of the reactor top when the reactor power level exceeds 100 kW. Which one of the following correctly describes this restriction?

a. 3 minutes per individual per day

b. 30 minutes per individual per day

c. 30 minutes per individual per week

d. 30 minutes per individual per quarter Answer: C.10 c.

Reference:

OSTROP 6 Question C.011 [1.0 point, 1/4 each] {11.0}

In addition to the operable instrumentation specified in the Technical Specifications, indicate which of the following instrumentation requires the specific permission of the Reactor Supervisor for the reactor to be taken critical? (Requires permission or Does Not require permission)

a. External scram circuits

b. Pool water temperature monitor

c. Liquid waste hold-up tank level indicator

d. Any area radiation monitor not required by the Technical Specifications Answer: C.11 a. = R; b. = R; c. = R; d. = R

Reference:

OSTROP 6 Question C.012 [1.0 point] {12.0}

Which ONE of the following is NOT a design feature of the Purification System?

a. Reduce radiation levels due to dissolved ions.

b. Reduce corrosion rate due to dissolved ions.

c. Reduce radiation levels due to suspended ions.

d. Reduce radiation levels due to soluble gases.

Answer: C.12 d.

Reference:

ORST Training Manual Volume I, p. 106

Section C: Facility and Radiation Monitoring Systems Question C.013 [1.0 point] {13.0}

Which ONE of the following is the method used to minimize mechanical shock to the standard control rods on a scram?

a. A piston, (part of the connecting rod) drives water out of a dashpot as the rod nears the bottom of its travel.

b. A piston (part of the connecting rod) drives air out of a dashpot as the rod nears the bottom of travel.

c. An electrical-mechanical brake energizes when the rod down limit switch is energized.

d. A small spring located at the bottom of the rod.

Answer: C.13 a.

Reference:

OSU Trn Man Volume 1, Control Rod Drives, p. 50.

Question C.014 [1.0 point] {14.0}

WHICH ONE of the following detectors is used primarily to measure Ar41 released to the environment?

a. Stack Gas Monitor

b. Air Particulate Monitor

c. Area Radiation Monitor above pool

d. NONE, Ar41 has too short a half-life to require environmental monitoring.

Answer: C.14 a.

Reference:

Per telcon with G. Wachs.

Question C.015 [1.0 point, 1/4 each ] {15.0}

Match each beam port in column A with its corresponding description in column B.

Column A Column B (Beam Port) (Description)

a. #1 1. Radial, terminating at outer edge of Graphite Reflector assembly

b. #2 2. Radial, terminating at inner surface of Graphite Reflector assembly

c. #3 3. Same as 1, with a cylindrical void in Graphite Reflector graphite.

d. #4 4. Tangential to the outer edge of the core.

Answer: C.15 a. = 3; b. = 1; c. = 4; d. = 2

Reference:

OSU Training Volume 1, Beam Port Facilities

Section B: Normal/Emergency Procedures & Radiological Controls Question C.016 [1.0 point] {16.0}

On a pipe rupture, according to OSTROP 7, you are to stop the pumps and shut some valves. If you are unable to shut the valves, what design feature prevents siphoning of the reactor tank water?

a. vacuum breaker located on the shell side of the heat exchanger.

b. primary system pipes only go down six inches below the normal water surface.

c. holes located in each water pipe about 22 inches below the normal water surface.

d. holes located in each water pipe about 6 inches below the normal water surface.

Answer: C.16 c.

Reference:

OSTROP 7, Operating Procedures for Reactor Water Systems Question C.017 [1.0 point] {17.0}

The neutron absorbing part of the standard control rods are ...

a. hafnium impregnated with aluminum oxide.

b. boron-carbide impregnated with hafnium

c. graphite impregnated with boron carbide.

d. aluminum impregnated with boron carbide.

Answer: C.17 c.

Reference:

OSU Training Manual Volume 1, Standard Control Rods, p. 42.

Question C.018 [1.0 point] {18.0}

Match the purification system conditions listed in column A with their respective causes listed in column B. Each choice is used only once.

Column A Column B

a. High Radiation at Demineralizer. 1. Channeling in Demineralizer.

b. High Radiation downstream of Demineralizer. 2. Fuel element failure.

c. High flow rate through Demineralizer. 3. High temperature in Demineralizer.

d. High pressure upstream of Demineralizer. 4. Clogged Demineralizer.

Answer: C.18 a. = 2; b. = 3; c. = 1; d. = 4

Reference:

Standard NRC Question

Section C: Facility and Radiation Monitoring Systems Question C.19 [1.0 point] {19.0} Question deleted from future examinations Why is Erbium added to TRIGA-FLIP fuel?

a. to improve the overall heat transfer coefficient, which is necessary due to higher temperatures generated when pulsing FLIP fuel.

b. to act as both a burnable poison, (allowing more fuel to be added), and as a resonance absorber, (enhancing prompt negative temperature coefficient).

c. to act as a burnable poison only (allowing more fuel to be added).

d. to act as a resonance absorber only, (enhancing prompt negative temperature coefficient).

Answer: C.19 b.

Reference:

Standard NRC question of FLIP fuel Question C.20 [1.0 point] {20.0}

The gas used to move pneumatic tube "rabbit" samples into and out of the reactor is ...

a. H

b. Air

c. CO2

d. N2 Answer: C.20 b.

Reference:

OSU Trn Man Volume 1, Pneumatic Transfer System