ML13183A357

Examination Report No. 50-128/OL-13-02, Texas A&M University

ML13183A357
Person / Time
Site:	05000128
Issue date:	07/17/2013
From:	Gregory Bowman Research and Test Reactors Branch B
To:	Reece W Texas A&M Univ
Young P
Shared Package
ML13079A291	List: ML13183A357 ML13183A409
References
50-128/OL-13-02
Download: ML13183A357 (30)

Text

July 17, 2013 Dr. Warren D. Reece, Director Texas A&M University System Nuclear Science Center 1095 Nuclear Science Road MS 3575 College Station, TX 77843-3575

SUBJECT:

EXAMINATION REPORT NO. 50-128/OL-13-02, TEXAS A&M UNIVERSITY

Dear Dr. Reece:

During the week of June 17, 2013, the U.S. Nuclear Regulatory Commission (NRC) administered an operator licensing examination at your TRIGA Reactor. The examination was 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 Phillip T. Young at (301) 415-4094 or via electronic 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. 50-128

Enclosures:

1. Examination Report No. 50-128/OL-13-02

2. Written examination cc without enclosures: See next page

Dr. Warren D. Reece, Director Texas A&M University System Nuclear Science Center 1095 Nuclear Science Road MS 3575 College Station, TX 77843-3575

SUBJECT:

EXAMINATION REPORT NO. 50-128/OL-13-02, TEXAS A&M UNIVERSITY

Dear Dr. Reece:

During the week of June 17, 2013, the U.S. Nuclear Regulatory Commission (NRC) administered an operator licensing examination at your TRIGA Reactor. The examination was 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 Phillip T. Young at (301) 415-4094 or via electronic 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. 50-128

Enclosures:

1. Examination Report No. 50-128/OL-13-02

2. Written examination cc without enclosures: See next page DISTRIBUTION:

PUBLIC RidsNrrDprPrta RidsNrrDprPrtb PYoung, NRR Facility File (CRevelle) O-07 F-08 PIsaac, NRR TLichatz, NRR ADAMS ACCESSION #: ML13183A357 TEMPLATE #: NRR-079 OFFICE NRR/DPR/PROB/CE NRR/DIRS/IOLB/OLA NRR/DPR/PROB/BC NAME PYoung CRevelle GBowman DATE 7/9/13 07/17/2013 07/17/2013 OFFICIAL RECORD COPY

TEXAS A&M UNIVERSITY Docket No. 50-128 cc:

Mayor, City of College Station P.O. Box Drawer 9960 College Station, TX 77840-3575 Governors Budget and Planning Office P.O. Box 13561 Austin, TX 78711 Texas A&M University System ATTN: Dr. Dimitris C. Lagoudas, Interim Deputy Director Nuclear Science Center Texas Engineering Experiment Station 1095 Nuclear Science Road MS 3575 College Station, Texas 77843 Texas A&M University System ATTN: Jim Remlinger, Associate Director Nuclear Science Center Texas Engineering Experiment Station 1095 Nuclear Science Road MS 3575 College Station, Texas 77843 Radiation Program Officer Bureau of Radiation Control Dept. Of State Health Services Division for Regulatory Services 1100 West 49th Street, MC 2828 Austin, TX 78756-3189 Susan M. Jablonski Technical Advisor Office of Permitting, Remediation & Registration Texas Commission on Environmental Quality P.O. Box 13087, MS 122 Austin, TX 78711-3087 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.: 50-128/OL-13-02 FACILITY DOCKET NO.: 50-128 FACILITY LICENSE NO.: R-83 FACILITY: TEXAS A&M UNIVERSITY EXAMINATION DATES: June 17- 19, 2013 SUBMITTED BY: _________/RA/__________ __07/09/13____

Philip T. Young, Chief Examiner Date

SUMMARY

During the week of June 17, 2013, the NRC administered the operator licensing examinations to one (1) Reactor Operator candidate, one (1) Senior Reactor Operator Instant candidate and two (2) Senior Reactor Operator Upgrade candidates.

REPORT DETAILS

1. Examiners: Philip T. Young, Chief Examiner Paulette Torres, Examiner in Training

2. Results:

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

3. Exit Meeting:

Philip T. Young, Chief Examiner Paulette Torres, Examiner in Training Jerry Newhouse, Reactor Supervisor, Texas A&M University TRIGA Greg Stasny, Manager of Reactor Operations, Texas A&M University TRIGA The examiner thanked the facility for their support during the examination and their comments on questions. The examiner indicated that several of the applicants had problems describing the characteristics, production and decay of Argon-41 and Nitrogen-16 ENCLOSURE 1

U. S. NUCLEAR REGULATORY COMMISSION RESEARCH AND TEST REACTOR OPERATOR LICENSING EXAMINATION FACILITY: Texas A&M University REACTOR TYPE: TRIGA DATE ADMINISTERED: 06/17/2013 CANDIDATE:

INSTRUCTIONS TO CANDIDATE:

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

% of Category  % of Candidates Category Value Total Score Value Category 20.00 33.33 A. Reactor Theory, Thermodynamics and Facility Operating Characteristics 20.00 33.33 B. Normal and Emergency Operating Procedures and Radiological Controls 20.00 33.33 C. Plant and Radiation Monitoring Systems FINAL GRADE

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

Candidate's Signature 1

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

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

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

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

10. When you have completed and turned in you examination, leave the examination area.

2

EQUATION SHEET

Q = m c p T = m H = UA T

( - )2 * -4

1 x 10 seconds P max

2 (k)

S CR 1 (1 - K eff 1 ) = CR 2 (1 - K eff 2 )

SCR =

1 - K eff eff = 0.1 sec -1 eff SUR = 26.06 1 - K eff 0 1 CR1

- M= M= =

1 - K eff 1 1 - K eff CR2 P = P0 10 SUR(t) P = P0 e t

(1 - )

P= P0 (1 - K eff ) * -

SDM = = =

+

K eff eff K eff 2 - K eff 1 0.693 ( K eff - 1)

T=

k eff 1 x K eff 2 K eff 6CiE(n) 2 DR1 d 1 = DR 2 d 2 2

DR = DR0 e- t DR = 2 R

2

( 2 - )2 ( 1 - )

=

Peak 2 Peak 1 1 Curie = 3.7 x 1010 dis/sec 1 kg = 2.21 lbm 3

1 Horsepower = 2.54 x 10 BTU/hr 1 Mw = 3.41 x 106 BTU/hr 1 BTU = 778 ft-lbf EF = 9/5 EC + 32 1 gal (H2O) . 8 lbm EC = 5/9 (EF - 32) 3

Section A L Theory, Thermo, and Facility Characteristics Question A.001 (1.0 point) {1.0}

Given a source strength of 100 neutrons per second (N/sec) and a multiplication factor of 0.8, the expected neutron count rate would be:

a. 125 N/sec

b. 250 N/sec

c. 400 N/sec

d. 500 N/sec Answer: A.01 d.

Reference:

C.R. = S/(1 - Keff) C.R. = 100/(1 - 0.8) = 100/0.2 = 500 Question A.002 (1.0 point) {2.0}

A reactor is slightly supercritical, with the thermal utilization factor = 0.700. A control rod is inserted to bring the reactor back to critical. Assuming all other factors remain unchanged, the new value for the thermal utilization factor is:

a. 0.698

b. 0.700

c. 0.702

d. 0.704 Answer: A.02 a.

Reference:

R. R. Burn, Introduction to Nuclear Reactor Operations, page 3-16.

In order to decrease K (return to critical), thermal utilization must decrease.

Question A.003 (1.0 point) {3.0}

A reactor is operating at a constant power level of 250 kW. The fission rate of this reactor is approximately:

a. 0.78x1012 fissions/sec.

b. 1.56x1014 fissions/sec.

c. 0.78x1016 fissions/sec.

d. 3.90x1018 fissions/sec.

Answer: A.03 c.

Reference:

R. R. Burn, Introduction to Nuclear Reactor Operations, page 2-51.

250 kW = 1.562x1018 Mev/sec. (From Equation Sheet)

(1.562x1018 Mev/sec)/(200 Mev/fission) = 0.78x1016 fissions/sec.

4

Section A L Theory, Thermo, and Facility Characteristics Question A.004 (1.0 point) {4.0}

Which ONE statement below describes a positive fuel temperature coefficient?

a. When fuel temperature increases, positive reactivity is added.

b. When fuel temperature decreases, positive reactivity is added.

c. When fuel temperature increases, negative reactivity is added.

d. When fuel temperature increases, reactor power decreases.

Answer: A.04 a.

Reference:

R. R. Burn, Introduction to Nuclear Reactor Operations, page 6-5.

Question A.005 (1.0 point) {5.0}

The Moderating Ratio measures the effectiveness of a moderator by combining the scattering cross section, the absorption cross section, and the average energy loss per collision. The Moderating Ratio is expressed as:

a. (absorption cross section)x(scattering cross section)/(average energy loss per collision).

b. (absorption cross section)x(average energy loss per collision)/(scattering cross section).

c. (scattering cross section)x(absorption cross section)x(average energy loss per collision).

d. (average energy loss per collision)x(scattering cross section)/(absorption cross section).

Answer: A.05 d.

Reference:

R. R. Burn, Introduction to Nuclear Reactor Operations, page 2-62.

Question A.006 (1.0 point) {6.0}

Which ONE of the following is an example of neutron decay?

87

a. 35Br 33As83 87

b. 35Br 35Br86 87

c. 35Br 34Se86 87

d. 35Br 36Kr87 Answer: A.06 b.

Reference:

Burn, R., Introduction to Nuclear Reactor Operations, © 1988, § 2.4.6, P. 2-23.

5

Section A L Theory, Thermo, and Facility Characteristics Question A.007 (1.0 point) {7.0}

During the neutron cycle from one generation to the next, several processes occur that may increase or decrease the available number of neutrons. Which ONE of the following factors describes an INCREASE in the number of neutrons during the cycle?

a. Thermal utilization factor.

b. Fast fission factor.

c. Thermal non-leakage probability.

d. Resonance escape probability.

Answer: A.07 b.

Reference:

R. R. Burn, Introduction to Nuclear Reactor Operations, page 3-16.

Question A.008 (1.0 point) {8.0}

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. 0.993

d. 0.893 Answer: A.08 c.

Reference:

R. R. Burn, Introduction to Nuclear Reactor Operations, pg. 4-1.

Question A.009 (1.0 point) {9.0}

Which ONE of the following is the reason for operating with thermal neutrons rather than fast neutrons?

a. Probability of fission is increased since thermal neutrons are less likely to leak out of the core.

b. As neutron energy increases, neutron absorption in non-fuel materials increases exponentially.

c. The absorption cross-section of U-235 is much higher for thermal neutrons.

d. The fuel temperature coefficient becomes positive as neutron energy increases.

Answer: A.09 c.

Reference:

R. R. Burn, Introduction to Nuclear Reactor Operations, pg. 2-39.

6

Section A L Theory, Thermo, and Facility Characteristics Question A.010 (1.0 point) {10.0}

When the reactor is shut down from full power, what is the main contributor to the constant -80 second period that results?

a. The amount of negative reactivity introduced to the core.

b. The decay constant of the longest lived delayed neutron precursors.

c. The degree of neutron absorption by the fission products in the core.

d. The level of the prompt neutron population.

Answer: A.10 b.

Reference:

Burn, R., Introduction to Nuclear Reactor Operations, 1988, page 4-12.

Question A.011 (1.0 point) {11.0}

While the reactor is operating at power in the automatic mode, a void is produced in the core. The regulating rod will:

a. drive out to add positive reactivity.

b. drive in to add positive reactivity.

c. drive out to add negative reactivity.

d. drive in to add negative reactivity.

Answer: A.11 a.

Reference:

Burn, R., Introduction to Nuclear Reactor Operations, 1988, page 6-13.

Question A.012 (1.0 point) {12.0}

What is the normal NSCR neutron startup source for a startup when the reactor has only been shut down for a few days?

a. Gammas produced from Sb124 result in a neutron from Be.

b. Spontaneous fission from Cf252.

c. Gammas produced from fuel result in a neutron from H2.

d. Betas produced from Ra result in a neutron from Li8.

Answer: A.12 a.

Reference:

Standard NRC Question.

7

Section A L Theory, Thermo, and Facility Characteristics Question A.013 (1.0 point) {13.0}

The term prompt critical refers to:

a. the instantaneous jump in power due to a rod withdrawal.

b. a reactor which is supercritical using only prompt neutrons.

c. a reactor which is critical using both prompt and delayed neutrons.

d. a reactivity insertion which is less then eff.

Answer: A.13 b.

Reference:

Burn, R., Introduction to Nuclear Reactor Operations, 1988, page 4-1.

Question A.014 (1.0 point) {14.0}

K The temperature coefficient of the NSCR core is 1.2 x104 K and the average control rod worth of the o

C regulating control rod is 0 .093 $ inch . If the reactor is in the automatic mode and the temperature increases by 25oC, the regulating rod will move: (assume eff =0.007 K K ).

a. 5.3 inches in

b. 4.6 inches out

c. 0.5 inches in

d. 7.8 inches out Answer: A.14 b.

Reference:

Standard NRC Question K

0 .093 inch x 0 .007 = 6 .51E 4 K inch K

1.2 E 4 o K

• 25o C = 3 E 3 K K C

Since the temperature rise results in a negative reactivity insertion, the control rod will need to drive out to add positive reactivity.

3 E 3 K D = K = 4 .61 inches K

6 .51E 4 K inch 8

Section A L Theory, Thermo, and Facility Characteristics Question A.015 (1.0 point) {15.0}

Which ONE of the following describes the characteristics of a good moderator?

a. Low scattering cross section and high absorption cross section.

b. Low scattering cross section and low absorption cross section.

c. High scattering cross section and low absorption cross section.

d. High scattering cross section and high absorption cross section.

Answer: A.15 c.

Reference:

Burn, R., Introduction to Nuclear Reactor Operations, 1988, page 2-45.

Question A.016 (1.0 point) {16.0}

What is the approximate amount of time that it will take the amount of Xenon in the core to reach negligible levels after the reactor is shut down from full power? The Xenon will be considered to be negligible after 7 half-lives have passed. (Xe-135 T1/2 = 9.2 hrs)

a. 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br />

b. 64 hours7.407407e-4 days <br />0.0178 hours <br />1.058201e-4 weeks <br />2.4352e-5 months <br />

c. 68 hours7.87037e-4 days <br />0.0189 hours <br />1.124339e-4 weeks <br />2.5874e-5 months <br />

d. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Answer: A.16 d.

Reference:

Burn, R., Introduction to Nuclear Reactor Operations, 1988, page 8-11.

Question A.017 (1.0 point) {17.0}

The reactor had been running for 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> straight at 1 megawatt when it was shutdown for maintenance. The maintenance took six hours, and you have just restarted the reactor and raised power to 1 megawatt and placed the reactor in automatic control. Which ONE of the following is the expected response of the regulating rod for the next half hour?

a. Drive in

b. Drive out

c. Not move

d. Drive out then back in Answer: A.17 a.

Reference:

Lamarsh, J.R., Introduction to Nuclear Engineering, 1983. § 7.4, pp. 316 - 322.

9

Section A L Theory, Thermo, and Facility Characteristics Question A.018 (1.0 point) {18.0}

If a $1.50 pulse has a peak power of 250 MW, a FWHM of 100 ms, and a fuel temperature rise of 145 C, what would you estimate the peak power, FWHM, and fuel temperature rise values would be for a $2.00 pulse?

a. Peak power: 780 MW FWHM: 80 ms Temp. rise: 210 C

b. Peak power: 1000 MW FWHM: 50 ms Temp. rise: 290 C

c. Peak power: 1200 MW FWHM: 50 ms Temp. rise: 350 C

d. Peak power: 900 MW FWHM: 80 ms Temp. rise: 210 C Answer: A.18 b.

Reference:

SAR § XI: Peak Power is proportional to $prompt2, FWHM is proportional to 1/$prompt and temperature increase is proportional to $prompt Question A.019 (1.0 point) {19.0}

During a reactor power decreases, the delayed neutron fraction, :

a. remains unchanged.

b. decreases because prompt neutrons are being produced at a slower rate.

c. decreases because delayed neutron precursors are being produced at a slower rate.

d. increases because delayed neutrons are being produced from precursors that were formed at the higher power level.

Answer: A.19 d.

Reference:

R. R. Burn, Introduction to Nuclear Reactor Operations, pg. 4-8.

Question A.20 (1.0 point) {20.0}

A reactor power calibration is being performed by measuring the rate of temperature increase in the reactor pool. Which ONE of the following conditions would result in calculated power being LESS THAN actual power?

a. The measured final temperature is greater than the true temperature.

b. The measured final temperature is less than the true temperature.

c. The calculated volume of water in the pool is greater than the true volume.

d. The calculated rate of temperature increase is greater than the true rate.

Answer: A.20 b.

Reference:

SOP Power Calibration.

10

Section B Normal/Emerg. Procedures & Rad Con Question B.001 [1.0 point] {1.0}

Which one of the following is a requirement for all fuel movements involving the core?

a. At least one fuel element temperature measuring channel must be operable.

b. A Health Physics technician must be on call.

c. All controls rods must be installed in the core.

d. The neutron source must be installed Answer: B.001 a.

Reference:

SOP-II-I Reactor Core Manipulation & TS 3.2.1 Question B.002 [1.0 point] {2.0}

An experiment with a reactivity worth of $0.40 is to be removed from the core. Prior to performing this operation:

a. reactor power must be less than 600 kW.

b. the reactor must be subcritical.

c. the reactor must be subcritical by at least $0.40.

d. the reactor must be shutdown.

Answer: B.02 d.

Reference:

SOP Steady State Operation.

Question B.003 [1.0 point] {3.0}

In accordance with SOP "Personnel Dosimetry," an Expected High Dose Individual is a person who:

a. may receive a dose greater than the annual limit.

b. may receive a dose greater than 10% of the annual limit.

c. will not be expected to exceed 10% of the annual limit.

d. has received an unknown amount of radiation resulting from an accident.

Answer: B.03 b.

Reference:

SOP Personnel Dosimetry.

11

Section B Normal/Emerg. Procedures & Rad Con Question B.004 [1.0 point] {4.0}

The area radiation monitor at the Reactor Bridge is out of service for maintenance. As a result:

a. the reactor cannot be operated.

b. the reactor can continue to operate.

c. the reactor can continue to operate only if the monitor is replaced with a portable gamma instrument with its own alarm.

d. the reactor can continue to operate only if the alarm setpoints of the remaining area radiation monitors are lowered.

Answer: B.04 c.

Reference:

TA&M TS, Section 3.5.1.

Question B.005 [1.0 point] {5.0}

The dose rate 10 feet from a point source is 25 mrem/hour. A person working for 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> at a distance of 3 feet from the source will receive a dose of:

a. 83 mrem.

b. 125 mrem.

c. 278 mrem.

d. 417 mrem.

Answer: B.05 d.

Reference:

DR1d12= DR2d22 ; (25)(100) = DR2(9) ; DR2 = 277 mrem/hour.

Total dose received = (277 mrem/hour)(1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />) = 417 mrem.

Question B.006 [1.0 point] {6.0}

Select the MODE from Column II when the Safety Channels from Column I are required to be operable.

Modes may be used once, more than once, or not at all.

Column I Column II (Safety Channel) (Mode)

a. Fuel Element Temperature 1. Steady State only

b. Preset timer 2. Both modes

c. Transient Rod Position 3. Pulse only

d. Log Power Answer: B.06 a. = 2; b. = 3; c. = 1; d. = 2

Reference:

Tech Spec - 3.2.1 Reactor Measuring Channels Table 1 and 3.2.2 Reactor Safety Systems and Interlocks Table 2a and 2b 12

Section B Normal/Emerg. Procedures & Rad Con Question B.007 [1.0 point] {7.0}

Limiting Safety System Settings used to prevent exceeding a Safety Limit:

a. can be exceeded and prevent exceeding the Safety Limit.

b. can be exceeded during transients.

c. can be changed by the Reactor Safety Board.

d. apply only in the steady state mode of operation.

Answer: B.07 a.

Reference:

TA&M Technical Specifications, Section 2.2.

Question B.008 [1.0 point] {8.0}

A Limited Access Worker must receive ____________and is issued a ________badge.

a. General Employee Training; green

b. Radiation Worker Training and General Employee Training; yellow

c. General Employee Training; orange

d. Radiation Worker Training and General Employee Training; blue Answer: B.08 c.

Reference:

SOP NSC Access Control.

Question B.009 [1.0 point] {9.0}

You observe a loss of reactor pool water which can be controlled by adding makeup water. In accordance with the Emergency Plan, your first course of action is to:

a. assess the severity of the pool water loss by observing the leakage rate and reactor bridge area radiation monitor readings.

b. send a member of Reactor Operations to the west end of the pool and position the emergency cover over the 10-inch cooling exit line.

c. dispatch teams to take appropriate action to determine source of leakage and correct by valve manipulation if possible.

d. shutdown the reactor.

Answer: B.09 d.

Reference:

SOP Implementing Procedure For A Pool Level Alarm. {IX D-4}

13

Section B Normal/Emerg. Procedures & Rad Con Question B.010 [1.0 point] {10.0}

In accordance with 10CFR20, the "Derived Air Concentration (DAC) refers to:

a. the amount of radioactive material taken into the body by inhalation or ingestion in one (1) year which would result in a committed effective dose equivalent of five (5) rems.

b. limits on the release of effluents to an unrestricted environment.

c. the dose equivalent to organs that will be received from an intake of radioactive material by an individual during the 50-year period following the intake.

d. the concentration of a given radionuclide in air which, if breathed for 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br />, would result in a committed effective dose equivalent of five (5) rems.

Answer: B.10 d.

Reference:

20CFR20.

Question B.011 [1.0 point] {11.0}

Match each of the following actions in Column I with the correct term from the Technical Specifications in Column II: Channel Check, Channel Test, or Channel Calibration. (Only one term per action).

Column I Column II

a. Immersing a thermometer in an ice 1. Check bath, then in boiling water and 2. Test noting the output. 3. Calibration

b. Placing a source next to a radiation detector and observing meter movement.

c. Performing a determination of reactor power with a heat balance, then adjusting a power meter to correspond to the heat balance.

d. Observing the overlap between two different neutron detectors as power increases.

Answer: B.11 a. = 2; b. = 2; c. = 3; d. = 1

Reference:

TA&M Technical Specifications, Section 1.0 14

Section B Normal/Emerg. Procedures & Rad Con Question B.012 [1.0 point] {12.0}

A person has received a serious injury which does not involve contamination. In accordance with the Emergency Plan, your first course of action is to:

a. notify the SRO on duty.

b. call for an ambulance, briefly describe the injury and explain the type of accident.

c. go to the injured person and assess the extent of the injury.

d. shutdown the reactor.

Answer: B.12 a.

Reference:

SOP Implementing Procedure For A Personnel Injury. {IX C-1 item a}

Question B.013 [1.0 point] {13.0}

An Emergency Action Level is:

a. a condition which calls for immediate action, beyond the scope of normal operating procedures, to avoid an accident or to mitigate the consequences of one.

b. a class of accidents for which predetermined emergency measures should be taken or considered.

c. a specific instrument reading or observation which may be used as a threshold for initiating appropriate emergency procedures.

d. a procedure that details the implementation actions and methods required to achieve the objectives of the emergency plan.

Answer: B.13 c.

Reference:

Emergency Plan Definition 2.8, pg. 9.

Question B.014 [1.0 point] {14.0}

Match the radiation reading from column A with its corresponding radiation area classification (per 10 CFR 20) listed in column B.

COLUMN A COLUMN B

a. 10 mRem/hr 1. Unrestricted Area

b. 150 mRem/hr 2. Radiation Area

c. 10 Rem/hr 3. High Radiation Area

d. 550 Rem/hr 4. Very High Radiation Area Answer: B.14 a. = 2; b. = 3; c. = 3; d. = 4 REF: 10 CFR 20.1003, Definitions 15

Section B Normal/Emerg. Procedures & Rad Con Question B.015 [1.0 point] {15.0}

The Quality Factor is used to convert

a. dose in rads to dose equivalent in rems.

b. dose in rems to dose equivalent in rads.

c. contamination in rads to contamination equivalent in rems.

d. contamination in rems to contamination equivalent in rads.

Answer: B.15 a.

Reference:

10CFR20.1004.

Question B.016 [1.0 point] {16.0}

Technical Specification 5.5 requires all fuel elements shall be stored in a geometrical array where Keff is less than:

a. 0.80

b. 0.85

c. 0.90

d. 0.95 Answer: B.16 a.

Reference:

Technical Specification 5.6 Question B.017 [1.0 point] {17.0}

SOP II-C Reactor Startup, requires placing the diffuser in operation if anticipated power level is equal to or greater than

a. 0.5 Kilowatts.

b. 5 Kilowatts.

c. 50 Kilowatts.

d. 500 Kilowatts.

Answer: B.17 d.

Reference:

SOP II-C Reactor Startup 16

Section B Normal/Emerg. Procedures & Rad Con Question B.018 [1.0 point] {18.0}

While performing a power calibration the difference between the indicated power and the measured power is 10%. Which ONE of the below statements is correct for this condition?

a. Position the detector to match indicated and measured power.

b. A difference this great is suspect and may be an indication of thermocouple grounding or improper ice bath preparation.

c. Adjustments to the power instrumentation cannot be performed under any circumstances, if the difference is greater than 5%.

d. A difference this great is suspect and may be an indication of a "shadowing effect" Answer: B.18 b.

Reference:

SOP § II-J.2.b, pp. 1 of 3.

Question B.019 [1.0 point] {19.0}

Select the correct sequence of rod withdrawal during a normal reactor startup.

a. Shim-Safeties in gang to upper limit, Transient to mid position, Regulating to criticality

b. Regulating to upper limit, Transient to mid position, Shim-Safeties in gang to criticality

c. Transient to upper limit, Regulating to mid position, Shim-Safeties in gang to criticality

d. Transient to upper limit, Shim-Safeties in gang to mid position, Regulating to criticality.

Answer: B.19 c.

Reference:

SOP II-C pg. 2 of 4 Question B.020 [1.0 point] {20.0}

You wish to store a small radioactive source temporarily in the reactor building. The source strength is estimated to be 500 millicuries and it emits gamma rays of an average energy of 1.3 Mev. Approximately how far from the source would you have to erect a "CAUTION - HIGH RADIATION AREA" barrier?

a. 780 feet

b. 39 feet

c. 15 feet

d. 6 feet Answer: B.20 d.

Reference:

High radiation area - 100 mr per hour.

R/hr = 6E / d2, d2 = (6)(.5)(1.3) / 0.1 = 39, d = 6.25 feet 17

Section C Facility and Radiation Monitoring Systems Question C.001 [1.0 point] {1.0}

Which one of the following describes the yellow light associated with the beam port water shutters?

a. An illuminated yellow light indicates that the shutter tube is evacuated and the beam is active.

b. An illuminated yellow light indicates that a shutter flood permissive has been selected by the reactor operator.

c. The yellow light tells the experimenter that the beam has been cut off.

d. The yellow light warns the experimenter of the commencement of a reactor startup.

Answer: C.01 a.

REF SOP IV-D.3.b.10 Question C.002 [1.0 point] {2.0}

An experimenter is attempting to open the door on beam port #5 while the reactor is in operation. As a result:

a. An alarm horn in the lower research area is activated to warn the experimenter that the reactor is in operation.

b. An annunciator occurs on the console in the control room indicating a beam port door is being opened.

c. The cameras in the lower research area are automatically scanned so the operator would observe the beam port door being opened.

d. Opening of the beam port door during operation will result in a scram.

Answer: C.02 b.

Reference:

SOP IV-D.2, Beam Port Experiments.

18

Section C Facility and Radiation Monitoring Systems Question C.003 [1.0 point] {3.0}

Which ONE of the following is the purpose of the stainless steel liner that encircles the reactor pool?

a. Reduce radiation exposure to people.

b. Contain the water within the pool.

c. Prevent outside contaminants from getting into the pool.

d. Support the biological shield structure.

Answer: C.03 b.

Reference:

SAR, 1.8 Facility Modifications and History page 4 Question C.004 [1.0 point] {4.0}

More than 95% of the facility's Ar-41 is produced in the:

a. beam ports.

b. pneumatic system.

c. reactor pool.

d. reactor building atmosphere.

Answer: C.04 c.

Reference:

SAR; 11.1.1.1- Airborne Radiation Sources page 142 Question C.005 [1.0 point] {5.0}

The Log Power Channel consists of a(n) ____________ and provides an input to the

a. Ion Chamber; period circuit.

b. Compensated Ion Chamber; low count rate (2 cps) interlock.

c. Fission Chamber; servo controller.

d. Fission Chamber; 1 kW pulse interlock.

Answer: C.05 d.

Reference:

SAR; - 7.2.3.1, Log Power Channel page 106.

19

Section C Facility and Radiation Monitoring Systems Question C.006 [1.0 point, 0.25 each] {6.0}

For the items labeled A through D on the attached figure, Transient Rod Drive, select the proper component from the item list in Column II. All items in Column II may not be necessarily used.

Column I (Figure label) Column II (Item list)

A._____ 1. Ball Nut B._____ 2. Vent holes C._____ 3. Piston D._____ 4. Bottom limit

5. Piston rod

6. Shock absorber Answer: C.06 a. = 2; b. = 6; c. = 3; d. = 1.

Reference:

SAR; - 7.3.1.1, Transient Rod Control Figure 7 - 9.

Question C.007 [1.0 point] {7.0}

Which ONE of the following lists the correct locations for the air handling system dampers?

a. Air inlet to all air handlers, exhaust stack, air inlet to central exhaust fan.

b. Air inlet to all air handlers, fresh air bypass to the exhaust fan, exhaust stack.

c. Fresh air bypass to the exhaust fan, air inlet to central exhaust fan, exhaust stack.

d. Air inlet to all air handlers, fresh air bypass to the exhaust fan, air inlet to central exhaust fan.

Answer: C.07 b.

Reference:

SAR; - 9.1.3, Dampers and Filters, page 128 Question C.008 [1.0 point] {8.0}

Which ONE of the following situations will cause the reactor to automatically SCRAM?

a. Low safety detector voltage (<150 V).

b. High Radiation level at top of pool (>100 mrem/hr).

c. Low pool water level (<90% of normal level).

d. Low air pressure applied to the transient rod (<10 psi).

Answer: C.08 a.

Reference:

SAR; - 7.2.3.5, Safety Power Channels, page 108 20

Section C Facility and Radiation Monitoring Systems Question C.009 [1.0 point] {9.0}

What type of detector does the building particulate monitor use to measure radiation?

a. Gamma scintillator.

b. Geiger-Mueller.

c. Ionization chamber.

d. Beta scintillator.

Answer: C.009 d.

Reference:

SAR; 11.1.1.1- Airborne Radiation Sources page 145 Question C.010 [1.0 point] {10.0}

What automatic action is associated with a high radiation alarm signal from the Building Particulate Monitor?

a. The air handler fans continue to operate and all inlet dampers close.

b. The air handler fans cease operation and all inlet dampers remain open.

c. The air handler fans cease operation and all inlet dampers close.

d. The air handler fans continue to operate and all inlet dampers remain open.

Answer: C.10 d.

Reference:

SAR; 11.1.1.1- Airborne Radiation Sources page 145 Question C.011 [1.0 point] {11.0}

Which set of measurements are chosen by the reactor console thermocouple selector?

a. Fuel temperature, irradiation cell temperature, heat exchanger primary outlet temperature.

b. Fuel temperature, pool water temperature, heat exchanger primary outlet temperature.

c. Fuel temperature, irradiation cell temperature, pool water temperature.

d. Pool water temperature, irradiation cell temperature, heat exchanger primary outlet temperature.

Answer: C.11 c.

Reference:

SAR; - 7.2.3.7, Fuel Temperature Channel page 109 21

Section C Facility and Radiation Monitoring Systems Question C.012 [1.0 point] {12.0}

A 1-3/4 inch diameter hole through the grid plate is located at the southwest corner of the four rod fuel assemblies. The purpose of these holes is to

a. accommodate a fuel followed control rod.

b. provide a mounting location for in-core experiments.

c. allow for accurate repositioning of the reactor core which is essential for numerous experiments.

d. provide a coolant flow path through the grid plate Answer: C.12 a.

Reference:

SAR; - 1.8, Facility Modifications and History page 5.

Question C.013 [1.0 point, 0.25 each] {13.0}

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 Level at Demineralizer. 1. Channeling in Demineralizer.

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

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

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

Answer: C.13 a, = 2; b, = 3; c, = 1; d, = 4

Reference:

Standard NRC Question:

22

Section C Facility and Radiation Monitoring Systems Question C.014 [1.0 point] {14.0}

During reactor operation, a leak develops in the primary to secondary heat exchanger. Which ONE of the following conditions correctly describes how the system will react?

a. Pool level will increase due to leakage from the secondary, the automatic level control will maintain level in the secondary.

b. Cooling tower basin level will decrease due to leakage from the secondary, pool level will increase.

c. Cooling tower level will increase due to leakage from the primary, automatic level control will maintain level in the primary.

d. Cooling tower basin level increase due to leakage from the primary, pool level will decrease.

Answer: C.14 d.

Reference:

SAR; - 5.2, Primary Coolant System page 95.

Question C.015 [1.0 point] {15.0}

The purpose of the diffuser above the core during operation is to

a. reduce dose rate at the pool surface due to N16.

b. enhance heat transfer across all fuel elements in the core.

c. better distribute heat throughout the pool.

d. ensure consistent water chemistry in the core.

Answer: C.15 a.

Reference:

SAR; - 5.6, Nitrogen-16 Control Diffuser System page 100.

Question C.016 [1.0 point] {16.0}

Fill in the blank: In the event of failure of the pool cooling system, the heat capacity of the reactor pool is sufficient to cool the reactor for several ____________, with the reactor operating at 1 Megawatt.

a. Minutes

b. Hours

c. Days

d. Weeks Answer: C.16 b.

Reference:

SAR; - 5.2, Primary Coolant System page 96 and 6.2.3, Emergency Core Cooling System, p. 104 23

Section C Facility and Radiation Monitoring Systems Question C.017 [1.0 point] {17.0}

Which one of the following describes the MINIMUM action an operator would have to take, to prevent excessive loss of pool water in the event of a catastrophic rupture of the primary side of the cooling system heat exchanger? NOTE: PW-1 = Coolant extraction (pump suction) line valve. PW-2 and PW-3 = Coolant (pool) return line valves.

a. Manually shut PW-1, PW-2, and PW-3, in the valve pit of the heat exchanger room.

b. Manually shut PW-1 in the valve pit of the heat exchanger room; PW-2 and PW-3 may remain open due to the check valve installed downstream of the heat exchanger.

c. Remotely shut PW-1, PW-2, and PW-3, using the control switches on the auxiliary panel of the reactor console.

d. No action needed; PW-1, PW-2, and PW-3 will shut automatically when pool water level reaches a preset low level.

Answer: C.17 b.

Reference:

SAR Question C.018 [1.0 point] {18.0}

On a decreasing pool level the university communications room will receive an alarm as a result of lowering level. What other automatic action will occur?

a. Core pump trip.

b. Purification pump trip.

c. Recirculation pump trip.

d. Skimmer pump trip.

Answer: C.18 c.

Reference:

SAR; - 6.2.3, Emergency Core Cooling System, p. 104.

24

Section C Facility and Radiation Monitoring Systems Question C.019 [1.0 point] {19.0}

The design basis for the confinement system ensures that:

a. the reactor building is maintained at a pressure lower than the atmosphere.

b. the reactor building is at a higher pressure than the atmosphere.

c. the reactor building is always equal to atmospheric pressure.

d. the reactor building and the adjacent laboratory are always at the same pressure.

Answer: C.19 a.

Reference:

SAR; - 6.2.2, Containment page 103 Question C.020 [1.0 point] {20.0}

Which one of the following does NOT describe how the reactor power instrumentation system operates differently during pulsing operations.

a. Period scram is bypassed.

b. Pulsing detector output is fed to an integrator.

c. Linear Power is placed on full scale.

d. The normal channels are rebiased to detect pulsing levels.

Answer C.20 d.

Reference:

SAR

• • • • • END OF EXAMINATION *****

25