ML24052A209

Examination Report Letter No. 50-020/OL-24-01, Massachusetts Institute of Technology

ML24052A209
Person / Time
Site:	05000200
Issue date:	06/06/2024
From:	Travis Tate NRC/NRR/DANU/UNPO
To:	Foster J Massachusetts Institute of Technology (MIT)
References
50-020/24-01 50-020/OL-24
Download: ML24052A209 (44)

Text

John P. Foster Director of Reactor Operations Nuclear Reactor Laboratory Massachusetts Institute of Technology 138 Albany Street, MS NW12-116A Cambridge, MA 02139

SUBJECT:

EXAMINATION REPORT NO. 50-020/OL-24-01, MASSACHUSETTS INSTITUTE OF TECHNOLOGY

Dear John Foster:

During the week of April 15, 2024, the U.S. Nuclear Regulatory Commission (NRC) administered operator licensing examinations at your Massachusetts Institute of Technology Research 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 component of NRC's Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the NRC website 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 Amy Beasten at (301) 415-8341 or via email at Amy.Beasten@nrc.gov.

Sincerely, Travis L. Tate, Chief Non-Power Production and Utilization Facility Oversight Branch Division of Advanced Reactors and Non-Power Production and Utilization Facilities Office of Nuclear Reactor Regulation Docket No.50-020

Enclosures:

1. Examination Report No. 50-020/OL-24-01

2. Facility Comments with NRC Resolutions

3. Written examination cc: w/enclosures to GovDelivery SubscribersJune 6, 2024 Signed by Tate, Travis on 06/06/24

ML24052A241 NRR-079 OFFICE NRR/DANU/UNPO/CE NRR/DANU/UNPO/OLA NRR/DANU/UNPO/BC NAME ABeasten NJones TTate DATE 6/6/2024 6/6/2024 6/6/2024 U. S. NUCLEAR REGULATORY COMMISSION OPERATOR LICENSING INITIAL EXAMINATION REPORT

REPORT NO.: 50-020/OL 24-01

FACILITY DOCKET NO.: 50-020

FACILITY LICENSE NO.: R-37

FACILITY: Massachusetts Institute of Technology

EXAMINATION DATES: Week of April 15, 2024

SUBMITTED BY: Amy Beasten 5/8/2024 Amy E. Beasten, PhD, Chief Examiner Date

SUMMARY

During the week of April 15, 2024, the NRC administered operator licensing examinations to four Senior Reactor Operating-Instant (SRO-I) candidates and one Reactor Operator (RO) candidate. A request for withdrawal was submitted by the facility for one candidate on April 26, 2024, in accordance with the requirements in 10 CFR 55.31. Because the candidate withdrew after the exam began and before exam results were issued, the candidate is considered a failure in accordance with NUREG-1478. One SRO-I candidate failed the written examination and passed the operating examination, and one RO candidate failed the written examination and the operating examination. The remaining two SRO-I candidates passed all applicable portions of the examination.

REPORT DETAILS

1. Examiner: Amy E. Beasten, PhD, Chief Examiner, NRC

2. Results:

RO PASS/FAIL SRO PASS/FAIL TOTAL PASS/FAIL

Written 0/1 2/2 2/3

Operating Tests 0/1 4/0 4/1

Overall 0/1 2/2 2/3

Enclosure 2

3. Exit Meeting:

Amy E. Beasten, PhD, Chief Examiner, NRC Margaret N. Goodwin, RTR Examiner, NRC Edward Lau, Assistant Director for Reactor Operations Susan Tucker, QA Supervisor Gordon Khose, Managing Director of Operations Frank Warmsley, Interim Superintendent

Prior to administration of the written exam, based on facility comments, adjustments were accepted. Comments provided corrections and additional clarity to questions/answers and identified where changes were appropriate based on current facility conditions. Upon completion of all operator licensing examinations, the NRC examiners met with facility staff representatives to discuss the results and observations. At the conclusion of the meeting, the NRC examiners thanked the facility for their support in the administration of the examination.

FACILITY COMMENTS AND NRC RESOLUTION

QUESTION B.01 [1.0 point]

In accordance with Emergency Procedure 4.4.4.11, NW12 Evacuation, all of the following actions should be taken in the event of a fire EXCEPT:

a. Shut down the reactor via ARI.

b. Attempt to fight the fire.

c. Stop ventilation.

d. Isolate containment building.

Answer: b.

Reference:

Emergency Procedure 4.4.4.11, NW12 Evacuation

Facility Comment B.01 The NRC answer key entry for this question is B: "Attempt to fight the fire" is NOT one of the actions to take in the event of a fire in NWl 2. However, we would like NRC to also consider A as an acceptable answer. This is because the general protocol stated by the reactor's Emergency Plan is to scram the reactor if there is a fire. Specifically, Emergency Plan section 4.7 Emergency Response, sub-section 4.7.3.5 Fire/Damage Control, stipulates the console operator will initiate a reactor scram, not a shutdown of the reactor via ARI, if a smoke detector system alarm is received in the control room.

NRC Resolution B.01 The NRC determined the justification for two correct answers is not supported and the NRC does NOT accept these changes.

QUESTION B.09 [1.0 point]

According to MITR-II Technical Specifications, which ONE of the following conditions is NOT permissible?

a. Operation with neutron sources < 500 W.

b. Control rod drop time < 1 s.

c. Shutdown margin < 1% K/K.

d. Maximum reactivity addition rate < 5 x 10-4 K/K.

Answer: c.

Reference:

MITR-II Technical Specifications 3.1.3, 3.1.4, 3.2.1, 3.2.2

Facility Comment B.09 The NRC answer key entry for this question is C: "Shutdown margin < 1 % delta-K/K" as a condition NOT permissible by the MITR-II Technical Specification (TS). However, we would like NRC to consider D as an acceptable answer as well. This is because TS 3.2.2 specifies "the maximum controlled reactivity addition rate is no more than 5 x 104 delta-K/K/s". Answer D of the exam question provided "5x104 delta-K/K" without "per second" in its unit, and was, as a result, considered NOT permissible by the TS by some candidates.

NRC Resolution B.09 The NRC determined the justification for two correct answers is not supported and the NRC does NOT accept these changes.

QUESTION B.18 [1.0 point]

All of the following are true statements regarding MITR-II Technical Specification surveillance requirements EXCEPT:

a. Surveillance tests for required equipment may be waived when an instrument, component or system is not required to be operable, but such instruments, components, or systems must be tested prior to being used as a required operable instrument, component, or system.

b. All surveillance requirements have a 25% grace period of the specified interval.

c. Surveillance requirements that are deferred during reactor shutdown must be performed as soon as possible after the reactor restarts.

d. The total maximum combined interval time for any three consecutive surveillance intervals may not exceed 3.25 times the specified interval.

Answer: a.

Reference:

MITR-II Technical Specifications, 1.3.11 (1.3.11 states this is true for EXPERIMENTS, not equipment)

Facility Comment B.18 The NRC answer key entry for this question is A: "Surveillance tests for required equipment may be waived when an instrument, component or system is not required to be operable, but... " as a false statement for the MITR-II Technical Specification (TS) surveillance requirements. The actual statement of this is in the last paragraph of TS Definition 1.3.11 Frequency. The actual statement applies to "experiments", such that it is "Surveillance tests for experiments (Section 6) may be waived when an instrument, component or system is not required to be operable, but...

" The "required equipment" versus "experiment" in this question statement is misleading.

Therefore, we recommend NRC to consider eliminating this question because it has no fully correct answer.

NRC Resolution B.18 The NRC determined the justification for eliminating this question is not supported and the NRC does NOT accept these changes.

QUESTION C.13 [1.0 point]

Which ONE of the following actions would be considered a Channel CHECK?

a. Comparison of Nuclear Channel 2 indications with the N-16 monitor indications of reactor power.

b. Comparison of current Nuclear Channel 2 indications with Nuclear Channel 2 indications at the same power levels during previous reactor operations.

c. Adjustment of Nuclear Channel 2 indications while at power to reflect the N-16 monitor indications of reactor power more closely.

d. Adjustment of Nuclear Channel 2 indications following performance of the thermal power calibration to correct for acceptable instrument drift.

Answer: a.

Reference:

MITR-II Technical Specifications 1.3.3

Facility Comment C.13 The NRC answer key entry for this question is A: "Comparison of Nuclear Channel 2 indications with the N-16 monitor indications of reactor power" would be considered a Channel CHECK.

However, we would like NRC to also consider B as an acceptable answer. This is because TS Definition 1.3.3 Channel Check allows "qualitative verification of acceptable performance by observation of channel behavior... 11 Therefore, we consider "comparison of current Nuclear Channel 2 indications with Nuclear Channel 2 indications at the same power levels during previous reactor operations" an acceptable, equally valid Channel Check.

NRC Resolution C.13 The NRC determined the justification for two correct answers is supported and the NRC accepts these changes.

QUESTION C.15 [1.0 point]

During normal, steady state reactor operation at 5.9 MW, the Control on Manual annunciator alarms. Which ONE of the following conditions could NOT have caused this alarm?

a. The regulating rod was inserted to within 1.5 of near-in.

b. An experiment with 0.5% K/K reactivity was inadvertently removed from the pneumatic transfer system.

c. Power set point and actual reactor power deviation was 1.5%.

d. The regulating rod reached its full out position.

Answer: d.

Reference:

MITR-II Reactor Systems Manual, 4.4

Facility Comment C.15 The NRC answer key entry for this question is D: "The regulating rod reached its full out position" could NOT cause the "Control on Manual" alarm. However, we interpret answer B as equally acceptable as well, unless it specifically states that the experiment with 0.5 delta-KIK.

reactivity was inadvertently removed from the *reactor*, not just from the "pneumatic transfer system". This is because we routinely park experiment samples at the transfer station, ready for transfer, either to the reactor or to the adjacent building NW13. So removal of an experiment from the pneumatic transfer system doesn't necessary mean removing an experiment from the reactor. Removing an experiment from the reactor will result in a reactivity transient scenario, but removing an experiment from the transfer system doesn't necessary cause a reactivity transient.

NRC Resolution C.15 The NRC determined the justification for two correct answers is not supported and the NRC does NOT accept these changes.

Massachusetts Institute of Technology Research Reactor

Operator Licensing Examination

Week of April 15, 2024

Enclosure 3 U. S. NUCLEAR REGULATORY COMMISSION NON-POWER REACTOR LICENSE EXAMINATION

FACILITY: MIT Research Reactor

REACTOR TYPE: Tank

DATE ADMINISTERED: April 16, 2024

CANDIDATE: _______________________

INSTRUCTIONS TO CANDIDATE:

Answers are to be written on the Answer sheet provided. Attach all Answer sheets to the examination. Point values are indicated in parentheses for each question. A 70% in each category is required to pass the examination. Examinations will be picked up three (3) hours after the examination starts.

% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY

20.00 33.3 A. REACTOR THEORY, THERMODYNAMICS AND FACILITY OPERATING CHARACTERISTICS

20.00 33.3 B. NORMAL AND EMERGENCY OPERATING PROCEDURES AND RADIOLOGICAL CONTROLS

20.00 33.3 C. FACILITY AND RADIATION MONITORING SYSTEMS

60.00 % TOTALS FINAL GRADE

All 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 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.

Category A: Reactor Theory, Thermodynamics, & Facility Operating Characteristics

A N S W E R S H E E T

Multiple Choice (Circle or X your choice)

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

A01 a b c d ___

A02 a b c d ___

A03 a b c d ___

A04 a b c d ___

A05 a b c d ___

A06 a b c d ___

A07 a b c d ___

A08 a b c d ___

A09 a _____ b _____ c _____ d _____ (0.25 each)

A10 a _____ b _____ c _____ d _____ (0.25 each)

A11 a b c d ___

A12 a b c d ___

A13 a _____ b _____ c _____ d _____ (0.25 each)

A14 a b c d ___

A15 a b c d ___

A16 a b c d ___

A17 a b c d ___

A18 a b c d ___

A19 a b c d ___

A20 a b c d ___

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

Category B: Normal/Emergency Operating Procedures and Radiological Controls

A N S W E R S H E E T

Multiple Choice (Circle or X your choice)

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

B01 a b c d ___

B02 a b c d ___

B03 a b c d ___

B04 a b c d ___

B05 a b c d ___

B06 a b c d ___

B07 a _____ b _____ c _____ d _____ (0.25 each)

B08 a b c d ___

B09 a b c d ___

B10 a _____ b _____ c _____ d _____ (0.25 each)

B11 a b c d ___

B12 a _____ b _____ c _____ d _____ (0.50 each)

B13 a b c d ___

B14 a b c d ___

B15 a b c d ___

B16 a b c d ___

B17 a b c d ___

B18 a b c d ___

B19 a b c d ___

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

Category C: Facility and Radiation Monitoring Systems

A N S W E R S H E E T

Multiple Choice (Circle or X your choice)

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

C01 a b c d ___

C02 a b c d ___

C03 a b c d ___

C04 a b c d ___

C05 a b c d ___

C06 a b c d ___

C07 a b c d ___

C08 a b c d ___

C09 a b c d ___

C10 a b c d ___

C11 a b c d ___

C12 a b c d ___

C13 a b c d ___

C14 a _____ b _____ c _____ d _____ (0.25 each)

C15 a b c d ___

C16 a b c d ___

C17 a b c d ___

C18 a b c d ___

C19 a b c d ___

C20 a b c d ___

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

(********** END OF EXAMINATION **********)

EQUATION SHEET

= = = 2 1 P eff0.1sec max 2

t S S

• 4 P P0e SCR 110 sec 1 Keff

+ CR CR CR 1 K CR 1 K

= 26.06 1 1 2 2 1 eff1 2 eff2

Gn, 1 M 1 CR2 P P 10SUR(t)

P P0 1 K CR 0 eff 1

M 1Keff1 1 Keff

• 1 K SDM eff2 Keff

= + 0.693 K K eff2 eff1 T1 2 K K eff1eff2

Keff1 DR DR et 2 2 K 0 DR1 d1 DR2d2 eff

6Ci E n 2 2 DR 2 1 R2 Peak2 Peak1

1 Curie = 3.7 x 1010 dis/sec 1 kg = 2.21 lb 1 Horsepower = 2.54 x 103 BTU/hr 1 Mw = 3.41 x 106 BTU/hr 1 BTU = 778 ft-lb °F = 9/5 °C + 32 1 gal (H2O) 8 lb °C = 5/9 (°F - 32) cP = 1.0 BTU/hr/lb/°F cp = 1 cal/sec/gm/°C Category A: Reactor Theory, Thermodynamics, and Facility Operating Characteristics

QUESTION A.01 [1.0 point]

What is the condition of the reactor when = 1 ?

1 eff

a. Subcritical

b. Critical

c. Supercritical

d. Prompt critical

QUESTION A.02 [1.0 point]

Which ONE of the following statements is NOT a result of increases in moderator temperature during reactor operation?

a. Thermal utilization factor increases.

b. Resonance escape probability decreases.

c. Rod worth decreases.

d. Thermal non-leakage probability decreases.

QUESTION A.03 [1.0 point]

The figure below shows a trace of reactor period as a function of time. Which ONE of the following describes reactor power from point A to point C?

a. Reactor power is increasing, then decreasing, then stable.

b. Reactor power is constant.

c. Reactor power is increasing, then stable.

d. Reactor power is increasing continually.

Category A: Reactor Theory, Thermodynamics, and Facility Operating Characteristics

QUESTION A.04 [1.0 point]

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

a. Moderators thermalize neutrons and reflectors scatter neutrons to decrease leakage from the core.

b. Moderators scatter fast neutrons and reflectors scatter thermal neutrons.

c. Moderators absorb thermal neutrons and reflectors scatter fast neutrons.

d. Moderators scatter neutrons to decrease leakage from the core and reflectors thermalize neutrons.

QUESTION A.05 [1.0 point]

Which ONE of the following best describes the difference between prompt and delayed neutrons?

a. Prompt neutrons ensure there is a sufficient neutron population to overcome the effects of fission product poisoning following a shutdown and delayed neutrons are responsible for lengthening the neutron generation time to ensure the reactor does not go prompt critical.

b. Prompt neutrons are responsible for the ability to control the rate at which power can rise in the reactor and delayed neutrons are responsible for the rate at which a reactor can be shut down.

c. Prompt neutrons are produced immediately and directly from the fission event and delayed neutrons are produced immediately following the first beta decay of fission fragments.

d. Prompt neutrons are produced from spontaneous fission of U-235 in the fuel, and delayed neutrons are the result of fission in U-238.

QUESTION A.06 [1.0 point]

Which ONE of the following best describes the effect of Xe-135 on normal reactor operation?

a. Xe-135 has a large thermal neutron absorption cross-section, which causes large removal of thermal neutrons from the core, causing negative reactivity addition.

b. When Xe-135 is formed, kinetic energy in the form of heat is released, causing increases in moderator temperature and subsequently decreases in reactor power over the course of normal reactor operation.

c. Xe-135 is inert and therefore has no impact on reactor operations.

d. Xe-135 has a large mass which causes increased scattering collisions, slowing more neutrons to thermal energies as the concentration builds up over time.

Category A: Reactor Theory, Thermodynamics, and Facility Operating Characteristics

QUESTION A.07 [1.0 point]

Given a reactor period of 17 seconds, how long will it take for reactor power to triple?

a. 28.6 seconds

b. 18.7 seconds

c. 15.5 seconds

d. 8.2 seconds

QUESTION A.08 [1.0 point]

Which ONE of the following best describes the effective multiplication factor, keff?

a. The ratio of the number of total neutrons produced by fission in one generation to the number of neutrons lost through absorption in the preceding generation.

b. The ratio of the number of thermal neutrons produced by fission in one generation to the number of thermal neutrons lost through leakage and absorption in the preceding generation.

c. The ratio of the number of fast neutrons produced by fission in one generation to the number of fast neutrons lost through leakage and absorption in the preceding generation.

d. The ratio of the number of total neutrons produced by fission in one generation to the number of neutrons lost through leakage and absorption in the preceding generation.

QUESTION A.09 [1.0 point, 0.25 each]

Match the decay mode in Column A with the definition in Column B. Answers in Column B may be used once, more than once, or not at all.

Column A Column B

a. Alpha decay 1. Conversion of a neutron to a proton and an electron, which is ejected from the

b. Beta-minus decay nucleus.

c. Beta-plus decay 2. Nucleus absorbs an electron from the innermost orbital, which combines with a

d. Electron capture proton to form a neutron.

3. Emission of a helium atom from an unstable nucleus.

4. Conversion of a proton to a neutron and a positron, which is ejected from the nucleus.

Category A: Reactor Theory, Thermodynamics, and Facility Operating Characteristics

QUESTION A.10 [1.0 point, 0.25 each]

Complete the decay chain below by matching the blanks in Column A with the isotopes listed in Column B. Answers in Column B may be used once, more than once, or not at all.

0 + 0 + 135 0 + 0 + (stable) 191 6.571 54 69.11 2.3E61

Column A Column B

a. A

1. 135I 53

b. B

2. 135Cs

c. C 55

3. 135Ba

d. D 56

4. 135Te 52

QUESTION A.11 [1.0 point]

Which ONE of the following best describes the importance of a negative temperature coefficient of reactivity?

a. As fuel temperature increases, the concentration of fission product poisons in the fuel matrix increase, adding negative reactivity through increased neutron absorption.

b. As fuel temperature increases, U-235 in the fuel is consumed, causing Pu-239 to form which becomes additional sources of fission causing control rods to be withdrawn farther, adding negative reactivity.

c. As fuel temperature increases, moderator temperature increases rapidly through conduction heat transfer, adding negative reactivity.

d. An increase in reactor power causes an increase in fuel temperature which results in a negative reactivity addition, causing the power increase to slow or stop.

QUESTION A.12 [1.0 point]

Which ONE of the following statements regarding fission with thermal neutrons is true?

a. U-238 fissions with thermal neutrons because the binding energy released by the absorption of a neutron is greater than the critical energy for fission.

b. U-235 fissions with thermal neutrons because the binding energy released by the absorption of a neutron is less than the critical energy for fission.

c. U-238 fissions with thermal neutrons because the binding energy released by the absorption of a neutron is less than the critical energy for fission.

d. U-235 fissions with thermal neutrons because the binding energy released by the absorption of a neutron is greater than the critical energy for fission.

Category A: Reactor Theory, Thermodynamics, and Facility Operating Characteristics

QUESTION A.13 [1.0 point, 0.25 each]

Match the isotope in Column A with the type of material in Column B. Options in Column B may be used once, more than once, or not at all.

Column A Column B

a. Th-232 1. Fissionable

b. U-233 2. Fertile

c. U-235 3. Fissile

d. U-238

QUESTION A.14 [1.0 point]

A subcritical reactor has a keff of 0.959. How much reactivity is added to change the keff to 0.872?

a. + 0.104

b. - 0.104

c. + 0.084

d. - 0.084

QUESTION A.15 [1.0 point]

Which ONE of the following parameters is the MOST significant in determining the differential rod worth of a control rod?

a. Fuel temperature

b. Reactor power

c. Flux shape

d. Rod speed Category A: Reactor Theory, Thermodynamics, and Facility Operating Characteristics

QUESTION A.16 [1.0 point]

As new fuel is being loaded into the core, the reactor operator is using a 1/M plot to monitor core loading. Which ONE of the following conditions could result in the reactor reaching criticality mass at a value greater than the predicted critical mass?

a. The detector is located so that core load starts away from the detector and subsequent loading proceeds towards the detector.

b. Too much time elapses between subsequent core loadings.

c. The detector and source are too close to each other.

d. The detector is located so that core load starts at a point close to the detector and subsequent loadings move farther from the detector.

QUESTION A.17 [1.0 point]

Which ONE of the following best describes inelastic scattering?

a. A neutron is absorbed by the target nucleus to form a compound nucleus, which ejects an alpha particle or proton.

b. A neutron is absorbed by the target nucleus to form a compound nucleus, which releases its excitation energy by emitting a gamma ray.

c. A neutron interacts with a target nucleus and momentum and kinetic energy are conserved.

d. A neutron is absorbed by the target nucleus to form a compound nucleus, which emits a neutron with lower kinetic energy, and the resulting excited state nucleus emits a gamma.

QUESTION A.18 [1.0 point]

Which ONE of the following defines eff?

a. The fraction of all fission neutrons born as delayed neutrons.

b. The fraction of neutrons at thermal energies born as delayed neutrons.

c. The average of the total delayed neutron fractions of the different types of fuel.

d. The fraction of neutrons at fast energies born as delayed neutrons.

Category A: Reactor Theory, Thermodynamics, and Facility Operating Characteristics

QUESTION A.19 [1.0 point]

During each fission event with a thermal neutron in U-235, ________ is released immediately, and _______ is delayed.

a. 167 MeV; 10 MeV

b. 187 MeV; 23 MeV

c. 210 MeV; 23 MeV

d. 200 MeV; 10 MeV

QUESTION A.20 [1.0 point]

Which ONE of the following statements best describes integral rod worth?

a. The total reactivity worth of the rod at a particular degree of withdrawal.

b. The reactivity change per unit movement of a rod.

c. The area under the curve represents the cumulative effect of withdrawing a control rod a specific amount from the core.

d. The area under the curve represents the cumulative effect of withdrawing a control rod a specific distance from the core.

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

Category B: Normal/Emergency Operating Procedures and Radiological Controls

QUESTION B.01 [1.0 point]

In accordance with Emergency Procedure 4.4.4.11, NW12 Evacuation, all of the following actions should be taken in the event of a fire EXCEPT:

e. Shut down the reactor via ARI.

f. Attempt to fight the fire.

g. Stop ventilation.

h. Isolate containment building.

QUESTION B.02 [1.0 point]

An experiment reading 135 mrem/hr was removed from the reactor. Three hours later, it reads 65 mrem/h. What is the half-life of the experiment?

a. 0.48 hr

b. 1.42 hr

c. 2.84 hr

d. 3.21 hr

QUESTION B.03 [1.0 point]

MITR-II procedure PM 2.4.3, One Pump Operation, requires all of the following changes to the standard procedure for reactor startup and continuous power operation be made EXCEPT:

a. The reactor must not be operated above 2.5 MW power.

b. Reactor Operating Mode selector switch is selected to < 2.5 MW.

c. Primary flow recorder and core inlet pressure switches must be set to scram above 900 gpm.

d. Safety System level scrams must be set at 2.8 MW.

Category B: Normal/Emergency Operating Procedures and Radiological Controls

QUESTION B.04 [1.0 point]

Which ONE of the following defines the term High Radiation Area?

a. Area where radiation exposure rates would result in a dose equivalent in excess of 0.1 rem (1 mSv) in one hour at 30 centimeters from the radiation source.

b. Any area to which access is limited for any reason.

c. Area where radiation exposure rates would result in a dose equivalent in excess of 5 mrem (0.05 mSv) in one hour at 30 centimeters from the radiation source.

d. Any area to which access is limited for the purpose of protecting individuals against undue risks from exposure to radiation and radioactive materials.

QUESTION B.05 [1.0 point]

In accordance with the MITR-II Emergency Plan, which ONE of the following events would be classified as an ALERT?

a. An unauthorized individual gains access to the containment building.

b. Loss of primary coolant such that pool level decreases below the anti-siphon valves.

c. Tornado touches down within the site area boundary.

d. A fire within the containment bay that cannot be extinguished within 15 minutes.

QUESTION B.06 [1.0 point]

An irradiated sample has a gamma dose rate of 15.0 rem/hr as indicated at a distance of 3 feet from the sample. How far from the irradiated sample will the dose rate read 400 mrem/hr?

a. 0.58 ft.

b. 10.61 ft.

c. 18.37 ft.

d. 33.75 ft.

Category B: Normal/Emergency Operating Procedures and Radiological Controls

QUESTION B.07 [1.0 point, 0.25 each]

Match the term in Column A with the Emergency Plan definition in Column B. Options in Column B may be used once, more than once, or not at all.

Column A Column B

a. Emergency Planning Zone 1. The area including the containment building, the fenced in area between the

b. Site Area Boundary containment building and the Boston &

Albany Railroad, the utilities service

c. Restricted Area building, the cooling tower equipment housing, and the one-story portion of

d. Operations Boundary NW12.

2. The MITR-II containment building

3. The area within a 100-meter radius round the containment building

4. The area including the containment building, the fenced in area between the containment building and the Boston &

Albany Railroad, the utilities service building, the cooling tower equipment housing, all of building NW-12, and the reactor parking lot.

QUESTION B.08 [1.0 point]

In accordance with MP 5.2.1, Low Flow MF-2, all of the following conditions could cause an alarm EXCEPT:

a. Primary to secondary leak inside the heat exchanger.

b. Coolant flow through the core is below 1900 gpm.

c. Cavitation of MM-2.

d. Blockage or other issue inside the primary cleanup loop.

Category B: Normal/Emergency Operating Procedures and Radiological Controls

QUESTION B.09 [1.0 point]

According to MITR-II Technical Specifications, which ONE of the following conditions is NOT permissible?

e. Operation with neutron sources < 500 W.

f. Control rod drop time < 1 s.

g. Shutdown margin < 1% K/K.

h. Maximum reactivity addition rate < 5 x 10-4 K/K.

_&@[]j vmU9 °Um9ypoint6ch Match the Technical Specification required surveillance in Column A with the surveillance frequency in Column B. Options in Column B may be used once, more than once, or not at all.

Column A Column B

a. Channel test of Stack Particulate 1. Monthly Monitor

2. Quarterly

b. Control blade withdraw permit interlock

3. Annually

c. Containment leak rate testing

4. Biennially

d. Conductivity of fuel storage pool water

QUESTION B.11 [1.0 point]

In accordance with 10 CFR 20, individual members of the public are limited to a dose in an unrestricted area from an external source:

a. 2000 mrem/hr.

b. 200 mrem/hr.

c. 20 mrem/hr.

d. 2 mrem/hr.

Category B: Normal/Emergency Operating Procedures and Radiological Controls

QUESTION B.12 [2.0 point, 0.50 each]

Match the Equipment Change in Column A to the Change Classification type in Column B.

Options in Column B may be used once, more than once, or not at all.

Column A Column B

a. Replacing regulating rod limit switches 1. Class A with spares

2. Class B

b. Replacing the Reactor Floor noble gas monitor with an energy-compensated 3. Class C GM

c. Replacing the regulating rod with a boron rod

d. Changing the frequency of the pressure relief system charcoal test to semiannually.

QUESTION B.13 [1.0 point]

The radiation level in the control room is 125 mrem/hour, and the operator is in the control room for 17 minutes. How much dose will the operator receive?

a. 7.35 mrem

b. 16.7 mrem

c. 22.8 mrem

d. 35.4 mrem

QUESTION B.14 [1.0 point]

In accordance with PM 5.2.2, No Overflow Core Tank, which ONE of the following conditions could NOT be the cause of the alarm?

a. Blockage in the cleanup loop.

b. Pump cavitation.

c. Primary coolant flow is less than 1900 gpm.

d. The tank overflow line shutter is open.

Category B: Normal/Emergency Operating Procedures and Radiological Controls

QUESTION B.15 [1.0 point]

You are currently a licensed operator at the MITR-II. Which ONE of the following conditions would be a violation of 10 CFR 55.53, Conditions of licenses?

a. The new requalification program cycle started 24 months ago.

b. Your last requalification written examination was 13 months ago.

c. Last quarter, you were the licensed operator for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

d. Your last medical exam was 30 months ago.

QUESTION B.16 [1.0 point]

Which ONE of the following conditions would require a Non-Routine Startup to be performed in accordance with PM 2.3.2?

a. Prior to bringing the reactor critical for the first time at the onset of cold weather.

b. Prior to restart following a major scram where the cause of the scram is known.

c. Following introduction of an experiment that changes the core radial power distribution.

d. Before performing control blade calibrations, maintenance, or surveillance.

QUESTION B.17 [1.0 point]

In accordance with PM 4.4.4.13, Reactor Reentry, in the event of a radiological emergency, all of the following statements are true EXCEPT:

a. Authorization for reentry is the responsibility of the Emergency Director.

b. MITR Radiation Protection Officer should be contacted for approval if it is anticipated personnel will receive exposure in excess of 10 CFR Part 20 limits.

c. Reentry crew must be comprised of volunteers.

d. Authorizations in excess of 25 rem are permissible to maintain nuclear safety.

Category B: Normal/Emergency Operating Procedures and Radiological Controls

QUESTION B.18 [1.0 point]

All of the following are true statements regarding MITR-II Technical Specification surveillance requirements EXCEPT:

e. Surveillance tests for required equipment may be waived when an instrument, component or system is not required to be operable, but such instruments, components, or systems must be tested prior to being used as a required operable instrument, component, or system.

f. All surveillance requirements have a 25% grace period of the specified interval.

g. Surveillance requirements that are deferred during reactor shutdown must be performed as soon as possible after the reactor restarts.

h. The total maximum combined interval time for any three consecutive surveillance intervals may not exceed 3.25 times the specified interval.

QUESTION B.19 [1.0 point]

In accordance with the PM 5.3.1, Low Flow D2O, all of the following actions are to be taken immediately EXCEPT:

a. Stop and attempt to restart the D2O pump.

b. Verify reactor power is decreasing.

c. Dump reflector if a major leak is suspected.

d. Notify Reactor Radiation Protection if a leak is suspected.

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

Category C: Facility and Radiation Monitoring Systems

QUESTION C.01 [1.0 point]

Which ONE of the following statements regarding the subcritical interlock is NOT true?

a. The subcritical interlock may be bypassed to provide a convenient reference point at which the operator may pause to make complete instrument checks before bringing the reactor critical.

b. The subcritical interlock is important to maintain the shim blade bank at a uniform height during the final approach to criticality.

c. When the subcritical interlock is engaged, the shim blades can be driven into the core but may not be withdrawn.

d. The subcritical interlock establishes a level below the critical position at which the shim blades may be individually withdrawn in one step.

QUESTION C.02 [1.0 point]

In accordance with the MITR-II Technical Specifications, which ONE of the following is NOT required to be powered by emergency electrical power during a loss of normal electrical power event?

a. Digital rod position indication.

b. Indication of reactor power levels.

c. Stack effluent radiation monitors.

d. Containment intercom system.

QUESTION C.03 [1.0 point]

Which ONE of the following isotopes would indicate a potential primary to secondary coolant leak in the heat exchanger?

a. Ar-41

b. I-135

c. H-3

d. Na-24 Category C: Facility and Radiation Monitoring Systems

QUESTION C.04 [1.0 point]

Which ONE of the following options correctly describes how primary coolant temperature is measured?

a. A semiconductor-based temperature sensor detects changes in pool temperature by utilizing two identical diodes with temperature-sensitive voltage vs current characteristics that are used to monitor changes in temperature.

b. A resistance temperature detector detects changes in pool temperature by measuring changes in resistance to the flow of electricity resulting from changes in temperature of the resistive metal element.

c. A thermistor detects changes in pool temperature by using a thermally sensitive resistor that exhibits a continuous, small, incremental change in resistance correlated to variations in temperature.

d. A thermocouple detects changes in pool temperature by measuring the voltage difference between two wires of dissimilar metals.

QUESTION C.05 [1.0 point]

All of the following statements regarding the Containment system are true EXCEPT:

a. If differential pressure exceeds +2.0 inches of water, the reactor will scram.

b. If differential pressure is below -0.10 inches of water, the reactor may not be started.

c. If differential pressure exceeds -0.10 inches of water, the ventilation fans will trip.

d. If differential pressure exceeds +2.0 inches of water, the Reactor Building Overpressure annunciator will alarm.

Category C: Facility and Radiation Monitoring Systems

QUESTION C.06 [1.0 point]

In accordance with MITR-II Technical Specifications, all of the following statements regarding the Radiation Monitoring systems are true EXCEPT:

a. Whenever the reactor floor is occupied, at least one area radiation monitor must be operable and capable of notifying personnel of high radiation levels. If any area monitor is inoperable and work is to be done in that area, portable instruments shall be used to survey radiation in that area.

b. Whenever containment is not isolated and integrity is required, a minimum of one stack effluent monitor shall be operable and capable of indicating and alarming in the control room.

c. An installed instrument capable of detecting fission products shall be used to monitor the effluent in the purge gas that is drawn through the space between the reactor top lid and the surface of the primary coolant.

d. Whenever the containment building is occupied, a continuous air monitor with an audible alarm and means of recording data shall be operable. Portable instruments, surveys, or analyses may be substituted for the installed monitor for periods of one week or until the next scheduled outage in cases where the MITR-II is scheduled to continuously operate.

QUESTION C.07 [1.0 point]

Within the exhaust air system, which ONE of the following statements best describes the purpose of the plenum?

a. The plenum provides indication of radiation levels inside the exhaust air system and will cause the intake and exhaust butterfly dampers to close and intake, exhaust and auxiliary fans to stop if operating limits are exceeded.

b. The plenum removes fission product gases and particulates entering the exhaust air system via a series of coarse and absolute filters to meet regulatory release limits.

c. The plenum delays the movement of air past the plenum monitors to ensure that if the plenum monitors alarm and ventilation trips the butterfly dampers will close and seal the building.

d. The plenum regulates the amount of air drawn from the reactor building to make it more or less negative relative to atmosphere to maintain building differential pressure within operating limits.

Category C: Facility and Radiation Monitoring Systems

QUESTION C.08 [1.0 point]

Which ONE of the following best describes the regulating rod drive assembly?

a. The regulating rod drive assembly consists of two 2-phase electric servo motors in parallel driving a lead-screw nut combination

b. The regulating rod drive assembly consists of a large compression spring with a ball nut and screw jack.

c. The regulating rod drive assembly consists of an electromechanical drive, with a stepper motor, magnetic rod coupler, and rack-and-pinion gear system.

d. The regulating rod drive assembly consists of an electromagnetically coupled motor, gear reduction system, and an acme screw type drive.

QUESTION C.09 [1.0 point]

Which ONE of the following statements best describes the series of events that result in the D2O reflector being dumped on a major scram?

a. CV-91 shuts off the air supply to dump valve DV-4 and vents the air chamber above the diaphragm. DV-4 opens and dumps the top 20 of the reflector.

b. CV-90 mechanically shuts off the air supply to dump valve DV-4, allowing air pressure to slowly bleed off. This will cause the dump valve to drift open, dumping the top 20 of the reflector.

c. CV-90 shuts off the air supply to dump valve DV-4 and vents the air chamber above the diaphragm. DV-4 opens and dumps the top 20 of the reflector.

d. CV-91 mechanically shuts off the air supply to dump valve DV-4, allowing air pressure to slowly bleed off. This will cause the dump valve to drift open, dumping the top 20 of the reflector.

QUESTION C.10 [1.0 point]

Which ONE of the following best describes a fuel element?

a. Each fuel element contains approximately 34 g U-238. Each fuel plate contains approximately 509 g enriched U-238 in the form of UAlx.

b. Each fuel element contains approximately 509 g U-238. Each fuel plate contains approximately 34 g enriched U-238 in the form of UAlx.

c. Each fuel element contains approximately 34 g U-235. Each fuel plate contains approximately 509 g enriched U-235 in the form of UAlx.

d. Each fuel element contains approximately 509 g U-235. Each fuel plate contains approximately 34 g enriched U-235 in the form of UAlx.

Category C: Facility and Radiation Monitoring Systems

QUESTION C.11 [1.0 point]

Which ONE of the following describes how the anti-siphon valves work?

a. In the event of a pipe break, the heat exchanger outlet valves close to prevent continued loss of coolant from the inlet plenum and ensure sufficient coolant remains above the core.

b. In the event of a pipe break, two float valves at the top of the core shroud replace the water in the inlet plenum with air to stop the loss of coolant and ensure sufficient coolant remains above the core.

c. In the event of a pipe break, the standpipe fills with air to stop the loss of coolant and ensure sufficient coolant remains above the core.

d. In the event of a pipe break, the air in the off-gas pipe is replaced with water to slow the loss of coolant and ensure sufficient coolant remains above the core.

QUESTION C.12 [1.0 point]

Which ONE of the following statements best describes the importance of maintaining a containment pressure differential?

a. Operating under a positive pressure differential ensures all leakage is into containment rather than out, which allows control of release of radioactive effluents to the environment during all normal and abnormal operating conditions.

b. Operating under a positive pressure differential ensures all leakage is out of containment rather than into it, which allows control of release of radioactive effluents to the environment through the stack and radiological controls for occupational workers.

c. Operating under a negative pressure differential ensures all leakage is into containment rather than out, which allows control of release of radioactive effluents to the environment during all normal and abnormal operating conditions.

d. Operating under a negative pressure differential ensures all leakage is out of containment rather than into it, which allows control of release of radioactive effluents to the environment through the stack and radiological controls for occupational workers.

Category C: Facility and Radiation Monitoring Systems

QUESTION C.13 [1.0 point]

Which ONE of the following actions would be considered a Channel CHECK?

e. Comparison of Nuclear Channel 2 indications with the N-16 monitor indications of reactor power.

f. Comparison of current Nuclear Channel 2 indications with Nuclear Channel 2 indications at the same power levels during previous reactor operations.

g. Adjustment of Nuclear Channel 2 indications while at power to reflect the N-16 monitor indications of reactor power more closely.

h. Adjustment of Nuclear Channel 2 indications following performance of the thermal power calibration to correct for acceptable instrument drift.

QUESTION C.14 [1.0 point, 0.25 each]

Match the reactor parameter in Column A with the appropriate action in Column B. Options in Column B may be used once, more than once, or not at all.

Column A Column B

a. MTS-1 at 47° C 1. Alarm only

b. MTS-1 at 50° C 2. Alarm and scram

c. MTS-1 at 53° C 3. Scram only

d. MTS-1 at 55° C 4. No action

QUESTION C.15 [1.0 point]

During normal, steady state reactor operation at 5.9 MW, the Control on Manual annunciator alarms. Which ONE of the following conditions could NOT have caused this alarm?

e. The regulating rod was inserted to within 1.5 of near-in.

f. An experiment with 0.5% K/K reactivity was inadvertently removed from the pneumatic transfer system.

g. Power set point and actual reactor power deviation was 1.5%.

h. The regulating rod reached its full out position.

Category C: Facility and Radiation Monitoring Systems

QUESTION C.16 [1.0 point]

Which ONE of the following best describes how Nuclear Channel 4 operates?

a. Nuclear Channel 4 is a dual fission chamber, with one chamber filled with BF3 gas and the other filled with an inert gas. Neutrons fission with the Boron to produce alpha particles which ionize the gas. Gammas also ionize the fill gases in both chambers. The gamma signal is subtracted from the neutron signal to provide an indication of reactor power over the full range of operation.

b. Nuclear Channel 4 is a fission chamber lined with highly enriched U-235. Neutrons interact with the U-235 to produce fission, which ionize the fill gas. Gammas also ionize the fill gas.

At low powers, the circuitry uses a pulse height discriminator to differentiate the neutrons from the gammas to provide an indication of reactor power.

c. Nuclear Channel 4 is a fission chamber lined with highly enriched U-238. Neutrons interact with the U-238 to produce fission, which ionize the fill gas. Gammas also ionize the fill gas.

At low powers, the circuitry uses a pulse height discriminator to differentiate the neutrons from the gammas to provide an indication of reactor power.

d. Nuclear Channel 4 is a fission chamber lined with B-10. Neutrons fission with the B-10 to produce alpha particles which ionize the gas. Gammas also ionize the fill gas. The combined signal provides an indication of reactor power.

QUESTION C.17 [1.0 point]

A gaseous effluent commonly produced from reactor operation is _______ which is _______.

a. Rn-222; a naturally occurring isotope

b. I-135; produced as a byproduct of fission

c. F-19; produced from irradiation of water

d. Ar-41; produced from irradiation of air

QUESTION C.18 [1.0 point]

All of the following statements regarding the core purge blower are true EXCEPT:

a. The core purge blower will trip when either of the main or auxiliary dampers close.

b. The core purge blower will immediately cause an automatic reactor scram if the blower is stopped or turned off.

c. The core purge blower will cause an automatic reactor scram approximately 4 minutes after system flow drops to 2.5 cfm.

d. The core purge blower will trip when the exhaust fan stops.

Category C: Facility and Radiation Monitoring Systems

QUESTION C.19 [1.0 point]

Which ONE of the following best describes the forced convection flow path for the primary coolant system through the reactor vessel?

a. Heated coolant is drawn from the upper core tank above the core by the primary coolant pumps into the main heat exchanger, then to the inlet plenum. The coolant flows through the annular space between the core tank and core shroud, flows around the outside of the core, then up through the core into the upper core tank.

b. Heated coolant is drawn from the outlet plenum by the primary coolant pumps into the main heat exchangers, then to the upper core tank. The coolant flows around the outside of the core, then up through the annular space between the core tank and core shroud

c. Heated coolant rises from the core to the level of the inlet plenum valves and anti-siphon valves, which force coolant down from the outlet plenum above the core to the bottom of the core.

d. Heated coolant is drawn from the outlet plenum above the core by the primary coolant pumps into the main heat exchanger, then to the inlet plenum. The inlet plenum valves and anti-siphon valves force coolant down from the outlet plenum above the core to the bottom of the core.

QUESTION C.20 [1.0 point]

Which ONE of the following best describes the reason for the high sensitivity of a Geiger-Mueller detector?

a. It is coated with special nuclear material that causes high ionizations at low concentrations.

b. It has a large tube, so the target area is bigger for all incident events.

c. The lower voltage applied to the detector helps to amplify all incident events.

d. Any incident radiation event causing primary ionization results in ionization of the entire detector,

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

(***** END OF EXAMINATION *****)

Category A: Reactor Theory, Thermodynamics, and Facility Operating Characteristics

A.01 Answer: d.

Reference:

LaMarsh, Introduction to Nuclear Engineering, Page 340-341 (1 ) = 1 is a prompt critical condition. Rearranging this equation results in

= 1 1

Answer: c.

Reference:

Burn, Introduction to Nuclear Reactor Operations, Section 3.3.2, p 3-16

A.03 Answer: d.

Reference:

Reactor power keeps increasing because period is positive.

A.04 Answer: a.

Reference:

Burn, Introduction to Nuclear Reactor Operations, Section 2.7, 2-63

A.05 Answer: c.

Reference:

DOE Fundamentals Handbook, Vol. 1, p. 29

A.06 Answer: a.

Reference:

DOE Fundamentals Handbook, Volume 2, p. 34

A.07 Answer: b.

Reference:

P = P0et/17 3 = 1e t/17 ln(3) = ln(et/17) 1.099 = t/17 T = 18.7 seconds

A.08 Answer: d.

Reference:

DOE Fundamentals Handbook, Vol. 2, p 15

A.09 Answer: a. 3; b. 1; c. 4; d. 2

Reference:

DOE Fundamentals Handbook, Module 1, p. 29

A.10 Answer: a. 4 (Te-135); b. 1 (I-135); c. 2 (Cs-135); d. 3 (Ba-135)

Reference:

DOE Fundamentals Handbook, Volume 2, p. 35

A.11 Answer: d.

Reference:

DOE Fundamentals Handbook, Volume 2, Module 3, p. 28 Category A: Reactor Theory, Thermodynamics, and Facility Operating Characteristics

A.12 Answer: d.

Reference:

DOE Fundamentals Handbook, Vol. 1, p. 55

A.13 Answer: a. 2 (Fertile); b. 3 (Fissile); c. 3 (Fissile); d. 1 (Fissionable)

Reference:

DOE Fundamentals Handbook, Volume 1, p. 50-52

A.14 Answer: b.

Reference:

Burn, Introduction to Nuclear Reactor Operations, Section 3.3.4, p 3-20-21

= (k eff2-keff1)/(keff1*keff2)

= (0.872-0.959)/(0.872*0.959)

= -0.104 k/k

Answer: c.

Reference:

Burn, Introduction to Nuclear Reactor Operations, Section 7, page 7-4

A.16 Answer: d.

Reference:

Burn, Section 5.5, p. 5-18

A.17 Answer: d.

Reference:

DOE Fundamentals Handbook, Vol. 1, p. 47

A.18 Answer: b.

Reference:

Burn, Introduction to Nuclear Reactor Operations, Section 3.2.3, p. 3-11

A.19 Answer: b.

Reference:

DOE Fundamentals Handbook, Volume 1, p. 61

A.20 Answer: a.

Reference:

DOE Fundamentals Handbook, Module 3, p. 58

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

Category B: Normal/Emergency Operating Procedures and Radiological Controls

B.01 Answer: b.

Reference:

Emergency Procedure 4.4.4.11, NW12 Evacuation

B.02 Answer: c.

Reference:

DR=DR, T1 = 0.693 0

2 DR = DR0 e-.693/T1/2 65 = 135 e-(.693)(3)/T1/2 0.481 = e-(.693)(3)/T1/2 ln(0.481) = ln(e-(.693)(3)/T1/2)

-0.731 = -2.079 / T1/2 T1/2 = -2.079/ -0.731 T1/2 = 2.84 hr

B.03 Answer: b.

Reference:

PM 2.4.3, One Pump Operation

B.04 Answer: a.

Reference:

10 CFR 20.1003

B.05 Answer: b.

Reference:

MITR-II Emergency Plan

B.06 Answer: c.

Reference:

DR1*(D1)² = DR2*(D2)² 15000 mrem (3 ft)² = 400 mrem (d)² 135000 mrem-ft2 = 400 mrem (d)2 337.5 ft2 = d2 D = 18.37 ft.

B.07 Answer: a. 3; b. 4; c. 1; d. 2

Reference:

MITR-II Emergency Plan

B.08 Answer: b.

Reference:

PM 5.2.1, Low Flow MF-2

B.09 Answer: c.

Reference:

MITR-II Technical Specifications 3.1.3, 3.1.4, 3.2.1, 3.2.2

B.10 Answer: a. 1 (Monthly); b. 3 (Annually); c. 4 (Biennially); d. 2 (Quarterly)

Reference:

MITR-II Technical Specifications 4.2, 4.3, 4.4, 4.7 Category B: Normal/Emergency Operating Procedures and Radiological Controls

B.11 Answer: d.

Reference:

10 CFR 20.1301

B.12 Answer: a. 3 (Class C); b. 2 (Class B); c. 1 (Class A); d. 2 (Class B)

Reference:

PM 1.4, Review and Approval of Plans, Procedures, and Facility

B.13 Answer: d.

Reference:

Dose = DR*T 125 mRem/hr/60 minutes = 2.08 mRem/min 2.08 mRem/min

• 17 min = 35.4 mRem

B.14 Answer: a.

Reference:

PM 5.2.2, No Overflow Core Tank

B.15 Answer: d.

Reference:

10 CFR 55.53, Conditions of Licenses

B.16 Answer: c.

Reference:

PM 2.3.2, Non-Routine Startup

B.17 Answer: d.

Reference:

PM 4.4.4.13, Reactor Reentry

B.18 Answer: a.

Reference:

MITR-II Technical Specifications, 1.3.11 (1.3.11 states this is true for EXPERIMENTS, not equipment)

B.19 Answer: a.

Reference:

PM 5.3.1, Low Flow D2O

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

Category B: Normal/Emergency Operating Procedures and Radiological Controls

C.01 Answer: a.

Reference:

MITR-II Reactor Systems Manual, 4.2

C.02 Answer: a.

Reference:

MITR-II Technical Specifications, 3.6

C.03 Answer: d.

Reference:

MITR-II Reactor Systems Manual, 3.4.2.3

C.04 Answer: b.

Reference:

MITR-II Reactor Systems Manual, 6.3.1

C.05 Answer: c.

Reference:

MITR-II Reactor Systems Manual, 6.7

C.06 Answer: d.

Reference:

MITR-II Technical Specifications 3.7.1

C.07 Answer: c.

Reference:

MITR-II Reactor Systems Manual, 8.3.6

C.08 Answer: a.

Reference:

MITR-II Reactor Systems Manual, 1.3.2

C.09 Answer: c.

Reference:

MITR-II Reactor Systems Manual 8.6.2

C.10 Answer: d.

Reference:

MITR-II Reactor Systems Manual 1.2.1

C.11 Answer: b.

Reference:

MITR-II Reactor Systems Manual 1.4.1.1

C.12 Answer: c.

Reference:

MITR-II Technical Specifications 1.3.5

C.13 Answer: a. or b.

Reference:

MITR-II Technical Specifications 1.3.3 Category B: Normal/Emergency Operating Procedures and Radiological Controls

C.14 Answer: a. 4 (No action); b. 4 (No action); c. 1 (Alarm only); d. 2 (Alarm and scram)

Reference:

MITR-II Reactor Systems Manual 9.1 and 9.2

C.15 Answer: d.

Reference:

MITR-II Reactor Systems Manual, 4.4

C.16 Answer: b.

Reference:

MITR-II Reactor Systems Manual 5.2.1, NRC Standard Question

C.17 Answer: d.

Reference:

MITR-II Reactor Systems Manual 7.3

C.18 Answer: b.

Reference:

MITR-II Reactor Systems Manual 8.3.4

C.19 Answer: a.

Reference:

MITR-II Reactor Systems Manual 3.2

C.20 Answer: d.

Reference:

NRC Standard question

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

(***** END OF EXAMINATION *****)