ML23199A293
| ML23199A293 | |
| Person / Time | |
|---|---|
| Site: | 05000083 |
| Issue date: | 07/20/2023 |
| From: | Travis Tate NRC/NRR/DANU/UNPO |
| To: | Wall D Univ of Florida |
| References | |
| 50-083/OL-023 | |
| Download: ML23199A293 (38) | |
Text
July 18, 2023 Dr. Donald Wall, Facility Director University of Florida Training Reactor 202 Nuclear Science Center P.O. Box 116400 Gainesville, FL 32611-8300
SUBJECT:
EXAMINATION REPORT NO. 50-083/OL-23-01, UNIVERSITY OF FLORIDA
Dear Dr. Wall:
During the week of June 13, 2023, the U.S. Nuclear Regulatory Commission (NRC) administered an operator licensing examination at your University of Florida training 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 me at (301) 415-3901 or via email at Travis.Tate@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.50-083
Enclosures:
- 1. Examination Report No. 50-083/OL-23-01
- 2. Written examination cc: w/enclosures to GovDelivery Subscribers Signed by Tate, Travis on 07/18/23
ML23199A293 NRR-079 OFFICE NRR/DANU/UNPO/OLA NRR/DANU/UNPO/BC NAME NJones TTate DATE 7/18/2023 7/18/2023
ENCLOSURE 1 U. S. NUCLEAR REGULATORY COMMISSION OPERATOR LICENSING INITIAL EXAMINATION REPORT REPORT NO.:
50-083/OL-23-01 FACILITY DOCKET NO.:
50-083 FACILITY LICENSE NO.:
R-56 FACILITY:
University of Florida Training Reactor EXAMINATION DATES:
June 13 - 14, 2023 SUBMITTED BY: _
__06/27/2023__
Paulette Torres, Chief Examiner Date
SUMMARY
During the week of June 13, 2022, the NRC administered operator licensing examinations to two Reactor Operator (RO) candidates. The two RO candidates passed all applicable portions of the examinations.
REPORT DETAILS 1.
Examiners:
Paulette Torres, Chief Examiner, NRC 2.
Results:
RO PASS/FAIL SRO PASS/FAIL TOTAL PASS/FAIL Written 2/0 0/0 2/0 Operating Tests 2/0 0/0 2/0 Overall 2/0 0/0 2/0 3.
Exit Meeting:
Paulette Torres, Chief Examiner, NRC Brian Shea, Reactor Manager, University of Florida Training Reactor Jyothier Kumar Nimmagadda, SRO, University of Florida Training Reactor Prior to administration of the written examination, based on facility comments, adjustments to the written examination were accepted. Comments provided corrections and additional clarity to questions/answers and identified where changes were appropriate based on current facility conditions. At the exit meeting, the NRC examiner thanked the facility for their support in the administration of the examination.
ENCLOSURE 2 U. S. NUCLEAR REGULATORY COMMISSION NON-POWER REACTOR LICENSE EXAMINATION FACILITY:
UFL REACTOR TYPE:
ARGONAUT DATE ADMINISTERED:
06/14/2023 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
A. Reactor Theory, Thermohydraulics & 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 ___
A10 a b c d ___
A11 a b c d ___
A12 a b c d ___
A13 a b c d ___
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 SECTION A *****)
B. Normal/Emergency 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 ___
B08 a b c d ___
B09 a b c d ___
B10 a b c d ___
B11 a b c d ___
B12 a b c d ___
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 ___
B20 a b c d ___
(***** END OF SECTION B *****)
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 ___
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 SECTION C *****)
(********** END OF EXAMINATION **********)
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.
EQUATION SHEET
=
DR - Rem, Ci - curies, E - Mev, R - feet 1 Curie = 3.7 x 1010 dis/sec 1 kg = 2.21 lbm 1 Horsepower = 2.54 x 103 BTU/hr 1 Mw = 3.41 x 106 BTU/hr 1 BTU = 778 ft-lbf
°F = 9/5 °C + 32 1 gal (H2O) 8 lbm
°C = 5/9 (°F - 32) cP = 1.0 BTU/hr/lbm/°F cp = 1 cal/sec/gm/°C 1ft = 30.48 cm
2 2
max
P 1
sec 1.0
eff
2 1
1 1
2 1
eff eff K
CR K
CR
eff SUR 06 26
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P 0
sec 10 1
4
eff K
S S
1
2 2
1 1
CR CR
0 1
P P
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(
0 10 t
SUR P
P 1
2 1
1 CR CR K
M eff
2 1
1 1
eff eff K
K M
eff eff K
K SDM
1
693
.0 2
1 T
eff 2
1 1
2 eff eff eff eff K
K K
K
eff eff K
K 1
2 2
2 2
1 1
d DR d
DR
t e
DR DR
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1 2
1 2
2 2
Peak Peak
2 6
R n
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University of Florida Operator Licensing Examination Week of June 13, 2023
Section A: Reactor Theory, Thermohydraulics & Facility Operating Characteristics Page 2 QUESTION A.01
[1.0 point]
Energy Yield (Q) from a nuclear fission reaction is in the range of (or is approximately):
a.
< 1 eV b.
1.86 keV c.
200 MeV d.
1000 MeV QUESTION A.02
[1.0 point]
A reactor is subcritical if:
a.
= 1.0 b.
Keff < 1.0 or < 0.0 c.
K= 1.0, =
d.
Keff > 1.0 or > 0.0 QUESTION A.03
[1.0 point]
Which ONE of the following correctly describes the relationship between differential rod worth (DRW) and integral rod worth (IRW)?
a.
DRW is the slope of the IRW curve at a given location.
b.
DRW is the area under the IRW curve at a given location.
c.
DRW is the square root of the IRW curve at a given location.
d.
There is no relationship between DRW and IRW.
Section A: Reactor Theory, Thermohydraulics & Facility Operating Characteristics Page 3 QUESTION A.04
[1.0 point]
Which ONE of the following changes does not require a movement of control rods in order to maintain constant reactor power?
a.
Pool water temperature decrease b.
U-235 burnup c.
Xe-135 buildup d.
N-16 formation QUESTION A.05
[1.0 point]
The effective multiplication factor (Keff) can be determined by dividing the number of neutrons produced from fission in the fourth generation by the number of neutrons produced from fission in the _________ generation.
a.
First b.
Second c.
Third d.
Fifth QUESTION A.06
[1.0 point]
Because the temperature of the fuel reacts immediately to changes in reactor power, the Fuel Temperature Coefficient is also called the:
a.
Prompt Temperature Coefficient b.
Moderator Temperature Coefficient c.
Nuclear Doppler Effect d.
Void Coefficient
Section A: Reactor Theory, Thermohydraulics & Facility Operating Characteristics Page 4 QUESTION A.07
[1.0 point, 0.25 each]
Match the items in Column A with the isotopes in Column B.
The most important fission product poison is 135Xe. The process that show how this isotope is formed and its decay is:
Column A Column B 1.
135 Ba 2.
135 Cs 3.
135 I 4.
135 Te QUESTION A.08
[1.0 point]
Delayed neutrons contribute more to reactor stability than prompt neutrons because they ________
the average neutron generation time and are born at a __________ kinetic energy.
a.
Decrease, lower b.
Increase, lower c.
Decrease, higher d.
Increase, higher QUESTION A.09
[1.0 point]
Which ONE of the following is the MOST effected factor in the six factor formula due to fuel burnup?
a.
Fast Fission Factor.
b.
Reproduction Factor.
c.
Thermal Utilization Factor.
d.
Resonance Escape Probability.
Section A: Reactor Theory, Thermohydraulics & Facility Operating Characteristics Page 5 QUESTION A.10
[1.0 point]
Reactivity is defined as the:
a.
Fractional change in neutron population per generation.
b.
Number of neutrons by which population changes per generation.
c.
Rate of change of reactor power in neutron per second.
d.
Change in the number of neutrons per second that causes a fission event.
QUESTION A.11
[1.0 point]
The term __________ defines the condition where no delay neutrons are required.
a.
Prompt Jump b.
Prompt Drop c.
Asyptotic Period d.
Prompt Critical QUESTION A.12
[1.0 point]
What order process is the radioactive decay differential equation?
a.
Zero b.
First c.
Second d.
Third
Section A: Reactor Theory, Thermohydraulics & Facility Operating Characteristics Page 6 QUESTION A.13
[1.0 point]
Which ONE of the following is the major source of heat generation after an operating reactor has been shut down and cooled down for several days?
a.
Resonance Capture b.
Fission Fragment Decay c.
Delayed Neutron Reactions d.
Corrosion Product Activation QUESTION A.14
[1.0 point]
What is the average number of neutrons produced from every fission () of Uranium-235 with thermal neutrons?
a.
2.42 neutrons b.
2.66 neutrons c.
2.81 neutrons d.
2.93 neutrons QUESTION A.15
[1.0 point]
The reaction 93Np239 _____ + 94Pu239 is an example of:
a.
Alpha Decay b.
Beta Decay c.
Gamma Emission d.
Electron Capture
Section A: Reactor Theory, Thermohydraulics & Facility Operating Characteristics Page 7 QUESTION A.16
[1.0 point]
What is the effect of delayed neutrons on the neutron flux decay following a scram from full power?
a.
Adds negative reactivity creating a greater shutdown margin.
b.
Adds positive reactivity due to the fuel temperature decrease following the scram.
c.
Limits the final rate at which power decreases to a -80 second period.
d.
Decreases the mean neutron lifetime.
QUESTION A.17
[1.0 point]
About two minutes following a reactor scram, period has stabilized, and is decreasing at a CONSTANT rate. If reactor power is 10-5 % full power what will the power be in three minutes?
a.
5 x 10-6 % full power b.
2 x 10-6 % full power c.
1 x 10-6 % full power d.
5 x 10-7 % full power QUESTION A.18
[1.0 point]
Which ONE of the following most accurately describes the reason that fission products such as Xenon-135 and Samarium-149 have the most substantial impact in reactor design and operation?
a.
Xenon-135 and Samarium-149 cause excess positive reactivity in the core.
b.
Xenon-135 and Samarium-149 burn up causes an increase in the thermal flux.
c.
Xenon-135 and Samarium-149 have large absorption cross sections resulting in a large removal of neutrons from the reactor.
d.
Xenon-135 and Samarium-149 produce fast fission neutrons, resulting in the net increase in the fast neutron population of the reactor core.
Section A: Reactor Theory, Thermohydraulics & Facility Operating Characteristics Page 8 QUESTION A.19
[1.0 point]
If Beta for U-235 is 0.0065 and Beta effective is approximately 0.007, how does this difference affect reactor period in the reactor period equation, T=(-p)/p? This difference produces a
_________ for a given addition of reactivity with Beta effective.
a.
Longer Period b.
Shorter Period c.
Stable Period d.
Decay Constant () Increase QUESTION A.20
[1.0 point]
Which ONE of the following is defined as the balance between production of neutrons and their absorption in the core for which core leakage can be neglected?
a.
Utilization Factor b.
Reproduction Factor c.
Infinite Multiplication Factor d.
Effective Multiplication Factor
- End of Section A *****************
Section B: Normal/Emergency Procedures and Radiological Controls Page 9 QUESTION B.01
[1.0 point]
What is the HALF LIFE of the isotope contained in a sample which produces the following count rates?
Time (Minutes)
Counts per Minute (cpm)
Initial count 840 30 696 60 575 90 475 180 270 a.
310 minutes b.
210 minutes c.
110 minutes d.
60 minutes QUESTION B.02
[1.0 point]
Which ONE of the following does NOT require NRC approval for changes?
a.
Technical Specifications incorporated in the license b.
Requalification Plan c.
Emergency Procedures d.
Section B: Normal/Emergency Procedures and Radiological Controls Page 10 QUESTION B.03
[1.0 point]
The exposure rate for a point source is 100 mR/hr at a distance of 4 m. What is the exposure rate at a distance of 2 m?
a.
200 mR/hr b.
400 mR/hr c.
600 mR/hr d.
800 mR/hr QUESTION B.04
[1.0 point]
Given the following instruments, which ONE is the best to check your hands and clothing for beta-gamma contamination upon leaving a contamination zone?
a.
Pancake Probe GM Survey Meter b.
Ionization Chamber Survey Instrument c.
Portable Sodium Iodide (NaI) Detector d.
Zinc Sulfide (ZnS) Detector QUESTION B.05
[1.0 point]
You are currently the licensed operator at UFTR. Which ONE of the following will violate your license per 10 CFR Part 55?
a.
Last requalification operating test was 14 months ago.
b.
Last requalification written examination was 20 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.
Last licensed renewal was 48 months ago.
Section B: Normal/Emergency Procedures and Radiological Controls Page 11 QUESTION B.06
[1.0 point]
Per UFTR Technical Specifications, the __________ is (are) based on measurement of first fission product release from the fuel at or above the blister threshold temperature described in NUREG-1313.
a.
Safety Limit b.
Administrative Controls c.
Limiting Safety System Setting d.
Limiting Condition for Operation QUESTION B.07
[1.0 point]
Per Emergency Plan, TLDs are routinely placed in various locations peripheral to the reactor cell and the operations boundary; these devices may be used to provide integrated dose information for post-accident assessment. This is an example of:
a.
Assessment Actions b.
Corrective Actions c.
Protective Actions d.
Recovery Actions QUESTION B.08
[1.0 point]
Per SOP, the resistivity of reactor coolant system water shall be verified no less than?
a.
0.4 Megohm-cm b.
0.5 Megohm-cm c.
0.6 Megohm-cm d.
0.8 Megohm-cm
Section B: Normal/Emergency Procedures and Radiological Controls Page 12 QUESTION B.09
[1.0 point]
In accordance with Technical Specifications, which ONE of the following statements is TRUE?
a.
Each fuel experiment shall be limited such that the total inventory of iodine isotopes 131 thru 135 in the experiment is no greater than 0.01 curies.
b.
The absolute value of the reactivity worth of any single movable experiment shall be less than or equal to 1400 pcm.
c.
Experiments containing explosive materials shall be double encapsulated.
d.
Experiments containing corrosive materials shall not be irradiated.
QUESTION B.10
[1.0 point]
Maintenance Procedure, E.4 Calibration of Reactor Measuring Channels, provides instructions for temperature channels __________.
a.
Alignment b.
Check c.
Calibration d.
Performance QUESTION B.11
[1.0 point]
The proper Reactor Cell Evacuation Alarm Interlock function shall be verified __________.
a.
Daily b.
Annual c.
Weekly d.
Semiannual
Section B: Normal/Emergency Procedures and Radiological Controls Page 13 QUESTION B.12
[1.0 point]
The reactor operator can attain criticality by incremental withdrawal of the __________ such that the stable reactor period is maintained at approximately 30 seconds or greater.
a.
Control Blade #1 b.
Control Blade #2 c.
Control Blade #3 d.
Regulating Blade QUESTION B.13
[1.0 point]
Operating circuit breakers or switches is an example of a __________ item.
a.
Corrective Maintenance b.
Major Maintenance c.
Minor Maintenance d.
Preventive Maintenance QUESTION B.14
[1.0 point]
Recovery operations shall be directed by the Emergency Director, with assistance from the Emergency Coordinator and the _________.
a.
Gainesville Fire Department b.
Radiation Control Officer c.
University of Florida Police Department d.
US Nuclear Regulatory Commission
Section B: Normal/Emergency Procedures and Radiological Controls Page 14 QUESTION B.15
[1.0 point]
The reactor core shall contain four scrammable control blades of swingarm type consisting of all of the following materials EXCEPT:
a.
Aluminum b.
Cadmium c.
Graphite d.
Magnesium QUESTION B.16
[1.0 point]
During an emergency, the first level Emergency Support Center is to be established in the:
a.
Control Room b.
NSB Room 108 c.
NSB Elevator Area d.
NSB Service Drive Area QUESTION B.17
[1.0 point]
Which ONE of the following requires a Level II Radiation Work Permit?
a.
Shielding (with the exception of experiment port shield blocks/plugs, primary equipment pit cover, and the spent fuel pit cover) is to be removed.
b.
Connecting the external electric hot water heater to the primary coolant system.
c.
Irradiated fuel handling.
d.
Fuel loading.
Section B: Normal/Emergency Procedures and Radiological Controls Page 15 QUESTION B.18
[1.0 point]
The __________ is established as an Emergency Planning Zone for the UFTR.
a.
Operations Area b.
Operations Boundary c.
Radiation Control Office d.
Emergency Support Center QUESTION B.19
[1.0 point]
Which ONE of the following is an example of Class 2 event?
a.
Bomb threat toward the reactor.
b.
Fire within the reactor cell that is extinguished within 15 minutes.
c.
Flood in the reactor facility environment.
d.
Fueled experiment failure resulting in significant release of fission products.
QUESTION B.20
[1.0 point]
The reactor condition shall be called MODE 4 when the reactor:
a.
Is secured.
b.
Is shutdown.
c.
Is in outage condition.
d.
Operation is at less than 1% rated thermal power.
- End of Section B ********************************
Category C: Facility and Radiation Monitoring Systems Page 16 QUESTION C.01
[1.0 point]
The typical gamma radiation levels in the reactor cell at full-power is 2 mR/hr for the:
a.
North ARM 1 b.
Control Console c.
Top of shield tank d.
Low level storage area QUESTION C.02
[1.0 point]
Per SAR Table 4-1, Nominal Design Parameters of the UFTR Fuel Assembly Core, 14 corresponds to the number of:
a.
Dummy Assemblies b.
Full Fuel Assemblies c.
Partial Fuel Assemblies d.
Plates per Full Fuel Assembly QUESTION C.03
[1.0 point]
Which ONE of the following events is designated as the Maximum Hypothetical Accident (MHA) for the UFTR?
a.
Fission products release due to severe damage to a fuel assembly.
b.
Malfunction of an experiment.
c.
Loss of Coolant Accident.
d.
Insertion of Excess Reactivity.
Category C: Facility and Radiation Monitoring Systems Page 17 QUESTION C.04
[1.0 point]
All of the following are core characteristics of the PuBe neutron source EXCEPT:
a.
1 Ci b.
Non-regenerable c.
Permanently installed in WVP d.
Alarm on Log-N meter at ~ 100 watts QUESTION C.05
[1.0 point]
The UFTR Fuel Plate Structure is made of __________ U-235 (nominal) per plate.
- a. 3.6 g
- b. 12.5 g c.
29 g
- d. 58 g QUESTION C.06
[1.0 point]
The Safety Channel 1 detector is connected to circuitry containing a __________.
a.
Linear Amplifier b.
Log Amplifier c.
Pre-Amplifier d.
Servo Amplifier
Category C: Facility and Radiation Monitoring Systems Page 18 QUESTION C.07
[1.0 point]
Which ONE of the following describes the compensated ion chamber method for gamma compensation?
a.
Supplies gamma compensation: Source/intermediate range uses pulse height discrimination.
b.
Supplies gamma compensation: Power range is intrinsic to Campbelling Mode.
c.
Gamma compensation is not required since gamma current is proportional to neutron flux in the power range.
d.
Active gamma compensation via negative gamma signal summing.
QUESTION C.08
[1.0 point]
Which ONE of the following conditions will initiate a Full-trip when 2 or more control blades are above their bottom RPS limit switch setting?
a.
Loss of A.C. power.
b.
Loss of power to Stack Dilution fan.
c.
Loss of power to Primary Coolant pump.
d.
Loss of power to the Deep Well pump when log-N 1 kW and using deep well for secondary cooling.
QUESTION C.09
[1.0 point]
On the Control Blade Drive circuit, which ONE of the following indicates the magnetic clutch current is energized?
a.
White DOWN light is illuminated b.
Red UP light is illuminated c.
Yellow ON light is ON d.
Yellow ON light is OFF
Category C: Facility and Radiation Monitoring Systems Page 19 QUESTION C.10
[1.0 point]
A major component of the reactor coolant system is the __________, which is a solenoid-operated valve located in the equipment pit that opens automatically when actuated by a demand or trip signal.
a.
Dump Valve b.
Flow Diverter Valve c.
Priming Vent Valve d.
Sampling Valve QUESTION C.11
[1 point, 0.25 each]
Identify the best answer from the components labeled 1 through 6 on the figure of the vertical section view of the UFTR core illustrating the fuel and fuel box arrangement.
(Note: Only one answer per number.)
a.
Rabbit System _____
b.
Regulating Blade _____
c.
Safety Blade #2 _____
d.
Safety Blade #3 _____
Category C: Facility and Radiation Monitoring Systems Page 20 QUESTION C.12
[1.0 point]
Which ONE of the following figures represent an Integral Worth Curve?
a.
b.
c.
d.
QUESTION C.13
[1.0 point]
Actuation of the evacuation alarm automatically trips the __________ system.
a.
Fire Alarm b.
Reactor Control c.
Reactor HVAC d.
Reactor Protection
Category C: Facility and Radiation Monitoring Systems Page 21 QUESTION C.14
[1.0 point]
Which ONE of the following will result in a control blade withdrawal inhibit?
a.
Log-N signal = 0.5 cps b.
Regulating Rod fully withdrawn c.
Safety Channel 1 Trip Test Switch in "OFF" d.
Reactor Period > 10 seconds QUESTION C.15
[1.0 point]
The __________ has the power to authorize operations in accordance with facility procedures.
a.
Second Person b.
Reactor Manager c.
Senior Operator on Call d.
Radiation Control Officer QUESTION C.16
[1.0 point]
The ________ is a Radiation Monitoring Equipment that functions to monitor airborne particulate, Argon-41 and other gases.
a.
Air Particulate Detector b.
Area Radiation Monitor(s) c.
Portable Air Sampler d.
Stack Monitor
Category C: Facility and Radiation Monitoring Systems Page 22 QUESTION C.17
[1.0 point]
Which ONE of the following reactor coolant system major components is designed to burst at approximately 2 psi above the normal operating system pressure?
a.
Dump Valve b.
Rupture Disk c.
Heat Exchanger d.
Coolant Storage Tank QUESTION C.18
[1.0 point]
Which ONE of the following correctly describes the characteristic of the fuel meat used at the UFTR?
a.
The fuel meat consists of Uranium-silicide-aluminum dispersion fuel with 19.25 wt.%
enriched uranium.
b.
The fuel meat consists of Uranium-silicide-aluminum dispersion fuel with 19.75 wt.%
enriched uranium.
c.
The fuel meat consists of Uranium-sulfate-aluminum dispersion fuel with 19.25 wt.%
enriched uranium.
d.
The fuel meat consists of Uranium-sulfate-aluminum dispersion fuel with 19.75 wt.%
enriched uranium.
QUESTION C.19
[1.0 point]
In the event of a loss of power at the UFTR, which ONE of the following will be provided power from a floating battery pack?
a.
Area Radiation Monitors b.
Emergency Lights c.
Primary Coolant Pump d.
Reactor Control Panel Annunciator Lights
Category C: Facility and Radiation Monitoring Systems Page 23 QUESTION C.20
[1.0 point]
Per Technical Specifications, the keff of all fuel, including fueled EXPERIMENTS and fueled devices, in storage shall be:
a.
no greater than 0.90 b.
equal to 1 c.
less than 0.8 d.
equal to 1.0178
- End of Section C ****************************
- End of the Exam ***************************
UFL OL 23-01 Section A: Theory, Thermohydraulics & Facility Operating Characteristics Page 24 A.01 Answer:
c REF:
Lamarsh 3rd, Table 3.6, pg. 88 A.02 Answer:
b REF:
Burns, Table 3.5, pg. 3-22 A.03 Answer:
a REF:
DOE Fundamentals Handbook, NP-03, pg. 52 A.04 Answer:
d REF:
Burns, Problem 7.7.4, pg. 7-17 A.05 Answer:
c REF:
Burns, Section 3.3.1, pg. 3-16 A.06 Answer:
a REF:
Lamarsh 3rd ed., Section 7.4, pg. 367 A.07 Answer:
a, 4 b, 3 c,2 d,1 REF:
Lamarsh 3rd ed., Section 7.5, pg. 377 Burns, Figure 8.1, pg. 8-6 A.08 Answer:
b REF:
Burns, Section 3.2.4, pg. 3-12 and Section 3.4.4, pg. 3-33 A.09 Answer:
c REF:
Burns, Section 3.3.2, pg. 3-18 A.10 Answer:
a REF:
Burns, Section 1.3.1, pg. 1-5 A.11 Answer:
d REF:
Knife, Nuclear Engineering, 2nd ed., pg. 142
UFL OL 23-01 Section A: Theory, Thermohydraulics & Facility Operating Characteristics Page 25 A.12 Answer:
b REF:
Burns, 2.4.6, pg. 2-30 Mathematically, radioactive decay can be represented by the first order, linear differential equation dA/dt = -A where A is the number density of radioactive atoms of a substance and is called the decay constant.
A.13 Answer:
b REF:
DOE Fundamentals Handbook, NP-03, pg. 34 A.14 Answer:
a REF:
DOE Fundamentals Handbook, NP-03, Table 1, pg. 7 A.15 Answer:
b REF:
DOE Fundamentals Handbook, NP-01, pg. 24 A.16 Answer:
c REF:
Burns, Section 4.10.12, pg. 4-32 to 4-33 A.17 Answer:
c REF:
P = P0 e-(t/T) = 10-5 x e-(180sec/80sec) = 10-5 x e-2.25 = 0.1054 x 10-5 = 1.054 x 10-6 A.18 Answer:
c REF:
DOE Fundamentals Handbook, NP-03, pg. 34 A.19 Answer:
a REF:
Burns, Example 3.4.3, pg. 3-32, 3-33 In the reactor period equation, T=(-p)/p, if Beta effective is used instead of Beta for U-235, the term (eff-p) is larger giving a longer period.
A.20 Answer:
c REF:
DOE Fundamentals Handbook, NP-03, pg. 9
Section B: Normal, Emergency and Radiological Control Procedures Page 26 B.01 Answer:
c REF:
A = A0e-t 270 = 840e-180, 180 = -ln (0.321), = 0.00631 min-1 t1/2 = 0.693 /, = 0.693 / 0.00631 min-1 = 109.8 minutes B.02 Answer:
c REF:
10 CFR 50.59: TS 10 CFR 50.54 (i-1): RP 10 CFR 50.54 (q): EP B.03 Answer:
b REF:
I2=I1D12/d22 = (100 mR/hr)(4m)2 / (2m)2 = 400 mR/hr B.04 Answer:
a REF:
Glasstone, Sesonske, Nuclear Reactor Engineering, Section 9.88, pg. 537 SAR Table 11-4, pg. 11-6 B.05 Answer:
a REF:
10 CFR Parts 55.53, 55.55, 55.59
55.53(h), 55.59(a)(2) - annual operating tests.
55.53(h), 55.59(a)(1), 55.59(c)(1) - The requalification program must be conducted for a continuous period not to exceed 2 years.
55.53(e) - The licensee shall actively perform the functions of a licensed operator for a minimum of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per calendar quarter.
55.55 (a) Each operator license and senior operator license expires 6 years after the date of issuance.
B.06 Answer:
a REF:
TS 2.1, pg. 6 of 43 B.07 Answer:
c REF:
EP 7.3.4, pg. 23 B.08 Answer:
d REF:
SOP A.1, pg. 6 of 26 B.09 Answer:
a REF:
TS 3.8.2 (4), pg. 25 of 43
Section B: Normal, Emergency and Radiological Control Procedures Page 27 B.10 Answer:
c REF:
SOP E.4, Section 7.1.2, pg. 7 of 31 SOP E.4, FORM SOP-E.4B, pg. 23 of 31 B.11 Answer:
c REF:
TS SR 3.4, pg. 18 of 43 B.12 Answer:
d REF:
SOP-A.2, Section 7.22, pg. 7 of 7 B.13 Answer:
c REF:
SOP-0.2, Appendix I, item #4, pg. 11 of 13 B.14 Answer:
b REF:
EP 3.10 (b), pg. 13 B.15 Answer:
c REF:
TS 5.3.1 (2), pg. 31 of 43 B.16 Answer:
b REF:
EP 2.0, pg. 4 EP 8.1, pg. 26 UFTR Training Material Emergency Plan, pg. 7 of 12 B.17 Answer:
a REF:
SOP-D.2, Section 4.4.2, pg. 4 of 16 B.18 Answer:
b REF:
EP 6.0. pg. 19 B.19 Answer:
d REF:
EP Table 5.1, UFTR Emergency Classification Guide, pg. 18 EP 4.3, pg. 16 EP 7.4.2, pg. 24 B.20 Answer:
a REF:
TS 1.2, pg. 4 of 43
Section C: Facility and Radiation Monitoring Systems Page 28 C.01 Answer:
a REF:
SOP A.2 C.02 Answer:
d REF:
SAR Table 4-1, pg. 4-7 C.03 Answer:
a REF:
SAR 13.2.1, pg. 13-2 C.04 Answer:
c REF:
UFTR Training Material, Design and Operating Characteristics,Section I.C.5.b.,
pg. 4 C.05 Answer:
b REF:
SAR Table 4-1, pg. 4-7 UFTR Training Material, Design and Operating Characteristics,Section I.A.1.b.,
pg. 2 C.06 Answer:
a REF:
SAR 7.1.3.1.1, pg. 7-1 C.07 Answer:
d REF:
UFTR Training Material, Instrumentation and Control,Section I.C.1.c., pg.5 C.08 Answer:
a REF:
UFTR Training Material, Reactor Protection System,Section II.B.2.e., pg.3 C.09 Answer:
c REF:
SAR 7.2.1, pg. 7-4 C.10 Answer:
a REF:
SAR 5.2, pg. 5-1 C.11 Answer:
a,5 b,2 c,3 d,1 REF:
UFTR Training Material, Design and Operating Characteristics, Figure 5, pg. 35 C.12 Answer:
d REF:
2_Blade_Worth_Curves per UFTR FORM SOP-A.7B
Section C: Facility and Radiation Monitoring Systems Page 29 C.13 Answer:
c REF:
SAR 7.6, pg. 7-6 C.14 Answer:
a REF:
SAR 7.2.2, pg. 7-4 UFTR Training Material, Instrumentation and Control,Section I.B.1.d.2., pg. 4 C.15 Answer:
b REF:
SAR 12.1.2, pg. 12-1 SOP-A.2, Section 4.1, pg. 2 of 7 C.16 Answer:
d REF:
SAR Table 11-4, pg. 11-6 C.17 Answer:
b REF:
SAR 5.2, pg. 5-1 C.18 Answer:
b REF:
SAR 4.2.1.2, pg. 4-9 TS 5.3.3, pg. 33 of 43 C.19 Answer:
a REF:
SAR 8.1.2, pg. 8-1 SAR 7.6, pg. 7-6 C.20 Answer:
a REF:
TS 5.4, pg. 33 of 43