ML070110112
| ML070110112 | |
| Person / Time | |
|---|---|
| Site: | Purdue University |
| Issue date: | 01/18/2007 |
| From: | Doyle P NRC/NRR/ADRA/DPR/PRTB |
| To: | Joel Jenkins Purdue University Research Reactor |
| Doyle P, NRC/NRR/DPR/PRT, 415-1058 | |
| Shared Package | |
| ML 062140365 | List: |
| References | |
| 50-182/OL-70-01 | |
| Download: ML070110112 (28) | |
Text
January 18, 2007 Mr. Jere Jenkins, Laboratory Director 1290 Nuclear Engineering Bldg.
Department of Nuclear Engineering Purdue University West Lafayette, IN 47907
SUBJECT:
INITIAL EXAMINATION REPORT NO. 50-182/OL-07-01, PURDUE UNIVERSITY
Dear Mr. Jenkins:
During the week of November 6, 2006, the NRC administered an operator licensing examination at your Purdue University Reactor. The examination was conducted according to NUREG-1478, "Non-Power Reactor Operator Licensing Examiner Standards," Revision 1.
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 10 CFR 2.390 of the Commission's regulations, 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 (the Public Electronic Reading Room) http://www.nrc.gov/reading-rm/adams.html.
The NRC is forwarding the individual grades to you in a separate letter which will not be released publicly. Should you have any questions concerning this examination, please contact Mr. Paul V. Doyle Jr. at (301) 415-1058 or via internet e-mail pvd@nrc.gov.
Sincerely,
/Ra/
Johnny Eads, Chief Research and Test Reactors Branch B Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Docket No. 50-182
Enclosures:
- 1. Initial Examination Report No. 50-182/OL-07-01
- 2. Examination with facility comments with NRC resolution cc w/encls:
Please see next page
Mr. Jere Jenkins, Laboratory Director January 18, 2007 1290 Nuclear Engineering Bldg.
Department of Nuclear Engineering Purdue University West Lafayette, IN 47907
SUBJECT:
INITIAL EXAMINATION REPORT NO. 50-182/OL-07-01, PURDUE UNIVERSITY
Dear Mr. Jenkins:
During the week of November 6, 2006, the NRC administered an operator licensing examination at your Purdue University Reactor. The examination was conducted according to NUREG-1478, "Non-Power Reactor Operator Licensing Examiner Standards," Revision 1.
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 10 CFR 2.390 of the Commission's regulations, 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 (the Public Electronic Reading Room) http://www.nrc.gov/reading-rm/adams.html.
The NRC is forwarding the individual grades to you in a separate letter which will not be released publicly. Should you have any questions concerning this examination, please contact Mr. Paul V. Doyle Jr. at (301) 415-1058 or via internet e-mail pvd@nrc.gov.
Sincerely,
/RA/
Johnny Eads, Chief Research and Test Reactors Branch B Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Docket No. 50-182
Enclosures:
- 1. Initial Examination Report No. 50-182/OL-07-01
- 2. Examination with facility comments with NRC resolution cc w/encls:
Please see next page DISTRIBUTION w/ encls.:
PUBLIC PRTB r/f JEads DHughes Facility File (EBarnhill) O-6 F-2 ADAMS ACCESSION NO.: Package ML062140365 ADAMS ACCESSION #: ML070110112 TEMPLATE #:NRR-074 OFFICE PRTB:CE IOLB:LA E
PRTB:SC NAME PDoyle:cah EBarnhill JEads DATE 01 /16 /2007 01/18/2007 01/18/2007 OFFICIAL RECORD COPY
Purdue University Docket No. 50-182 cc:
Mayor City of West Lafayette 609 W. Navajo West Lafayette, IN 47906 John H. Ruyack, Manager Epidemiology Res Center/Indoor & Radiological Health Indiana Department of Health 2525 N. Shadeland Ave., E3 Indianapolis, IN 46219 Howard W. Cundiff, P.E., Director Consumer Protection Indiana State Department of Health 2 North Meridian Street, 5D Indianapolis, IN 46204 Mr. Ed Merritt Reactor Supervisor Department of Nuclear Engineering Purdue University West Lafayette, IN 47907 Test, Research, and Training Reactor Newsletter University of Florida 202 Nuclear Sciences Center Gainesville, FL 32611
U. S. NUCLEAR REGULATORY COMMISSION OPERATOR LICENSING INITIAL EXAMINATION REPORT REPORT NO.:
50-182/OL-07-01 FACILITY DOCKET NO.:
50-182 FACILITY LICENSE NO.:
R-87 FACILITY:
Purdue University EXAMINATION DATES:
November 07, 2006 SUBMITTED BY:
___/RA/________________________
01/10/2007 Paul V. Doyle Jr., Chief Examiner Date
SUMMARY
On November 7, 2007, the NRC administered an Operator Licensing examination to one SRO (Instant) candidate. The candidate passed all portions of the administered examination.
REPORT DETAILS 1.
Examiners:
Paul V. Doyle Jr., Chief Examiner 2.
Results:
RO PASS/FAIL SRO PASS/FAIL TOTAL PASS/FAIL Written 0/0 1/0 1/0 Operating Tests 0/0 1/0 1/0 Overall 0/0 1/0 1/0 3.
Exit Meeting:
Paul V. Doyle Jr., NRC, Chief Examiner Jere Jenkins, Purdue University, Laboratory Director Edward Merritt, Purdue University, Reactor Supervisor The examiner thanked the facility for their support of the examination. The facility had minor comments on the examination (e.g. two correct answers) the examination included with this report has all comments included.
ENCLOSURE 1
OPERATOR LICENSING EXAMINATION With Answer Key PURDUE UNIVERSITY November 07, 2006 ENCLOSURE 2
Section A: L Theory, Thermodynamics & Facility Operating Characteristics Page 1 QUESTION A.01
[1.0 point]
Which ONE of the following describes the characteristics of good moderators and reflectors?
a.
High scattering cross-section and low absorption cross-section.
b.
Low scattering cross-section and high absorption cross-section.
c.
Low scattering cross-section and low absorption cross-section.
d.
High scattering cross-section and high absorption cross-section QUESTION A.02
[1.0 point]
You enter the control room and note that all nuclear instrumentation show a steady neutron level, and no rods are in motion. Which ONE of the following conditions CANNOT be true?
a.
The reactor is critical.
b.
The reactor is subcritical.
c.
The reactor is supercritical.
d.
The neutron source has been removed from the core..
QUESTION A.03
[2.0 points, 1/2 each]
A fissile material is one which will fission upon the absorption of a THERMAL neutron. A fertile material is one which upon absorption of a neutron becomes a fissile material. Identify each of the listed isotopes as either fissile or fertile.
a.
Th232 b.
U233 c.
U235 d.
Pu239 QUESTION A.04
[1.0 point]
A reactor is xenon free, with no experiments in the core. Given the following reactivity worths, calculate the Shutdown Margin.
worth (%K/K) worth (%K/K)
Shim-Safety Blade #1:
2.41 Shim-Safety Blade #2:
2.32 Shim-Safety Blade #3:
2.49 Shim-Safety Blade #4:
2.60 Regulating rod:
0.084 Excess Reactivity:
3.42 a.
9.90%
b.
6.48%
c 6.40%
d.
3.80%
Section A: L Theory, Thermodynamics & Facility Operating Characteristics Page 2 QUESTION A.05
[1.0 point]
The reactor is operating at 100 KW. The reactor operator withdraws the Regulating Rod allowing power to increase.
The operator then inserts the same rod to its original position, decreasing power. In comparison to the rod withdrawal, the period due to the rod insertion will be a.
longer due to long lived delayed neutron precursors.
b.
shorter due to long lived delayed neutron precursors.
c.
same due to equal amounts of reactivity being added.
d.
same due to equal reactivity rates from the rod.
QUESTION A.06
[1.0 point]
Which of the following power manipulations would take the longest to complete assuming the same period is maintained?
a.
1 Kilowatt: from 1 kW to 2 kW b.
1.5 Kilowatts
from 2 kW to 3.5 kW c.
2 Kilowatts: from 3.5 kW to 5.5 kW d.
2.5 Kilowatts
from 5.5 kW to 8 kW QUESTION A.07
[1.0 point]
Which ONE of the following will cause the reactor to reach criticality at a higher rod height?
a.
Increase in moderator temperature b.
Increase in barometric pressure c.
adding new fuel d.
inserting a positive reactivity experiment QUESTION A.08
[1.0 point]
Suppose the source strength in the core is 250 neutrons per second (N/sec) and the effective multiplication factor is 0.80. Select the closest stable neutron count rate from the list below:
a.
313 N/sec b.
750 N/sec c.
1250 N/sec d.
1500 N/sec
Section A: L Theory, Thermodynamics & Facility Operating Characteristics Page 3 QUESTION A.09
[1.0 point]
What is the approximate amount of time that it will take the amount of Xenon being produced to reach a peak after the reactor is shut down? On the attached Xenon reactivity curve it is noted as the difference between time TSD and TPeak.
a.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> b.
15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> c.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> d.
33 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br /> QUESTION A.10
[1.0 point]
During a fuel loading of the core, as the reactor approaches criticality, the value of 1/M:
a.
Increases toward one b.
Decreases toward one c.
Increases toward infinity d.
Decreases toward zero QUESTION A.11
[1.0 point]
The reactor supervisor tells you that the Keff for the reactor is 0.955. How much reactivity must you add to the reactor to reach criticality?
a.
+0.0471 b.
+0.0450 c.
-0.0471 d.
-0.0450 QUESTION A.12
[2.0 points, 1/2 each]
Match each term in column A with the correct definition in column B.
Column A Column B a.
Prompt Neutron 1.
A neutron in equilibrium with its surroundings.
b.
Fast Neutron 2.
A neutron born directly from fission.
c.
Thermal Neutron 3.
A neutron born due to decay of a fission product.
d.
Delayed Neutron 4.
A neutron at an energy level greater than its surroundings.
Section A: L Theory, Thermodynamics & Facility Operating Characteristics Page 4 QUESTION A.13
[1.0 point]
and eff both describe the total fraction of delayed neutrons. The difference between the two is that eff is a.
smaller than since delayed neutrons are born at lower energy levels than prompt neutrons.
b.
larger than since delayed neutrons are born at lower energy levels than prompt neutrons.
c.
smaller than since delayed neutrons are born at higher energy levels than prompt neutrons.
d.
larger than since delayed neutrons are born at higher energy levels than prompt neutrons.
QUESTION A.14
[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.
QUESTION A.15
[1.0 point]
Which ONE of the following atoms will cause a neutron to lose the most energy in an ELASTIC collision?
a.
Uranium (U238) b.
Carbon (C12) c.
Deuterium (H2) d.
Hydrogen (H1)
QUESTION A.16
[1.0 point]
A thin foil target of 10% copper and 90% aluminum is in a thermal neutron beam. Given a Cu = 3.79 barns, a Al = 0.23 barns, s Cu = 7.90 barns, and s Al =1.49 barns, which ONE of the following reactions has the highest probability of occurring? A neutron a.
scattering reaction with aluminum b.
scattering reaction with copper c.
absorption in aluminum d.
absorption in copper
Section B Normal, Emergency and Radiological Control Procedures Page 5 QUESTION B.01
[2.0 points, 1/2 each]
Identify each of the following actions as either a channel CHECK, a channel TEST, or a channel CALibration.
a.
Prior to startup you place a radioactive source near a radiation detector, noting meter movement and alarm function operation.
b.
During startup you compare all of your nuclear instrumentation channels ensuring they track together.
c.
At power, you perform a heat balance (calorimetric) and determine you must adjust Nuclear Instrumentation readings.
d.
During a reactor shutdown you note a -80 second period on Nuclear Instrumentation.
QUESTION B.02
[1.0 point]
When performing a reactor startup, to avoid a reactor setback or scram turn the range switch for channel 3 COUNTER-CLOCKWISE when the console meter reads ? and CLOCKWISE when the console meter reads ?.
COUNTER-CLOCKWISE CLOCKWISE a.
above 75%
between 20 and 25%
b.
between 70 and 75%
below 20%
c.
above 65%
between 20 and 25%
d.
between 65 and 70%
below 20%
QUESTION B.03
[1.0 point]
Identify the condition which is a reportable occurrence.
a.
A shutdown margin of 1.5% K/K.
b.
Disagreement between actual and expected critical positions of more than 0.25% K/K.
c.
An unexpected short-term reactivity change causing a period of 9 seconds.
d.
Discovering the setback setpoint for the linear channel was set at 105% power.
d.
Transporting corrosive material beyond the chains of either pool.
QUESTION B.04
[1.0 point]
A survey instrument with a window probe was used to measure an irradiated experiment. The results were 100 millirem/hr window open and 60 millirem/hr window closed. What was the gamma dose?
a.
100 millirem/hr b.
60 millirem/hr c.
40 millirem/hr d.
140 millirem/hr
Section B Normal, Emergency and Radiological Control Procedures Page 6 QUESTION B.05
[1.0 point]
Which ONE of the following is the definition of Emergency Action Level?
a.
a condition that calls for immediate action, beyond the scope of normal operating procedures, to avoid an accident or to mitigate the consequences of one.
b.
Specific instrument readings, or observations; radiation dose or dose rates; or specific contamination levels of airborne, waterborne, or surface-deposited radioactive materials that may be used as thresholds for establishing emergency classes and initiating appropriate emergency methods.
c.
classes of accidents grouped by severity level for which predetermined emergency measures should be taken or considered.
d.
a document that provides the basis for actions to cope with an emergency. It outlines the objectives to be met by the emergency procedures and defines the authority and responsibilities to achieve such objectives.
QUESTION B.06
[2.0 points, 1/2 each]
Match the 10CFR55 requirements for maintaining an active operator license in column A with the corresponding time period from column B.
Column A Column B a.
Renew License 1 year b.
Medical Exam 2 years c.
Pass Requalification Written Examination 4 years d.
Pass Requalification Operating Test 6 years QUESTION B.07
[1.0 point]
Which ONE of the following is the SAFETY LIMIT as set by Technical Specifications?
a.
Reactor Power TRUE VALUE shall not exceed 50 Kilowatts.
b.
Reactor Power MEASURED VALUE shall not exceed 50 Kilowatts.
c.
Reactor Power TRUE VALUE shall not exceed 1.2 Kilowatts.
d.
Reactor Power MEASURED VALUE shall not exceed 1.2 Kilowatts QUESTION B.08
[1.0 point]
Two isotopes are monitored for (Regions of Interest) during the Reactor Water Gamma Spec Analysis Standard Operating Procedure. This procedure is used to detect fission products? Which ONE of the following choices shows the two (2) radioisotopes monitored?
a.
1H3 and 7N16 b.
6C14 and 8O16 c.
11Al24 and 18Ar41 d.
53I137 and 55Cs137
Section B Normal, Emergency and Radiological Control Procedures Page 7 QUESTION B.09
[1.0 point]
There are two SROs and one non-licensed person on shift when one of the SROs becomes sick. Which ONE of the listed situations is allowed by Technical Specifications?
a.
The non-licensed person takes over the control console. The SRO remains within the confines of the annex basement.
b.
The SRO takes over the console. The non-licensed person remains within the confines of the Purdue campus.
c.
The non-licensed person takes over the control console. The SRO remains at the console to direct the non-licensed person.
d.
The SRO takes over the control console. The non-licensed person is free to leave.
QUESTION B.10
[2.0 points, 1/2 each]
Identify whether each of the following conditions requires (REQ) or does not require (NOT) SRO presence for recovery from an unplanned (unscheduled) shutdown.
a.
Building evacuation due to a fire drill initiated by the school administration.
b.
High school student on tour accidentally hits scram button.
c.
Reactor scram due to trainee generating a 6 second period d.
Loss of electrical power due to electrical storms.
QUESTION B.11
[1.0 point]
All of the following conditions are REPORTABLE OCCURRENCES except a.
dropping material into the shutdown core, which causes an unplanned criticality.
b.
A core excess reactivity of 0.025 K/K.
c.
An unexpected short-term reactivity change causing a period of 9 seconds d.
Discovering the scram setpoint for the safety channel was set at 125% power.
QUESTION B.12
[1.0 point]
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
Section B Normal, Emergency and Radiological Control Procedures Page 8 QUESTION B.13
[1.0 point]
An experimenter wants to perform an experiment. The experiment has approximately the same nuclear characteristics as an experiment previously tried. Which ONE of the following is the minimum level of approval necessary to perform the experiment? The Radiation Management Office and a.
a licensed Reactor Operator.
b.
a licensed Senior Reactor Operator.
c.
the Reactor Supervisor.
d.
the Reactor Supervisor with the approval of the Committee of Reactor Operations.
QUESTION B.14
[1.0 point]
You (a licensed Reactor Operator) and a Senior Reactor Operator (SRO) are operating the reactor on the weekend.
No one else is available. In order to meet Technical Specifications requirements if you are on the console the SRO must be a.
within the reactor control room b.
within the Reactor Building c.
within the confines of the Campus d.
within 30 minutes drive of the Reactor Facility.
QUESTION B.15
[1.0 point]
Technical Specification 5.3 requires All reactor fuel assemblies shall be stored in a geometrical array where the Keff is less than ____ for all conditions of moderation and reflection.
- a. 0.80
- b. 0.85 c.
0.90
- d. 0.95
QUESTION C.01
[2.0 points, 0.4 each]
Match the Nuclear Instrumentation channels with their correct functions. (Protection refers to scrams and setbacks, while control relates to input to the automatic circuit for the servo control system.)
Column A Column B a.
Startup Channel 1.
Setback and Slow Scram b.
Log-N Period Channel 2.
Setback and Fast Scram c.
Pool Top Radiation Area Monitor 3.
Slow Scram Only d.
Safety Channel 4.
None e.
Continuous Air Monitor QUESTION C.02
[1.0 point]
Which ONE of the following components within the purification system is responsible for maintaining primary conductivity low?
a.
conductivity cells b.
water filter c.
chiller unit d.
demineralizer QUESTION C.03
[1.0 point]
Which one of the following actions are the result of a rod insert (set back) signal?
a.
The two shim-safety rods drive downward into the core until the initiating signal is less than the setpoint.
b.
All three rods drive downward into the core until the initiating signal is less than the setpoint.
c.
The two drive downward into the core until the set back signal is reset manually.
d.
All three drive downward into the core until the set back signal is reset manually.
QUESTION C.04
[2.0 points, a each]
Identify whether each of the following traits is associated with the Shim-Safety Rods (SSR) or the Regulating Rod (RR).
a.
Hollow b.
Solid c.
made of Stainless-Steel d.
made of borated stainless-steel e.
magnetic clutch f.
direct drive
Section C Facility and Radiation Monitoring Systems Page 10 QUESTION C.05
[1.0 point]
A visitor inadvertently hits the gang-lower switch. How may the operator stop the rods from inserting?
a.
Reenergizing the Log Count Rate Channel.
b.
Ranging UP on the Linear Level Channel.
c.
Momentarily placing the Raise-Lower switch in either the raise or lower position.
d.
Not required. Rod insertion will stop when the gang lower switch is released.
QUESTION C.06
[1.0 point]
What is the purpose of the Startup channel drive mechanism?
a.
To maintain the instrument on-scale.
b.
To prevent burnout of the U235 coating on the detector during high power.
c.
To maintain coupling between the fission chamber and the neutron source as they are both withdrawn from the core.
d.
To prevent shorting out the preamplifier due to detector saturation.
QUESTION C.7
[1.0 point]
Maximum design flow rate for the Ion Exchanger (Demineralizer) is 20 gpm, but the design flow rate of the pump is 30 gallons per minute. How is the system setup to prevent too much flow to the Demineralizer?
a.
A pipe with an orifice bypasses some flow around the Demineralizer.
b.
A throttle valve in the system is adjusted to supply proper flow.
c.
Cavitation due to inadequate Net Positive Suction Head (NPSH) is used to derate the pump flow rate.
d.
Differential pressure across the cartridge type filters is kept high to reduce flow.
QUESTION C.8
[1.0 point]
What is the purpose of the Dash Pot Cylinder in the Shim-Safety Control Rod Assembly?
a.
To provide indication of the control rod at the top of the core.
b.
To provide indication of the control rod at the bottom of the core.
c.
To minimize reactivity addition during a rod withdrawal accident.
d.
To minimize the mechanical stress of a rod drop.
Section C Facility and Radiation Monitoring Systems Page 11 QUESTION C.9
[1.0 point]
Which ONE of the following correctly describes how gamma radiation is compensated for in the Log-N channel?
a.
A compensating current equal and opposite to the signal due to gammas is generated by the detector.
b.
The detector is positioned in towards and out away from the core to compensate for gammas.
c.
The output of the detector is put through a discriminator circuit which passes only pulses caused by neutron interactions.
d.
Lead shielding around the detector decreases the signal due to gammas low enough such that compensation is not required.
QUESTION C.10
[1.0 point]
You just finished working on the Nuclear Instrumentation rate (period) channel. Tests are completed and the reactor supervisor want you to verify channel operation. You perform a startup and you establish a period of 100 seconds-1. Which one of the listed approximate times to double reactor power would indicate the instrument is working properly?
a.
30 seconds b.
50 seconds c.
70 seconds d.
100 seconds QUESTION C.11
[2.0 points, 1/2 each]
Match the radiation detection equipment in column A, with its primary use in column B.
Column A: Radiation Equipment Column B: Primary use a.
Ion Chamber portable radiation detector 1.
To measure total cumulative dose received by a visitor.
b.
Geiger-Müller portable radiation detector 2.
To detect the presence of contamination.
c.
Film Badge/TLD 3.
To measure radiation field strength.
d.
Pocket Dosimeter.
4.
To measure total cumulative dose received by a radiation worker.
QUESTION C.12
[1.0 point]
Which ONE of the following is the detector type used in the Radiation Area Monitors?
a.
Scintillation Detector b.
Ion Chamber c.
Geiger Muller d.
Proportional Counter
Section C Facility and Radiation Monitoring Systems Page 12 QUESTION C.13
[1.0 point]
Which ONE of the following conditions will NOT prohibit withdrawal of the shim-safety rods?
a.
Reactor period at 12 seconds.
b.
Linear channel reading 112% range.
c.
Startup channel Count Rate at 1 cps.
d.
Log Count Rate Selector in USE position.
QUESTION C.14
[2.0 points, a each]
Using the figure provided, identified each of shim-safety rod micro switches.
a.
Engage b.
Jam c.
Lower Limit d,
Rod Lower Limit e.
Shim Range f.
Upper Limit
Section A: L Theory, Thermodynamics & Facility Operating Characteristics Page 13 A.01 a
REF:
x A.02 c
REF:
A.03 a, fertile; b, fissile; c, fissile; d, fissile REF:
A.04 d
REF:
SDM (cold/clean) = Total Rod worth - Kexcess - Most reactive blade - Reg Rod SDM = (2.41 + 2.32 + 2.49 + 2.60 + 0.084) - 3.42 - 2.60 - 0.084 = 3.80%
A.05 a
REF:
A.06 a
REF:
P = P0 et/T A.07 a
REF:
A.8 c
REF:
Count = S/(1 - Keff) = 250/(1 - 0.8) = 250/0.2 = 1250 A.09 b or a (2nd correct answer added per facility comment)
REF:
A.10 d
REF:
A.11 a
REF:
= (Keff1 - Keff2) ÷ (Keff1
- Keff2)
= (0.9550 - 1.0000) ÷ (0.9550
- 1.0000)
= -0.0450 ÷ 0.9550 = -0.0471 A.12 a, 2; b, 4; c, 1; d, 3 REF:
A.13 b
REF:
x A.14 a
REF:
Burn, R., Introduction to Nuclear Reactor Operations, © 1988, § p.7-5 A.15 d
REF:
Burn, R., Introduction to Nuclear Reactor Operations, © 1988, § 2.5.3 A.16 a
REF:
Section B Normal, Emergency and Radiological Control Procedures Page 14 B.01 a, Check; b, Check; c, Cal; d, Test REF:
Technical Specifications § 1.0 Definitions B.02 b or d (2nd correct answer added per facility comment)
REF:
PUR Procedure 91-1, Reactor Startup, Operation and Shutdown. a, 20; b, 1; c, 1; d, 10 B.03 c
REF:
Technical Specifications 1.27, 3.1.a, 3.2 Table 1.
B.04 b
REF:
Standard NRC Health Physics Question B.05 b
REF:
Emergency Plan, § 2.0 Definitions, p. 6.
B.06 a, 6; b, 2; c, 2; d, 1 REF:
B.07 a
REF:
Technical Specification 2.1 B.08 d
REF:
SOP Reactor Water Sampling and Analysis (Revised Jan. 2005)
B.09 c
REF:
PUR Technical Specifications, §§ 6.1.6 through 6.1.11.
B.10 a, NOT; b, NOT; c, REQ; d, NOT REF:
Technical Specification 6.1.9.
B.11 d
REF:
Technical Specifications §§ 3.2 & 1.27 B.12 a
REF:
B.13 c
REF:
TS § 1.31 Tried Experiment, TS § 1.51 New Experiment, TS § 6.2.5(c), Form B.
B.14 b
REF:
TS § 1.25 B.15 a
REF:
Technical Specification 5.3(1)
Section C Facility and Radiation Monitoring Systems Page 15 C.01 a, 1; b, 2; c, 3; d, 2; e, 4 REF:
Instrumentation Block Diagram C.02 d
REF:
SAR §3.6 C.03 d
REF:
SAR § 3.7.1 C.04 a, RR; b, SSR; c, RR; d, SSR; e, SSR; f, RR REF:
SAR § 1.2 C.05 c
REF:
PUR-1 Operating Manual 1962, Lockheed Nuclear Products, § 1.5.6.5.
C.06 a or b (2nd correct answer added per facility comment)
REF:
PUR-1 Operation Manual 1962, Lockheed Nuclear Products, § 1.5.6.3 Fission Chamber Drive System C.07 b
REF:
PUR-1 Operations Manual 1962, Lockheed Nuclear Products, § 1.6 Process System C.8 d
REF:
SAR § 3.1.2.3, p. 33; OM-15 Figure from Attachment C.
C.9 a
REF:
Standard NRC Question C.10 c
REF:
x time = Ln(2) x 100 = 69.3 sec. (Choice a. represents using Log vs. Ln.)
C.11 a, 3; b, 2; c, 4; d, 1 REF:
Standard NRC question.
C.12 a
REF:
PUR-1 Operating Procedure M-5, Procedure for Radiation Area Monitor, figure 1, also SAR § 3.7.7.2 C.13 d
REF:
PURDUE Requalification Examination 12/78, question E C.14 a, 6; b, 1; c, 4; d, 5; e, 3; f, 2 REF:
Facility supplied Figure (Shim Safety Rod)
U. S. NUCLEAR REGULATORY COMMISSION RESEARCH AND TEST REACTOR INITIAL OPERATOR LICENSING EXAMINATION FACILITY:
Purdue University REACTOR TYPE:
PUR (pool)
DATE ADMINISTERED: 2006/11/08 CANDIDATE:
INSTRUCTIONS TO CANDIDATE:
Circle correct answer on the answer sheet provided. If you wish to change your answer, cross out the incorrect answer and write the correct answer on the line provided to the right. If you change you wish to change your answer again, cross the incorrect answer and write the correct answer to the right of the last answer you put down. Points for each question are indicated in brackets for each question. A 70%
in each section is required to pass the examination. The examiner will pick up these handouts three (3) hours after the examination starts.
% of Category % of Candidates Category Value Total Score Value Category 18.00 33.3 A.
Reactor Theory, Thermodynamics and Facility Operating Characteristics 18.00 33.3 B.
Normal and Emergency Operating Procedures and Radiological Controls 18.00 33.3 C.
Facility and Radiation Monitoring Systems 54.00 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 OPERATOR LICENSING EXAMINATIONS The following rules apply during the administration of this examination:
1.
Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
2.
After you complete the examination you must sign the statement on the cover sheet of the handout indicating that the work is your own and you have neither received nor given assistance in completing the examination.
3.
Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
4.
Use black ink or dark pencil only to facilitate legible reproductions.
5.
Print your name in the blank provided in the upper right-hand corner of the examination cover sheet.
6.
Mark your answers by drawing a circle around the correct answer on the answer sheets provided.
If you wish to change your answer cross out the previous answer and write the correct answer on the underlined space 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 number.
8.
If the intent of a question is unclear, ask questions of the examiner only.
9.
When finished turn in this handout, with the cover sheet signed.
10.
Ensure all information you wish to have evaluated as part of your answer is on your answer sheet.
Scrap paper will be disposed of immediately following the examination.
11.
To pass the examination you must achieve a grade of 70 percent or greater in each category.
12.
There is a time limit of three (3) hours for completion of the examination.
13.
When you have completed and turned in you examination, leave the examination area. If you are observed in this area while the examination is still in progress, your license may be denied or revoked.
EQUATION SHEET 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 EF = 9/5 EC + 32 1 gal (H2O). 8 lbm EC = 5/9 (EF - 32) cP = 1.0 BTU/hr/lbm/EF cp = 1 cal/sec/gm/EC
Section A L Theory, Thermo, and Facility Characteristics Page 19 A.01 a b c d ___
A.09 a b c d ___
A.02 a b c d ___
A.10 a b c d ___
A.03a fertile fissile ___
A.11 a b c d ___
A.03b fertile fissile ___
A.12a 1 2 3 4 ___
A.03c fertile fissile ___
A.12b 1 2 3 4 ___
A.03d fertile fissile ___
A.12c 1 2 3 4 ___
A.04 a b c d ___
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Section B Normal/Emerg. Procedures & Rad Con Page 20 B.01a Check Test Cal ___
B.07 a b c d ___
B.01b Check Test Cal ___
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B.01c Check Test Cal ___
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B.01d Check Test Cal ___
B.10a REQ NOT ___
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Section C Facility and Radiation Monitoring Systems Page 21 C.01a 1 2 3 4 ___
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C.01b 1 2 3 4 ___
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C.01c 1 2 3 4 ___
C.11a 1 2 3 4 ___
C.01d 1 2 3 4 ___
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C.04a RR SSR ___
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C.14a 1 2 3 4 5 6 ___
C.04d RR SSR ___
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C.14f 1 2 3 4 5 6 ___