ML021350400
| ML021350400 | |
| Person / Time | |
|---|---|
| Site: | Oregon State University |
| Issue date: | 06/05/2002 |
| From: | Madden P NRC/NRR/DRIP/RORP |
| To: | Binney S Oregon State University |
| Doyle, P, NRR/DRIP/RORP 415-1058 | |
| References | |
| 50-243/OL-02-01 50-243/OL-02-01 | |
| Download: ML021350400 (35) | |
Text
June 5, 2002 Dr. Steven E. Binney, Director Oregon State University Radiation Center, A100 Corvallis, OR 97331-5903
SUBJECT:
INITIAL EXAMINATION REPORT NO. 50-243/OL-02-01, OREGON STATE UNIVERSITY
Dear Dr. Binney:
On April 28, 2002, the NRC administered an operator licensing examination at your Oregon State University Reactor. The examination was conducted according to NUREG-1478, "Non-Power Reactor Operator Licensing Examiner Standards," Draft 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 10 CFR 2.790 of the Commissions 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 NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at (the Public Electronic Reading Room) http://www.nrc.gov/NRC/ADAMS/indesx.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 Paul Doyle at (301)415-1058 or pvd@nrc.gov.
Sincerely,
/RA/
Patrick M. Madden, Section Chief Research and Test Reactors Section Operating Reactor Improvements Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation
Docket No. 50-243
Enclosures:
- 1. Initial Examination Report No. 50-243/OL-02-01
- 2. Examination and answer key (RO/SRO) cc w/encls:
Please see next page
Oregon State University Docket No. 50-243 cc:
Mayor of the City of Corvallis Corvallis, OR 97331 David Stewart-Smith Oregon Office of Energy 625 Marion Street, N.E.
Salem, OR 97310 George Holdren, Interim Vice Provost for Research Oregon State University Administrative Services Bldg., Room A-312 Corvallis, OR 97331-5904 Dr. Steven Reese Reactor Administrator Oregon State University Radiation Center, A-100 Corvallis, OR 97331-5904 Dr. Jack F. Higginbotham, Chairman Reactor Operations Committee Oregon State University Radiation Center, A-100 Corvallis, OR 97331-5904 Test, Research, and Training Reactor Newsletter University of Florida 202 Nuclear Sciences Center Gainesville, FL 32611
June 5, 2002 Dr. Steven E. Binney, Director Oregon State University Radiation Center, A100 Corvallis, OR 97331-5903
SUBJECT:
INITIAL EXAMINATION REPORT NO. 50-243/OL-02-01, OREGON STATE UNIVERSITY
Dear Dr. Binney:
On April 28, 2002, the NRC administered an operator licensing examination at your Oregon State University Reactor. The examination was conducted according to NUREG-1478, "Non-Power Reactor Operator Licensing Examiner Standards," Draft 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 10 CFR 2.790 of the Commissions 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 NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at (the Public Electronic Reading Room) http://www.nrc.gov/NRC/ADAMS/indesx.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 Paul Doyle at (301) 415-1058 or pvd@nrc.gov.
Sincerely,
/RA/
Patrick M. Madden, Section Chief Research and Test Reactors Section Operating Reactor Improvements Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Docket No. 50-243
Enclosures:
- 1. Initial Examination Report No. 50-243/OL-02-01
- 2. Examination and answer key (RO/SRO) cc w/encls:
Please see next page
DISTRIBUTION w/ encls.:
PUBLIC RORP/R&TR r/f PMadden AAdams Facility File (EBarnhill) O6-D17 ADAMS ACCESSION #: ML021350400 TEMPLATE #:NRR-074 OFFICE RORP:CE IEHB:LA E
RORP:SC NAME PDoyle:rdr EBarnhill PMadden DATE 05/ 25 /2002 05/ 24 /2002 05/ 28 /2002 C = COVER E = COVER & ENCLOSURE N = NO COPY OFFICIAL RECORD COPY
ENCLOSURE 1 EXAMINATION REPORT NO: 50-243/OL/02-01 Oregon State University TRIGA Reactor Facility Docket No.:
50-243 Facility License No.:
R-160 SUBMITTED BY:
/RA/
__05/25/02__
Paul V. Doyle Jr., Chief Examiner Date Summary:
On April 29, 2002, the NRC administered a licensing examination to one SRO (Instant) license applicant. The applicant passed the examination.
Additional Examiners:
None Exit Meeting:
Attendees:
Gary Wachts, Reactor Supervisor, Oregon State Triga Reactor (OSTR)
Steven Reese, Reactor Administrator, OSTR Paul Doyle, Examiner, NRC At the conclusion of the site visit, the examiner met with representatives of the facility staff to discuss the results of the examinations.
The examiners made the following observations concerning your training program:
a.
The examiner did not note any areas of generic weakness during the administration of the operating tests.
ENCLOSURE 1
OREGON STATE UNIVERSITY With Answer Key
OPERATOR LICENSING EXAMINATION April 29, 2002
QUESTION A.1 [1.0 point]
Which ONE of the following isotopes has the largest microscopic cross-section for absorption for thermal neutrons?
- a. Sm149
- b. U235 c.
Xe135
- d. B10 QUESTION A.2 [1.0 point]
Which ONE of the following factors is the most significant in determining the differential worth of a control rod?
- a. The rod speed.
- b. Reactor power.
c.
The flux shape.
- d. The amount of fuel in the core.
QUESTION A.3 [1.0 point]
A reactor (not OSTR) has the following reactivity characteristics.
Kexcess... $2.50 Standard Rod 1.... $2.25 Standard Rod 2.... $2.30 Reg Rod.... $1.10 Which ONE of the following is the shutdown margin allowable by Technical Specifications. (NOTE: All rods are able to scram, same condition as OSTR Tech Spec.)
- a. $5.65
- b. $4.00 c.
$3.10
- d. $0.85
QUESTION A.4 [1.0 point]
Which ONE of the following conditions describes a critical reactor?
- a. Keff = 1; k/k ( ) = 1
- b. Keff = 1; k/k ( ) = 0 c.
Keff = 0; k/k ( ) = 1
- d. Keff = 0; k/k ( ) = 0 QUESTION A.5 [1.0 point]
Which ONE of the following is an example of beta decay?
a.
35Br87 33As83 b.
35Br87 35Br86 c.
35Br87 34Se86 d.
35Br87 36Kr87 QUESTION A.6 [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
QUESTION A.7 [1.0 point]
You are increasing reactor power on a steady +26 second period. How long will it take to increase power by a factor of 1000?
- a. 60 seconds (1 minute)
- b. 180 seconds (3 minutes) c.
300 seconds (5 minutes)
- d. 480 seconds (8 minutes)
QUESTION A.8 [1.0 point]
Which ONE of the following statements is the definition of REACTIVITY?
- a. A measure of the cores fuel depletion.
- b. A measure of the cores deviation from criticality.
c.
Equal to 1.00 K/K when the reactor is critical.
- d. Equal to 1.00 K/K when the reactor is prompt critical.
QUESTION A.9 [1.0 point]
Which ONE of the following correctly describes the generation of neutrons from the Am-Be source?
a.
95Am241 93Np237 + 2 4;
2 4 + 4Be9
[6C13]* 6C12 + 0n1 b.
95Am241 96Np241 + -1 0 + ;
0 0 + 4Be9
[4Be9]* 4Be8 + 0n1 c.
95Am241 96Np241 + -1 0 + ;
-1 0 + 4Be9
[3Li9]* 3Li8 + 0n1 d.
95Am241
[S.F.]
2 fission products + 0n1
QUESTION A.10
[1.0 point]
A complete core load is in progress on a research reactor. The following data has been taken.
Number of Elements Detector A (cpm)
Detector B (cpm)
Installed 0
11 13 2
13 15 4
17 18 6
22 22 8
34 30 Using the graph paper provided, determine which of the following is the approximate number of fuel elements that will be required to be loaded for a critical mass.
- a. 8
- b. 10 c.
12
- d. 14 QUESTION A.11
[1.0 point]
Initially Nuclear Instrumentation is reading 30 CPS and the reactor has a Keff of 0.90. You add an experiment which causes the Nuclear instrumentation reading to increase to 60 CPS. Which ONE of the following is the new Keff?
- a. 0.91
- b. 0.925 c.
0.95
- d. 0.975
QUESTION A.12
[1.0 point]
Which ONE of the following describes the difference between a moderator and reflector?
- a. A reflector increases the fast non-leakage factor and a moderator increases the thermal utilization factor.
- b. A reflector increases the neutron production factor and a moderator increases the fast fission factor.
c.
A reflector decreases the thermal utilization factor and a moderator increases the fast fission factor.
- d. A reflector decreases the neutron production factor and a moderator decreases the fast non-leakage factor.
QUESTION A.13
[1.0 point]
After a week of full power operation, Xenon will reach its peak following a shutdown in approximately:
- a. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
- b. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-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. 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> QUESTION A.14
[1.0 point]
Regulating rod worth for a reactor is 0.001 K/K/inch. Moderator temperature increases by 9F, and the regulating rod moves 41/2 inches inward to compensate. The moderator temperature coefficient Tmod is
- a. +5 x 10-4
- b. -5 x 10-4 c.
+2 x 10-5
- d. -2 x 10-5
QUESTION A.15
[1.0 point]
Which ONE of the following is the difference between prompt and delayed neutrons? Prompt neutrons
- a. account for less than 1% of the neutron population, while delayed neutrons account for the rest.
- b. are released during fast-fission events, while delayed neutrons are released during the decay process.
c.
are released during the fission process (fast & thermal), while delayed neutrons are release during the decay process.
- d. are the dominating factor in determining reactor period, while delayed neutrons have little effect on reactor period.
QUESTION A.16
[1.0 point]
Using the Integral Rod Worth Curve provided identify which ONE of the following represents Kexcess
- a. Area under curve B b.
C c.
max -
C
- d. Area under curve A and B QUESTION A.17
[1.0 point]
When performing rod calibrations, many facilities pull the rod out a given increment, then measure the time for reactor power to double (doubling time), then calculate the reactor period. If the doubling time is 42 seconds, what is the reactor period?
- a. 29 sec
- b. 42 sec c.
61 sec
- d. 84 sec
QUESTION A.18
[1.0 point]
Shown below is a trace of reactor period as a function of time. Between points A and B reactor power is:
- a. continually increasing.
- b. continually decreasing.
c.
increasing, then decreasing.
- d. constant.
QUESTION A.19
[1.0 point]
Which ONE of the following is the correct definition of effective for a TRIGA reactor? The relative amount of delayed neutrons
- a. per generation corrected for resonance absorption.
- b. per generation corrected for leakage.
c.
per generation corrected for time after the fission event.
- d. per generation corrected for both leakage and resonance absorption.
QUESTION A.20
[1.0 point]
During a reactor startup, criticality occurred at a lower rod height than the last startup. Which ONE of the following reasons could be the cause?
- a. Adding an experiment with positive reactivity.
- b. Xe135 peaked.
c.
Moderator temperature increased.
- d. Maintenance on the control rods resulted in a slightly faster rod speed.
Section B Normal/Emergency Procedures and Radiological Controls Page 18 QUESTION B.1 [1.0 point]
Which ONE of the following correctly defines a Safety Limit?
- a. Limits on important process variables which are found to be necessary to reasonably protect the integrity of certain physical barriers which guard against the uncontrolled release of radioactivity.
- b. The Lowest functional capability of performance levels of equipment required for safe operation of the facility.
c.
Settings for automatic protective devices related to those variables having significant safety functions.
- d. a measuring or protective channel in the reactor safety system.
QUESTION B.2 [2.0 points, 0.5 each]
Match the values from column B for the Technical Specification limits listed in column A. (Values in Column B may be used more than once or not at all. Each limit in section A should have only one answer.)
Column A Column B
- a. Minimum Shutdown margin with the most reactive control rod fully withdrawn
$0.57 cold, no xenon, experimental facilities and experiments in place, with highest worth non-secured experiment in its most reactive state.
$1.00
- b. Total Maximum Reactivity worth of all experiments.
$2.55 c.
Total Maximum Reactivity worth of any single experiment
$3.00
- d. Maximum allowable pulse (by Technical Specifications).
$4.25
Section B Normal/Emergency Procedures and Radiological Controls Page 19 QUESTION B.3 [1.0 point]
A radiation survey instrument was used to measure an irradiated experiment. The results were 100 mrem/hr with the window open and 60 mrem/hr with the window closed. What was the beta dose rate?
- a. 40 mrem/hr
- b. 60 mrem/hr c.
100 mrem/hr
- d. 140 mrem/hr QUESTION B.4 [1.0 point]
You use a Geiger-Müller detector at the same distance from two point sources having the same curie strength. Source As gammas have an energy of 1.0 MeV, while Source Bs gammas have an energy of 2.0 MeV. Which ONE of the following would you expect for the readings due to each source?
- a. The reading from source B is four times that of source A.
- b. The reading from source B is twice that of source A.
c.
Both readings are the same.
- d. The reading from source B is half that of source A.
QUESTION B.5 [1.0 point]
Which ONE of the following is the correct definition of a CHANNEL CHECK?
- a. The combination of sensor, line, amplifier, and output devices which are connected for the purposes of measuring the value of a parameter.
- b. An adjustment of the channel such that its output corresponds with acceptable accuracy to known values of the parameter which the channel measures.
c.
A qualitative verification of acceptable performance by observation of channel behavior. This verification, may include comparison with independent channels measuring the same variable or other measurements of the variable.
- d. The introduction of a signal into the channel for verification that it is operable.
Section B Normal/Emergency Procedures and Radiological Controls Page 20 QUESTION B.6 [1.0 point]
While working in an area marked "Caution, Radiation Area," you discover your dosimeter is off scale and leave the area. Assuming you had been working in the area for 45 minutes, what is the maximum dose you would have received?
- a. 3.8 mr
- b. 35.6 mr c.
75 mr
- d. 100 mr QUESTION B.7 [1.0 point]
Which ONE of the following is the definition for Annual Limit on Intake (ALI)?
- a. The concentration of a radio-nuclide in air which, if inhaled by an adult worker for a year, results in a total effective dose equivalent of 100 millirem.
- b. 10CFR20 derived limit, based on a Committed Effective Dose Equivalent of 5 Rems whole body or 50 Rems to any individual organ, for the amount of radioactive material inhaled or ingested in a year by an adult worker.
c.
The effluent concentration of a radio-nuclide in air which, if inhaled continuously over a year, would result in a total effective dose equivalent of 50 millirem for noble gases.
- d. Projected dose commitment values to individuals, that warrant protective action following a release of radioactive material.
Section B Normal/Emergency Procedures and Radiological Controls Page 21 QUESTION B.8 [1.0 point]
You been assigned to decrease the dose rate from a point source by about a factor of 10. The point source emits a 1.5 MeV gamma. Your shielding consists of 1/2 inch thick lead sheets. How many sheets (minimum) are required? Given: the mass attenuation coefficient for lead (for 1.5 MeV gammas = 0.051 cm2/gram and density of lead is 11.4 gram/cm3.
- a. 1 sheet
- b. 2 sheets c.
3 sheets
- d. 5 sheets QUESTION B.9 [1.0 point]
Which ONE of the following conditions is a Reportable Occurrence per Technical Specifications?
- a. Operation of the reactor with a fuel temperature scram set at 500C.
- b. Operation of the reactor with bulk water temperature at 45C.
c.
Irradiation of a sample containing 20 milligrams of explosive material.
- d. Operation with pool water level 13 feet above the core.
QUESTION B.10
[1.0 point]
Your Annual limit (Occupational Dose Limit for an adult) for Total Effective Dose Equivalent is
- a. 1.25 rems
- b. 5.0 rems c.
15.0 rems
- d. 50 rems
Section B Normal/Emergency Procedures and Radiological Controls Page 22 QUESTION B.11
[1.0 point]
A small experiment sample reads 200 mR/hr with the sample 1 foot under water and the meter at the surface of the water. A reading taken 1/2 hour ago with both the sample and the meter in the same positions was 400 mR/hr. Approximately how long will it take for the reading to drop to 20 mR/hr with the sample and the meter in the same positions?
- a. 40 minutes
- b. 70 minutes c.
100 minutes
- d. 130 minutes QUESTION B.12
[1.0 point]
Which ONE of the following is the Technical Specification BASIS for the Limiting Condition of Operation for pool water temperature being maintained below 120F?
- a. To prevent damage to the resin in the purification system.
- b. To prevent cavitation in the primary coolant pump.
c.
To maintain the integrity of the Aluminum Reactor Tank.
- d. to ensure correct operation of the conductivity cells in the purification system.
QUESTION B.13
[1.0 point]
Which ONE of the following is the maximum number of times the reactor may be pulsed in a one hour period WITHOUT Reactor Supervisor permission?
- a. Three
- b. Six c.
Nine
- d. Twelve
Section B Normal/Emergency Procedures and Radiological Controls Page 23 QUESTION B.14
[2.0 points, 1/2 each]
Classify each of the experiments listed below as either Class A, Class B or Class C, according to OSTROP 18, Procedures for the Approval and Use of Reactor Experiments.
- a. Placing an empty containment tube in the Lazy Susan to test new sample containers.
- b. Placing a new experiment into a Beam Tube.
c.
An experiment requiring the movement of reactor shielding.
- d. An experiment containing explosives.
QUESTION B.15
[1.0 point]
Which ONE of the following materials would require suspension of reactor operations until approval from the Reactor Operations Committee, if that material were dropped (even in minute quantities) into the reactor tank?
- a. Mercury
c.
Glass
- d. Type 18-8 Stainless Steel QUESTION B.16
[1.0 point]
Which ONE of the listed measuring channels is REQUIRED by Technical Specifications for steady-state, pulsing and square wave modes of operation?
- a. Linear Power Level
- b. Log Power Level c.
Nvt circuit
- d. Fuel Element Temperature
Section B Normal/Emergency Procedures and Radiological Controls Page 24 QUESTION B.17
[1.0 point]
Which ONE of the following correctly defines the Emergency Plan term Protective Action Guide(s)?
- a. The person or persons appointed by the Emergency Coordinator to ensure that all personnel have evacuated the facility or a specific part of the facility.
- b. a condition or conditions which call(s) for immediate action, beyond the scope of normal operating procedures, to avoid an accident or to mitigate the consequences of one.
c.
Projected radiological dose or dose commitment values to individuals that warrant protective action following a release of radioactive material.
- d. Specific instrument readings, or observations; radiological dose or dose rates; or specific contamination levels of airborne, waterborne, or surface-deposited radioactive materials that may be used as thresholds for establishing emergency classes and initiating appropriate emergency measures.
QUESTION B.18
[1.0 point]
Per the Emergency Plan the primary Emergency Support Center is Room A100. However, for many of the more minor emergencies, the control point may be moved up to the
- a. Health Physics Office (D204)
- b. Reactor Conference Room (D300) c.
Common Room containing Large Emergency Cabinet (B134)
- d. Reactor Control Room (D302)
Section C Facility and Radiation Monitoring Systems Page 25 QUESTION C.1 [1.0 point]
Primary system water returning to the pool is ejected from an angled nozzle, causing a swirling motion of the water in the pool. Which ONE of the following is the PRIMARY purpose for this design?
- a. To increase the heat transfer rate due to increased convective flow.
- b. To decrease the activation rate of O16 to N16 due to a decrease in time within the core.
c.
To increase the transport time for N16 to reach the surface of the pool.
- d. To break up O16 bubbles in the pool thereby decreasing the production of N16.
QUESTION C.2 [1.0 point]
Which ONE of the listed Nuclear Instrumentation Channels/circuits listed below does NOT provide an input to the Regulating Rod Automatic Control Circuit.
- a. Linear Power
- b. Percent Power c.
Log-N
- d. Percent Demand QUESTION C.3 [1.0 point]
Which ONE of the following is NOT a design feature of the Purification System?
- a. Reduce radiation levels due to dissolved ions.
- b. Reduce corrosion rate due to dissolved ions.
c.
Reduce radiation levels due to suspended ions.
- d. Reduce radiation levels due to soluble gases.
Section C Facility and Radiation Monitoring Systems Page 26 QUESTION C.4 [2.0 points, 1/2 each]
Identify each of the control rods as having either a fuel follower or an air follower.
- a. Shim
- b. Safety c.
- d. Regulating QUESTION C.5 [1.0 point]
Which ONE of the following electrical load is NOT powered by the Emergency Generator or inverter batteries on a loss of site power?
- a. Argon Fan
- b. Television Monitor c.
Stack Monitor Pump
- d. Cypher Lock QUESTION C.6 [1.0 point]
Which ONE of the following is NOT an interlock associated with pulsing operations.
- a. Switch in Pulse Mode.
- b. Transient Rod fully inserted.
c.
Period greater than 50 seconds.
- d. Power less than 1 Kwatt.
Section C Facility and Radiation Monitoring Systems Page 27 QUESTION C.7 [1.0 point]
What design feature minimizes flux peaking in the central thimble.
- a. Filling with N2.
- b. An aluminum plug.
c.
A cadmium plug.
- d. A Zirconium plug QUESTION C.8 [1.0 point]
Which ONE of the following is the reason for the holes located at the bottom of the Central Thimble Assembly?
- a. To allow cooling flow through the thimble, at power.
- b. To allow for evacuation of the water in the thimble.
c.
To allow for fasteners to bolt the thimble to the bottom support plate.
- d. To fit over pins in the safety support plate for proper alignment.
QUESTION C.9 [1.0 point]
Which ONE of the following methods is the normal procedure for preventing basin water in the cooling tower from freezing?
- a. Use of Hand-held heat guns.
- b. Running fans in reverse.
c.
Heaters built into water sump.
- d. Steam connection from University facilities.
Section C Facility and Radiation Monitoring Systems Page 28 QUESTION C.10
[1.0 point]
Which ONE of the following components in the purification system is PRIMARILY responsible for maintaining the primary coolant system conductivity low.
- a. The surface skimmer
- b. The pre-demineralizer filter c.
The demineralizer
- d. The post-demineralizer filter QUESTION C.11
[1.0 point]
What is the purpose of the Cadmium Lined In-Core Irradiation Tube (CLICIT)
- a. To allow irradiation of samples by neutrons with an energy level greater than 0.5 ev.
- b. To allow irradiation of samples by neutrons with an energy level of less than 0.5 ev.
c.
To allow irradiation of samples by gammas within the core.
- d. To allow irradiation of samples by alphas produced by the neutron interaction with the cadmium.
QUESTION C.12
[1.0 point]
The ventilation system is designed to maintain reactor bay pressure slightly negative pressure with respect to the atmospheric pressure. If the outside atmospheric pressure increases, which ONE of the following actions will automatically occur to compensate the reactor bay pressure? A pressure regulator will generate a signal to
- a. Increase the Reactor Bay Supply fan speed to increase bay pressure.
- b. Decrease the Reactor Bay Exhaust fan speed to increase bay pressure.
c.
Go more closed on a damper in the ventilation exhaust ducting increasing bay pressure.
- d. Go more open on a damper in the ventilation supply ducting increasing bay pressure.
Section C Facility and Radiation Monitoring Systems Page 29 QUESTION C.13
[1.0 point]
While operating in AUTOMATIC mode, the reactor operator depresses the UP button for a control rod. At the same time, the AUTOMATIC circuit energizes to drive the regulating rod up. Which ONE of the following will actually take place.
- a. Due to the ONE ROD WITHDRAWAL interlock, only the CONTROL ROD will move.
- b. Due to the ONE ROD WITHDRAWAL interlock, only the REGULATING ROD will move.
c.
Due to the ONE ROD WITHDRAWAL neither rod will move.
- d. THE ONE ROD WITHDRAWAL does not apply and both rods will move.
QUESTION C.14
[1.0 point]
Which ONE of the following is the actual method used to generate the rod position indication on the control panel?
- a. Voltage changes generated by the movement of a lead screw between two coils of a transformer.
- b. A potentiometer linked to the rod drive motor c.
A series of several reed switches which as the rod moves up close to generate a current proportional to rod position.
- d. A servo motor connected to the UP and DN buttons which when either button is depressed generates a signal proportional to rod speed.
QUESTION C.15
[1.0 point]
WHICH ONE of the following detectors is used primarily to measure N16 release to the environment?
- a. NONE, N16 has too short a half-life to require environmental monitoring.
- b. Stack Gas Monitor c.
Air Particulate Monitor
- d. Area Radiation Monitor Channel # 5
Section C Facility and Radiation Monitoring Systems Page 30 QUESTION C.16
[1.0 point]
You (the console operator) receive a report of thick black smoke coming from the demin pump. Where would you send someone to deenergize the breaker that supplies the pump?
- a. Room 106 to Sub Distribution Panel A
- b. Room 106 to Panel G c.
Reactor Bay Panel A
- d. First Floor Hallway Panel F QUESTION C.17
[1.0 point]
To detect Neutrons, the Uncompensated Ion Chambers are lined with a.
5B10 b.
6C12 c.
92U235 d.
94Pu239 QUESTION C.18
[1.0 point]
Match each purpose in column A with its associated fuel element component listed in column B.
Column A Column B
- a. moderator
- 1. Graphite
- b. reflector
- 2. Zirconium-Hydride c.
resonance absorber
- 3. Erbium
- d. burnable poison
Section A Theory, Thermo & Fac. Operating Characteristics Page 31 A.1 c
REF:
Glasstone, S. and Sesonske, A, Nuclear Reactor Engineering, Kreiger Publishing, Malabar, Florida, 1991, § A.2 c
REF:
OSU Training Manual Vol. 3, p. 17 A.3 d
Worth of rods: $2.30 + $2.25 + $1.10 = $5.65. SDM = Worth of rods less Kexcess less reactivity of most worth rod. SDM = $5.65 - $2.50 - 2.30 =
5.65 - $4.80 = 0.85 REF:
OSU Training Manual Vol. 3, p. 29 A.4 b
REF:
OSU Training Manual Vol. 3, p. 11 A.5 d
REF:
Standard NRC Question A.6 a
REF:
Glasstone, S. and Sesonske, A, Nuclear Reactor Engineering, Kreiger Publishing, Malabar, Florida, 1991, § A.7 b
REF:
ln (P/P0) x period = time, ln(1000) x 26 = 6.908 x 26 = 179.6 180 seconds A.8 b
REF:
OSU Training Manual Vol. 3, p. 10 A.9 a
REF:
OSU Training Manual Vol. 3, pp. 41-43.
A.10 c
REF:
OSU Training Manual Vol. 3, pp. 45-50 A.11 c
REF:
CR2/CR1) = (1 - Keff1)/(1 - Keff2) 60/30 = (1 - 0.900)/(1 - Keff2) 1-Keff2 = 1/2 x 0.1 = 0.05 Keff2 = 1 - 0.05 = 0.95 A.12 a
REF:
OSU Training Manual Vol. 3, pp. 12 and 15.
Section A Theory, Thermo & Fac. Operating Characteristics Page 32 A.13 b
REF:
OSU Training Manual Vol. 3, p. 24.
A.14 a
REF:
(4.5 x 0.001) ÷ 9 = 0.0045 ÷ 9 = 0.0005 = 5 x 10-4, also OSU Training Manual Vol. 3, p. 19 A.15 c
REF:
OSU Training Manual Vol. 3, p. 30.
A.16 c
REF:
OSU Training Manual Vol. 3, p. 29.
A.17 c
REF:
ln (2) = -time/
= time/(ln(2)) = 60.59 61 seconds A.18 a
REF:
Standard NRC Question A.19 d
REF:
OSU Training Manual Vol. 3, p. 30.
A.20 a
REF:
OSU Training Manual Vol. 3, p. 30.
Section B Normal/Emergency Procedures and Radiological Controls Page 33 B.1 a
REF:
Technical Specifications § 1.22 B.2 a, $0.57; b, $3.00; c, $2.55 d, $2.55 REF:
Technical Specification §§ 3.2, 3.3 and 3.8 B.3 a
REF:
Instrument reads only dose with window closed. Instrument reads both and dose with window open. Therefore, dose is window open dose less window closed dose.
B.4 c
REF:
GM tubes are NOT sensitive to energy level.
B.5 c
REF:
Technical Specifications § 1.31 B.6 c
REF:
10 CFR 20.1003 Maximum dose in a radiation area is 100 mr/hr. 100 mr/hr x 0.75 hr = 75 mr.
B.7 b
REF:
10CFR20.1003 B.8 c
REF:
First calculate = 0.051 cm2/g x 11.4 g/cm3 = 0.5814 cm-1.
Next calculate thickness. I = I0 e-Fx ln(1/10) = - x x = - [ln(1/10)]/ = ln (0.1)/0.5814 = 3.96 4.0 cm Finally calculate number of sheets:
(4.0 cm)/(2.54cm/in) = 1.57 inches or about 3 sheets.
B.9 d
REF:
Technical Specifications, 2.2, 3.3.a, 3.6.d and 3.2.2 (Table 2)
B.10 b
REF:
10CFR20.2001.a(1)
B.11 c
REF:
A = A0 e-t Solve for 200 = 400 e-30minutes ln (200/400) = - x 30minutes ln(1/2)/30 minutes = - = 0.0231 Next solve for time 20 = 200 e(-0.0231 x time) ln (1/10)/-0.0231 = time = 99.7 minutes 100 minutes B.12 c
Section B Normal/Emergency Procedures and Radiological Controls Page 34 REF:
B.13 b
REF:
OSTROP 4, Reactor Operation Procedures, p. 9.
B.14 a, A; b, B; c, B; d, C REF:
Technical Specifications, 1.0 Definitions.
B.15 a
REF:
OSTROP 7, Operating Procedures for Reactor Water Systems, § I.F, Warning p. 3.
B.16 d
REF:
Technical Specification Table in § 3.5.2.
B.17 c
REF:
Emergency Plan § 2.0 Definitions.
B.18 d
REF:
Emergency Plan § 8.0 Emergency Equipment and Facilities.
Section C Facility and Radiation Monitoring Systems Page 35 C.1 c
REF:
Oregon State (OSTR) Training Manual Vol. I, page 106.
C.2 b
REF:
ORST Training Manual Volume II, Figure 2.16 C.3 d
REF:
ORST Training Manual Volume I, p. 106 C.4 a, Fuel; b, Fuel; c, Air; d, Fuel REF:
NRC Examination Question Bank, also Volume 1, pages 40-44, OSU Triga Manual C.5 a
REF:
OSTROP 22 C.6 c
REF:
NRC Exam Question Bank, also: Volume 2, pages 23-28, OSU Triga Manual C.7 b
REF:
NRC Exam administered 1998, also, OSU Training Manual Vol I, p, 89 second paragraph.
C.8 b
REF:
OSU Training Manual, p. 90 C.9 c
REF:
OSU Training Manual, p. 126, last paragraph.
C.10 c
REF:
NRC exam administered 1996, also OSTR Training Manual Vol. I, p. 116.
C.11 a
REF:
OSTR Training Manual, Vol. I, p. 81 C.12 c
REF:
OSTR Training Manual, Vol. I p. 148.
C.13 d
REF:
OSTR Training Manual Vol II, p. 9.
Section C Facility and Radiation Monitoring Systems Page 36 C.14 b
REF:
OSTR Training Manual, Vol I, fig. 1.26 TRIGA Control Rod Drive Mechanism, p. 48.
C.15 a
REF:
Standard NRC Question C.16 b
REF OSTROP 22.0 Emergency Power System, Fig. 22.1 One-Line Schematic Power Distribution C.17 a
REF:
ORST Training Manual Volume 2, § III.C.
C.18 a, 2; b, 1; c, 3; d, 3 REF:
ORST Training Manual Volume 1.
U. S. NUCLEAR REGULATORY COMMISSION NON-POWER INITIAL REACTOR LICENSE EXAMINATION FACILITY:
Oregon State University REACTOR TYPE:
TRIGA (Pulsing)
DATE ADMINISTERED:
2002/04/29 CANDIDATE:
INSTRUCTIONS TO CANDIDATE:
Write answers on the answer sheet provided. Attach answer sheets to the examination. Points for each question are indicated in brackets. A 70% overall is required to pass the examination. Examinations will be picked up three (3) hours after the examination starts.
Category Value
% of Total Candidates Score
%of Category Value Category 20 33 A. Reactor Theory, Thermodynamics, and Facility Operating Characteristics 20 33 B. Normal and Emergency Operating Procedures and Radiological Controls 20 33 C. Plant and Radiation Monitoring Systems
60 TOTALS FINAL GRADE All work done on this examination is my own. I have neither given nor received aid.
Candidates 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 answer sheets.
10.
Ensure all information you wish to have evaluated as part of your answer is on your answer sheet.
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.
Pmax ( )2 2 (k)
Q mcp T m H UA T
( 1 x 10&4 seconds SCR
S
S 1Keff eff 0.1 seconds &1 CR1(1Keff1) CR2(1Keff2)
CR1(
- 1) CR2(
2)
SUR 26.06 eff
M
1Keff0 1Keff1 M
1 1Keff
CR1 CR2 P P0 10SUR(t)
SDM (1Keff)
(
-
P
(1 )
P0
Keff2Keff1 keff1xKeff2
(
-
eff T1/2 0.693
(Keff1)
Keff DR DR0 e & t DR1d1 2 DR2d2 2
DR 6CiE(n)
R 2
(
2 )2 Peak2
(
1 )2 Peak1 EQUATION SHEET
DR Rem, Ci curies, Fuchs Pulse Model Equations (Estimates)
E Mev, R feet
Section A Theory, Thermo & Fac. Operating Characteristics Page 41 T
T C
max
(
)
=
° 0
2
P k I MW max
(
)
( )
=
2 2
E k
MWS tot =
2(
)
( )
I = 39 x10-6 sec.
= 1.26 x 10-4 k/k/C k = 9.6 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 A.1 a b c d ___
A.11 a b c d ___
A.2 a b c d ___
A.12 a b c d ___
A.3 a b c d ___
A.13 a b c d ___
A.4 a b c d ___
A.14 a b c d ___
A.5 a b c d ___
A.15 a b c d ___
Section A Theory, Thermo & Fac. Operating Characteristics Page 42 A.6 a b c d ___
A.16 a b c d ___
A.7 a b c d ___
A.17 a b c d ___
A.8 a b c d ___
A.18 a b c d ___
A.9 a b c d ___
A.19 a b c d ___
A.10 a b c d ___
A.20 a b c d ___
Section B Normal/Emerg. Procedures & Rad Con Page 43 B.1 a b c d ___
B.10 a b c d ___
($)
B.2a 0.57 1.00 2.55 3.00 5.00 ___
B.11 a b c d ___
($)
B.2b 0.57 1.00 2.55 3.00 5.00 ___
B.12 a b c d ___
($)
B.2c 0.57 1.00 2.55 3.00 5.00 ___
B.13 a b c d ___
($)
B.2d 0.57 1.00 2.55 3.00 5.00 ___
B.14a A B C ___
B.3 a b c d ___
B.14b A B C ___
B.4 a b c d ___
B.14c A B C ___
B.5 a b c d ___
B.14d A B C ___
B.6 a b c d ___
B.15 a b c d ___
Section B Normal/Emerg. Procedures & Rad Con Page 44 B.7 a b c d ___
B.16 a b c d ___
B.8 a b c d ___
B.17 a b c d ___
B.9 a b c d ___
B.18 a b c d ___
Section C Plant and Radiation Monitoring Systems Page 45 C.1 a b c d ___
C.10 a b c d ___
C.2 a b c d ___
C.11 a b c d ___
C.3 a b c d ___
C.12 a b c d ___
C.4a Air Fuel ___
C.13 a b c d ___
C.4b Air Fuel ___
C.14 a b c d ___
C.4c Air Fuel ___
C.15 a b c d ___
C.4d Air Fuel ___
C.16 a b c d ___
C.5 a b c d ___
C.17 a b c d ___
C.6 a b c d ___
C.18a 1 2 3 ___
Section C Plant and Radiation Monitoring Systems Page 46 C.7 a b c d ___
C.18b 1 2 3 ___
C.8 a b c d ___
C.18c 1 2 3 ___
C.9 a b c d ___
C.18d 1 2 3 ___
Critical Rod Height 0
Rod fully out Integral Rod Worth Curve HC C
Max 0
A B