ML20134H063
| ML20134H063 | |
| Person / Time | |
|---|---|
| Site: | Oregon State University |
| Issue date: | 10/30/1996 |
| From: | Doyle P NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20134G028 | List: |
| References | |
| 50-243-OL-96-01, 50-243-OL-96-1, NUDOCS 9611130512 | |
| Download: ML20134H063 (40) | |
Text
,
U. S. NUCLEAR REGULATORY COMMISSION OPERATOR LICENSING INITIAL EXAMINATION REPORT REPORT NO.:
50-243/OL-96-01 FACILITY DOCKET NO.:
50-243 FACIUTY LICENSE NO.:
R-160 FACILITY:
Oregon State University EXAMINATION DATES:
October 21,1996 EXAMINER:
Paul Doyle, Chief Examiner N'bO[94 SUBMITTED BY:
~
PdulDoylfhief/xaminer
/ Date
SUMMARY
During the week of October 21,1996, the NRC administered a licensing examination to a Reactor Operator candidate at Oregon State University. The candidate passed all portions of the NRC administered examination.
REPORT DETAILS 1.
Examiners, j
Paul Doyle, Chief Examiner 2.
Results:
RO SRO PASS / Fall TOTAL PASS / Fall
_ PASS / Fall Written 1/0 0/0 1/0 Operating Tests 1/0 0/0 1/0 Overall 1/0 1/0 1/0 3.
Exit Meeting:
Paul Doyle, NRC, Chief Examiner Brian Dodd, Oregon State Univ., Radiation Center Director Jack Higginbotham, Oregon State Univ., Reactor Supervisor Arthur Hall, Oregon State Univ., Reactor Administrator j
During the' exit meeting Mr. Doyle thanked the facility staff for their cordial assistance in the adminstration of the examinations, and reminded them of the importance to submit their comments on the written examination to expedite the grading process.
1 ENCLOSURE 1 1
n 9611130512 961107 PDR ADOCK 05000243 V
4 Facility Comments with NRC Resolution QUESTION (A.15)
Which ONE of the following conditions would result in a DECREASE in core excess reactivity?
a.
Replace four fuel bundles with four new bundles (same original specifications).
b.
Replace a control rod with a new control rod (same original specifications).
c.
Bumout of xenon following restart from a scram (12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> following scram).
d.
Placement of an experiment containing xenon into the lazy susan.
ANSWER (A.15)
REFERENCE (A.15)
OSTR Training Manual Vol.111, p. 28 FACILITY COMMENT Answer a is also acceptable because our current fuel element worth is greater than that of new fuel due to the erbium being bumed at a greater rate than the uranium. Answer b is correct because boron burnup in the control rod. A new control rod could also contain a fuel follower and the argument of Answer a also applies. Answer d is also acceptable because such an experiment would cause a small decrease in the core excess reactivity, based upon actual experiments involving loading the rack with cadmium covered samples.
NRC RESOLUTION Agree with facility comments. Question deleted from this examination.
QUESTlON (A.19)
At which ONE of the following times would the MAXIMUM amount of xenon in the core?
(Assume initial condition was in effect for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> before power change.)
a.
4 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a power increase from 50% to 100%
b.
4 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a power decrease frotn 100% to 50%
c.
8 to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> following a startup to 100%.
d.
8 to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> following a reactor shutdown from 100%
ANSWER (A.19) a l
. REFERENCE (A.19)
OSTR Training Manual Vol.111, pp. 22 - 24.
FACILITY COMMENT The correct answer is d, according to Figure 3.7 and the text of page 24 of the OSTR Training Manual, Volume 3. We suspect that since the correct references are cited that the answer key contains a typographical error.
NRC RESOLUTION Agree with facility comments. Anwer key modified to accept d as correct response.
ENCLOSURE 2
QUESTION (B.5)
Identify whether each experiment listed is classified as Class A, B or C from OSTROP 18 Procedures for the Approval and Use of Reactor Experiments.
a.
Placing an empty containment tube in 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 requiring the movement of reactor fuel.
ANSWER (B.5)
I a, A; b, B; c, B; d, C 1
REFERENCE (B.5)
OSTROP 18.0 518.4, Classification of Reactor Experiments p. IV.18.3.
FACILITY COMMENT The problem statement has a typographical error, the second word should be whether l
Instead of wether. Also, answer d is acceptable because specific Class B experiments authorize fuel movement, for example B-3, when the CLICIT is inserted or removed.
NOTE: Per telephone conversation with J. Higgenbotham, this comment refers to matching part D of question.
NRC RESOLUTION Agree with facility comment. Matching choice D has been modified to accept either B or C as correct responses.
QUESTION (B.7)
During an EMERGENCY, you received 2 REM performing actions that you volunteered to do to mitigate the accident. How is this radiation dose tracked?
a.
It is tracked as part of the normal dose (5 REM / year) allowed for a radiation worker.
b.
It is tracked as part of your Planned Special Exposure limit (5 REM per year,15 REM per lifetime.
c.
As an emergency dose it is not tracked on-site, but is reported to NRC.
d.
As an emergency dose it is not tracked at all.
ANSWER (B.7) b REFERENCE (8.7) 10 CFR 20.1206 FACILITY COMMENT We find that answer a is acceptable because of an OSTR institutional policy of not implementing Planned Special Exposures and that 2 REM is within the allowed annual occupational dose limit.
NRC RESOLUTION Agree with facility comments. Answer key modified to accept either b or a as correct response.
QUESTION (B.17)
Which ONE of the following is the SAFETY LIMIT for the maximum temperature for a FLIP TRIGA fuel element.
a.
800'C b.
950*C c.
1000*C d.
1150'C ANSWER (B.17) l c
REFERENCE (B.17)
OSTR Technical Specifications, % 2.1 Safety Limit-Fuel Element Temperature, p. 6 FACILITY COMMENT The correct answer is d as per the cited reference in the answer key.
NRC RESOLUTION Agree with facility comments. Answer key modified to accept d as correct response.
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Operator Licensing Examination u
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Oregon State University 10/21/96 ENCLOSURE 3
Section A R Theory. Thermo. and Facility Characteristics Page 2 I
QUESTION (A.1)
[1.0)
The reactor was shutdown, with a shutdown margin of 2.5 dollars and count rate on the source range reads 15 counts / minute. After placing a sample into the reactor the count rate increased to 60 counts / minute. What is the worth of the sample?
a.
-90p b.
= + 90 c.
= +1.50$
d.
-1.50$
QUESTION (A.2)
[1.0)
Which ONE of the following is the correct reason that indicated power will stabilize several hours following a reactor scram. (Assume source inserted into core, source range j
instrumentation operable reading 3 cpm, and no reactivity changes, i.e. no temperature l
changes, fuel movement, experiments inserted or removed, etc.)
a.
Continuing decay of the shortest lived neutron precursor b.
Gamma saturation of the source range detector.
c.
Suberitical multiplication of Source Neutrons d.
Neutron activation of the Source Range detector QUESTION (A.3)
[1.0)
A fast neutron will lose the most energy per collision when striking an atom of which ONE of the following elements?
A.
H' b.
H' c.
C'*
l d.
U" i
i i
i Section A R Theory. Thermo. and Facility Characteristics Page 3 QUESTION (A.4)
[1.0)
Which ONE of the following elements has the SMALLEST cross-section for absorption of
]
THERMAL neutrons?
i a.
,H' B.
H2 3
C.
sC'*
D.
U2'8 s2 i
QUESTION (A.5)
[2.0)
Which ONE of the following describes the MAJOR contributor to the production and depletion 2
of Xenon respectively in a STEADY-STATE OPERATING reactor?
4 Production Deoletion a.
Radioactive decay of lodine Radioactive Decay 4
b.
Radioactive decay of lodine Neutron Absorption c.
Directly from fission Radioactive Decay d.
Directly from fission Neutron Absorption QUESTION (A.6)
[1.0) p for U235 is 0.0065, but p, at the Oregon State TRIGA reactor is 0.0070. Why is p, larger?
a.
The reactor contains U238 which for fast fissioning has a larger p than U235 2
b.
The reactor contains Pu which for thermal fissioning has a larger s than U23s c.
Delayed neutrons are born at a higher average energy level than fission neutrons resulting in a greater amount of neutrons from fast fissions.
d.
Delayed neutrons are born at a lower average energy level than fission neutrons resulting in fewer being lost to fast leakage.
i l
Section A R Theory. Thermo. and Facility Characteristics Page 4 QUESTION (A.7)
[1.0)
Which ONE of the following is the correct definition of REACTIVITY?
a.
A measure of the core's deviation from criticality b.
A measure of the core's fuel depletion.
c.
A measure of the core's state with all control rods fully withdrawn.
d.
A measure of the core's state at prompt criticality.
QUESTION (A.8)
[1.0)
Which ONE of the following is the difference between prompt and delayed neutrons? Prompt neutrons.
1 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.
1 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 i
little effect on reactor period.
l QUESTION (A.9)
[1.0)
K, for the reactor is 0.85. If you place an experiment worth +17.6% into the core, what will the new K, be?
a.
0.995 b.
0.9995 c.
1.005 d.
1.05 4
Section A R Theorv. Thermo. and Facility Characteristics Page5 QUESTION (A.10)
[1.0)
Given an average rod reactivity worth of 0.1%/ inch, and cr of-0.005%Ak/*C. If fuel rm temperature were to increase by 150*C, how far and in what direction would you have to move the rod to compensate?
a.
7.5 inches, inward b.
0.75 inches,inward c.
7.5 inches, outward d.
0.75 inches, outward QUESTION (A.11)
[1.0]
The primary pump is operating at 490 gpm. The AT across the primary side of the heat exchanger is 14*F. What is the power level being generated in the reactor?
a.
12.1 kilowatts b.
96.6 kilowatts c.
121 kilowatts d.
966 kilowatts QUESTION (A.12)
[1.0)
Five minutes following a shutdown you note a Nuclear instrumentation reading of about 3 x 105 counts. What reading would you expect after another three minutes?
5 a.
10 counts b.
3 x 10d counts c.
1.5 x 104 counts d.
1 04 counts i
i f
4 a
1 Section A R Theory. Thermo. and Facility Characteristics Page 6 QUESTION (A.13)
[1.0]
Which ONE of the following is the reason for an installed neutron source within the core? A startup without an installed neutron source...
a.
is impossible as there would be no neutrons available to start up the reactor.
i l
l b.
would be very slow due to the long time to build up neutron population from so low a level.
l 1
l c.
could result in a very short period due to the reactor going critical before neutron population built up high enough to be read on nuclear instrumentation.
l d.
can be compensated for by adjusting the compensating voltage on the source range i
detector.
QUESTION (A.14)
(1.0)
The source was removed from an operating reactor. Later, the source was reinstalled and the Reactor Operator noted reactor power increasing LINEARLY. What was the condition of the reactor when the source was inserted? (Assume source has no reactivity worth, and no other changes in reactor parameters.) The reactor was...
a.
very subcritical b.
slightly suberitical c.
exactly critical d.
slightly suberitical QUESTION (A.15)
[1.0]
QUESTION DELETED PER FACILITY COMMENT
'"t ;h ON5 Of th; f;";;;'n; andnl;; ;;;;'d Tsu". ln ; 050S5.'05 ln s; ;;nn rxt.l?ny?
n;pl;; f;;r fu;l bund:n ;;;th f;ur n;;; bund:s (am ed;l=l ;p=F stl=;).
b.
n;p';; ; 2..ti;l red ;;lth ; n;;; s..t.el i;d (;;m; edgl=l ;pxl';;t en;).
Su- ;;;t of ;== %ll;;;lng ist;f. 77;m ; nam (12 h= ; %ll;;;;n; nam).
d.
P:;nmat of = 0xp-dm;nt ant;lnia; an= lnt; th; lny ann l
.. - ~
l Section A R Theorv. Thermo. and Facility Characteristics Page 7 QUESTION (A.16)
[12.0) l Given the height of a $1.50 Pulse. By what factor would you expect pulse height to increase by I
for a $2.00 pulse?
a.
1.333 b.
2 j
c.
1.778 d.
4 QUESTION (A.17)
[1.0) l Which ONE of the following listed factors is MOST affected by a change in poison level in the l
core?
a.
Fast Fission (c) l l
b.
Thermal Utilization (f) l c.
Resonance Escape (p) d.
Reproduction Factor (q)
QUESTION (A.18)(1.0) l During a startup, you are withdrawing a control rod in equal increments (distance) as the reactor approaches criticality. Which ONE of the following statements best describes reactor behavior? (Assume reactor remains slightly subcritical.)
i a.
Each rod withdrawal will add the same amount of reactivity.
1 b.
Reactor power will increase by the same amount for each rod withdrawal.
c.
The time for power to stabilize will increase.
d.
Decreasing time between withdrawals will result in a lower critical rod height.
l i
1 Section A R Theorv. Thermo. and Facility Characteristics Page 8 QUESTION (A.19)
[1.0)
At which ONE of the following times would the MAXIMUM amount of xenon in the core?
- (Assume initial condition was in effect for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> before power change.)
a.
4 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a power increase from 50% to 100%
b.
4 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a power decrease from 100% to 50%
c.
8 to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> following a startup to 100%.
d.
8 to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> following a reactor shutdown from 100%
QUESTION (A.20)
[1.0)
Which ONE of the following is the major source of the recoverable energy released during fission?
a.
Kinetic energy of the fission neutrons b.
Kinetic energy of the fission fragments.
c.
Decay of the fission fragments.
d.
Prompt Gamma Rays
Section B Normal. Emeroency and Radioloaical Control Procedures Page 9 QUESTION (B.1)
[1.0)
A point source generates a radiation field of 1 Rem /hr (y) at 3 feet. Addition of % inch of lead shielding reduces the radiation field to 500 mrem /hr at 3 feet. What will be the radiation field reading if you add another % inch of lead shielding?
a.
250 mrem /hr b.
125 mrem /hr c.
62.5 mrem /hr d.
31.25 mrem /hr i
QUESTION (B.2)
[1.0)
Which ONE of the following is the 10 CFR 20 definition of TOTAL EFFECTIVE DOSE EQUlVALENT (TEDE)?
I a.
The sum of the deep does equivalent and the committed effective dose equivalent.
i b.
The dose that your whole body receives from sources outside the body.
c.
The sum of the extemal deep dose and the organ dose.
d.
The oose to a specific organ or tissue resulting from an intake of radioactive material QUESTION (B.3)
[2.0)
Match the radiation reading from column A with its corresponding radiation area classification (per 10 CFR 20) listed in column B.
COLUMN A COLUMN B a.
10 mrem /hr 1.
Unrestricted Area b.
150 mrem /hr 2.
Rediation Area c.
10 Rem /hr 3.
550 Rem /hr 4.
Very High Radiation Area
.._._m..._
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Sectiony Normal. Emeroenev and Radioloaical Control Procedures Page 10 QUESTION (B.4)
[2.0) l Identify each of the following as either a Channel Check, Channel Test or Channel Calibration, as defined by Technical Specifications.
a.
Observe overlap between startup channel and intermediate range.
b.
Replace resistance temperature detector (RTD) with a precision resistance bridge to check proper temperature circuit operation.
c.
Monitor nuclear instrumentation after shutdown verifying that power decreases by a factor of 10 in three minutes.
d.
Based on a heat balance (calorimetric) performed on the primary system, adjust Nuclear Instrumentation.
l
[
QUESTION {B.5)
[2.0)
Identify whether each experiment listed is classified as Class A, B or C from OSTROP 18 Procedures for the Approval and Use of Reactor Experiments.
l a.
Placing an empty containment tube in 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 requiring the movement of reactor fuel.
QUESTION (B.6)
[1.0)
Which one of the following is the correct reason that the crane bridge is restricted to the NORTH side of the bay while the reactor is operating at or above 300 Kilowatts?
a.
To reduce the potential of neutron embrittlement of the crane parts.
b.
To reduce the buildup of fission gases due to restricted ventilation flow.
c.
To reduce the buildup of N and Ar due to restricted ventilation flow.
d d.
To reduce the possibility of damage to the core during crane operations.
_ __. _. ~. _ _.. _ _
Section B Normal. Emeraency and Radiolooical Control Procedures Page 11 QUESTION (B.7)
(1.0)
During an EMERGENCY, you received 2 REM performing actions that you volunteered to do to mitigate the accident. How is this radiation dose tracked?
a.
It is tracked as part of the normal dose (5 REM / year) allowed for a radiation worker, l
b.
't is tracked as part of your Planned Special Exposure limit (5 REM per year,15 REM per lifetime.
c.
As an emergency dose it is not tracked on-site, but is reported to NRC.
d.
As an emergency dose it is not tracked at all.
QUESTION (B.8)
[1.0)
Who of the following listed personnel has responsibility for authorizing reentry following an i
evacuation of the facility?
l l
a.
Emergency Director b.
Campus Radiation Safety Officer l
c.
Emergency Coordinator d.
Senior Health Physicist j
QUESTION (B.9)
[1.0)
During maintenance in the reactor bay a hot spot is reading 5,000 mrem /hr at a distance of 2 feet. Assuming no shielding available, at what distance from the source must you set up a j
HIGH radiation area boundary?
a.
4% feet b.
15 feet c.
21 feet d.
101 feet i
4 4
Section B Normal. Emeroency and Radioloaical Control Procedures Page 12 QUESTION (B.10)
[1.0)
You are given a personnel dosimeter with a maximum reading of 200 mr/hr. You are assigned to work in an area with a general radiation dose of 15 mr/hr. How long can you work without having to leave to rezero your dosimeter? (Assume no hot spots, and all equipment is working fine.)
a.
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-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.
16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> d.
20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> QUESTION (B.11)
[1.0)
The Emergency Plan discusses all four of the listed Emergency Classifications. Wt. ch ONE of the four listed is described as NOT CREDIBLE based on the Safety Analysis Report?
a.
Alert b.
Notification of Unusual Event c.
Personnel and Operational Events d.
Site Area Emergency l
QUESTION (B.12)
[1.0)
Which ONE of the following is the Technical Specification BASIS for the Limiting Condition for Operation for pool water temperature being maintained below 120*F7 l
a.
To prevent damage to 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.
I t
Section B Normal. Emeraencv and Radioloaical Control Procedures Page 13 QUESTION (B.13)
[1.0]
Who by title is the lowest level of OSTR management who may authorize restart of the reactor.
a.
Reactor (Console) Operator b.
Reactor Supervisor c.
Reactor Administrator d.
Radiation Center Director QUESTION (B.14)
[1.0]
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 l
b.
Six l
l c.
Nine d.
Twelve QUESTION (B.15)
[1.0]
During an emergency, plant conditions have degraded badly enough that in order to ensure the health and safety of the public you must deviated from technicai specifications, per 10 CFR 50.54(x). What is the minimum level of OSTR management approval you must have to take this action?
l
- a.
Licensed Reactor Operator l
b.
Any licensed Senior Reactor Operator c.
Reactor Administrator d.
A quorum of the Reactor Operations Committee.
i l
4 r
4
i Section B Normal. Emeraency and Radioloaical Control Procedutta Page 14 QUESTION (B.16)
[1.0)
During operations the console operator encounters an unusual condition, prompting him/her to shutdown the reactor and suspend operations. Which one of the following is the concurrences required to restart the reactor 7 a.
The Console Operator, and the Reactor Supervisor.
b.
The Reactor Supervisor and the Reactor Administrator.
c.
The Console Operator, and the Reactor Supervisor.
d.
The Console Operator, the Reactor Supervisor and the Reactor Administrator.
QUESTION (B.17)
[1.0)
Which ONE of the following is the SAFETY LIMIT for the maximum temperature for a FLIP TRIGA fuel element.
l l
a.
800*C l
b.
950*C 1
c.
1000*C l
d.
1150*C l
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i A
1
Section C Plant and Radiation Monitorina Systems Page 15 QUESTION (C.1)
[2.0) identify whether the equipment listed remains energized (System A), reenergizes after emergency generator starts [20 seconds) (System B) or remains deenergized (No Power) following a loss of normal AC power to the facility.
a.
Argon Fan b.
Public Address System c.
Fire Alarm System d.
Stack Monitor Pump e.
Cypher Locks f.
Rabbit Fan QUESTION (C.2)
[1.0)
Which ONE of the following is the main function performed by the DISCRIMINATOR circuit in the startup channel?
a.
To generate a current signal equal and of opposite polarity as the signal due to gammas generated within the Log-N Channel Detector.
)
b.
. To filter out small pulses due to gamma interactic as, bdeng only pulses due to neutron events within the Log-N Channel Detector.
c.
To convert the linear output of the Log-N Channel Detector to a logarithmic signal for metering purposes.
d.
To convert the logarithmic output of the metering circuit to a 5t (differential time) output for period metering purposes.
l I
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Section C Plant and Radiation Monitorina Systems Page 16 l
QUESTION (C.3)
[1.0) l Water returning to the pool from the primary system is ejected through an angled nozzle, which causes a swirling motion 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 O to N due to a decrease in time within the core.
c.
To increase the transport time for N to reach the surface of the pool.
d.
To break up O bubbles in the pool thereby decreasing the production of N.
QUESTION (C.4)
[1.0)
Which ONE of the Nuclear instrumentation channels / circuits listed below does NOT provide an input to the Regulating Rod Automatic Control circuit?
a.
Linear Power b.
% Power c.
Log-N d.
Percent Demand QUESTION (C.5)
[1.0)
Which ONE of the following is NOT a design function of the purification system.?
t a.
Reduce radiation level due to dissolved ions.
I b.
Reduce corrosion due to dissolved ions, c.
Reduce radiation levels due to suspended solids.
d.
Reduce radiation level due to gases in solution.
l l
I
i Section C Plant and Radiation Monitorina Systems Page 17 QUESTION (C.6)
[1.0)
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 W/THDRAWAL interlock, only the REGULATING ROD will move.
c.
Due to the ONE ROD WITHDRAWAL neither rod will move.
d.
THE ONE ROD WlTHDRAWAL does not apply and both rods will move.
QUESTION (C.7)
[1.0]
Which ONE of the following is the gas used in the pneumatic tube system?
a.
Air b.
CO2 c.
N2 d.
He QUESTION (C.8)
[1.0)
Which ONE of the valve lineups listed below will result in sending a " rabbit" lNTO the core?
(Use drawing provided with handout.
OPEN SHUT a.
A&B C&D b.
C&D A&B c.
A&C B&D d.
B&D A&C i
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Section C Plant and Radiation Monitorina Svstems Page 18 QUESTION (C.9)
[1.0]
Which ONE of the following detectors is the most likely to detect a fuel element failure first?
a.
Continuous Air Monitor b.
Stack Gas Monitor c.
Stack Particulate Monitor d.
Primary Coolant Reactivity Monitor QUESTION (C.10)
[1.0)
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.11)
[1.0)
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" l
i
i Section C Plant and Radiation Monitorina Systems Page 19 QUESTION (C.12) [1.0)
Which ONE of the following is the neutron source utilized at OSTR?
i a.
Am
'Be b.
23 Pu
'Be i
c.
ziopo _ eBe d.
12'Sb
'Be l
l l
QUESTION (C.13) [1.0]
What is the purpose of the Cadmium Lined In-Core Irradiation Tube (CLICIT)
L f
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.
t QUESTION (C.14)
[1.0]
Which ONE of the following channels recorded by the BLUE pen on the console recorder, when the MODE switch is in the PULSE LO position?
a.
Fuel Temperature b.
Linear Channel l
c.
Power Range Monitor nv circuit i
d.
Safety Channel I
i
. -- - -- -. -. -.. - - _ _ ~.
l Section C Plant and Radiation Monitorina Systems Page 20 QUESTION (C.15) [1.0)
The ventilation system is designed to maintain reactor bay pressure slightly negative pressure l
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.
l.
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.
l l
QUESTION (C.16) [1.0)
Which ONE of the following components in the purification system is PRIMARILY responsible for maintaining the primary coolant system conductivity low.
l a.
The surface skimmer b.
The pre-demineralizer filter c.
The demineralizer d.
The post-demineralizer filter QUESTION (C.17) [1.0)
During a survey of the demineralizer % hour after shutdown, you note that the dose rate has increased by a factor of 10 over the previous day's reading. Is this normal or abnormal, and why?
a.
Normal, due to N'8 in the coolant.
b.
Abnormal, due to the concentration of H8 in the domineralizer.
l c.
Abnormal, due to fission products in the demineralizer.
d.
Normal, due to Ar* entrained in the coolant system.
k
.s
_=.... - - -
Section C Plant and Radiation Monitorina Systems Page 21 QUESTION (C.18) [1.0]
Which ONE of the following scrams is available during pulsing operations?
a.
Loss of Detector High Voltage b.
Safety Power Level c.
Percent Power Level d.
Wide Range Log Power Level QUESTION (C.19) [1.0)
Which ONE of the following interlocks is NOT required for steady-state operations?
a.
Wide Range Log Power Level Channel b.
Transient Rod Cylinder Air c.
Shim, Safety and Regulating Rod Drive Circuit d.
Transient Rod Cylinder Position.
i 1
i i
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l
. -.. -. -.. - -.. _.. -. - = - _.
Section A R Theorv. Thermo. and Facility Characteristics Page 22 ANSWER (A.1) a REFERENCE (A.1)
OSTR Training Manual Vol. Ill, p. 29 Initial SDM = 2.5$ = 2.5 x 0.007 = 0.0175 AK/K. K
= 1/(1+SDM) = 1/1.0175 = 0.983 CR,/CR = (1 - K.)/(1 - K.)~ (1 - K.) = 15/30 (1 - 0.983) = %( 0.0170 = 0.00850 2
K. = 1 - 0.00850 = 0.9915 SDM = (1 - 0.9915)/0.9915 = 0.00857 Experiment worth = 0.0086 -0.0189 = 0.0103 AK/K + 0.007 = $1.475 ANSWER (A.2)
C i
. REFERENCE (A.2)
OSTR Training Manual Vol. lil, Example on p. 41 (After prompt drop.)
ANSWER (A.3) a REFERENCE (A.3)
)
Glassione, S & Sesonske, A., Nuclear Reactor Engineering,3* Edition, Kreiger Publishing, j
Malabar, Florida,1991, Table 3.3, p.' 164 ANSWER (A.4) b REFERENCE (A.4)
Glasstone, S & Sesonske, A., Nuclear Reactor Engineering,3* Edition, Kreiger Publishing, Malabar, Florida,1991, Table 2.6, p.100 ANSWER (A.5) b REFERENCE (A.5)
OSTR Training Manual Vol.111, pp. 22 & 23 ANSWER (A.6) d REFERENCE (A.6)
OSTR Training Manual Vol. lil, p. 30 ANSWER (A.7) a REFERENCE (A.7)
OSTR Training Manual Vol. Ill, p.10 ANSWER (A.8) c 1
i REFERENCE (A.8) l OSTR Training Manual Vol. Ill, p. 30
~
Section A R Theorv. Thermo. and Facility Characteristics Page 23 ANSWER (A.9) b REFERENCE (A.9)
OSTR Training Manual Vol. Ill, p. 30 SDM = (1-4)/h = (1-0.85)/0.85 = 0.15/0.85 = 0.1765, or a reactivity worth (p) of -0.1765.
Adding + 0.176 reactivity will result in a SDM of 0.1765 - 0.1760 = 0.0005. K, = 1/(1+SDM) =
1/(1 + 0.0005) = 0.9995 ANSWER (A.10) l C
l REFERENCE (A.10) i OSTR Training Manual Vol. Ill, pp.16 and 66 - 68 Reactivity due to temperature increase: 150*C x -0.00005/*C = -0.0075. To compensate you must ADD +0.0075 worth of reactivity (move rods out) 0.075/0.0001 inch" = 7.5 inches i
ANSWER (A.11) d i
REFERENCE (A.11)
Q = thc bT
\\
p g _4gggallons x8 x60 m/nutes,1 BN 1W
, 3 4.p,
minute gallon hour
- F-Ibm 3.41 x 10* BTU l
l
= 0.966 Mw or 966 kilowatts
]
ANSWER (A.12) b REFERENCE (A.12)
]
l OSTR Training Manual Vol. Ill, Example on p. 41 (After prompt drop.)
ANSWER (A.13) c l
REFERENCE (A.13)
OSTR Training Manual Vol. Ill, p. 41 l
ANSWER (A.14) l c
i REFERENCE (A.14)
OSTR Training Manual Vol. Ill, p. 51
(
ANSWER (A.15) i DELETED PER FACILITY COMMENT REFERENCE (A.15)
OSTR Training Manual Vol.111, p. 28
Section A R Theory. Thermo. and Facility Characteristics Page 24 ANSWER.(A.16) d REFERENCE (A.16)
OSTR Training Manual Vol. Ill, pp. 84 - 87.
ANSWER (A.17) b REFERENCE (A.17)
OSTR Training Manual Vol. til, pp. 6 - 8.
ANSWER (A.18) e REFERENCE (A.18)
OSTR Training Manual Vol. Ill, p. 48 ANSWER (A.19) d Answer changed per facility comment REFERENCE (A.19)
OSTR Training Manual Vol.111, pp. 22 - 24.
ANSWER (A.20) b i
REFERENCE (A.20)
Glasstone, S & Sesonske, A., Nuclear Reactor Engineering,3d Edition, Kreiger Publishing, Malabar, Florida, 1991,6 6.25, p. 337 1
Section B Normal. Emeraency and Radioloaical Control Procedures Page 25 ANSWER (B.1) b REFERENCE (8.1)
% inch of lead reduces the radiation level by %. Adding another % inch gives a total of three
%-thicknesses. Final radiation level will be I = l x (%)* = 1000 mrem /hr x % = 125 mrem /hr o
I 1
ANSWER (B.2) a l
REFERENCE (B.2) l l
10 CFR 20.1003 Definitions l
l ANSWER (B.3) a, 2; b, 3; c, 3; d, 4 i
REFERENCE (8.3) 10 CFR 20.1003, Definitions l
l ANSWER (B.4) i a, Check; b, Test; c, Check; d, Cal.
REFERENCE (B.4)
OSTR Technical Specifications, 91.0 Definitions ANSWER (B.5) a, A; b, B; c, B; d, B or C Second correct answer added per facility comment.
REFERENCE (B.5)
OSTROP 18.0 918.4, Classification of Reactor Experiments p. IV.18.3.
ANSWER (B.6) c REFERENCE (B.6)
OSTROP 23, Crane Operation Procedures,9 23.3, Limitations on Crane Operation p. IV.23.4.
ANSWER (B.7) b or a Second correct answer added per facility comment.
REFERENCE (B.7) 10 CFR 20.1206 ANSWER (8.8) c I
REFERENCE (B.8)
Emergency Response Plan,9 3.3.2 Authorities and Responsibilities of Facility Emergency Personnel, p. 3-7.
- ANSWER (B.9) b e
REFERENCE (B.9) 2 100 mrem (x ft): = 5000 mrem /hr (2 ft)2 X8 = 5000/100 x 4 = 200 ft X = 14.14 feet = 15 feet
_ _. -.. _ _ -. _. _.. _ _. ~ _. _ _ _ _ _ _ _ _ - _ _. _. _ _ _ _ _ _ _ _
Section B Normal. Emeroenev and Radiolooical Control Procedures Page 26 ANSWER (B.10) b REFERENCE (8.10) 200 mr/hr + 15 mr = 40/3 = 13.333 hr ANSWER (B.11) d REFERENCE (B.11)
Emergency R esponse Plan, 6 4.0 Emergency Classification System ANSWER (B.12) l c
l l
REFERENCE (B.12)
L Technical Specification 3.7, pp.14 - 15.
ANSWER (B.13) l b
l l'
REFERENCE (B.13)
Emergency Operating Procedures 91.2 Automatic Scrams, p. IV.1.2 j
ANSWER (B.14)
- b j
i REFERENCE (B.14) l OSTROP 4, Reactor Operation Procedures, p. 9.
i l
ANSWER (B.15).
b REFERENCE (B.15)
OSTROP 1, Emergency Operating Procedures, $ 1.E, p.1 L-ANSWER (B.16) d REFERENCE (B.16)
OSTROP 1, 6, Administrative and Personnel Procedures, 5 VI.B.5, p. 24.
ANSWER (B.17) d Answer changed por facility comment.
l_
REFERENCE (B.17)
OSTR Technical Specifications, 9 2.1 Safety Limit - Fuel Element Temperature, p. 6 i
1 i
l
.. -. ~ -.. - - -. -...
l i
k t
Section C Plant and Radiation Monitorina Systems Page 27 ANSWER (C.1) a, N; b, A; c, B; d, B; e, B; f, N REFERENCE (C.1) i OSTROP 22.0 Emergency Power System, Figures 22.1, and 22.2 j
ANSWER (C.2) b l
REFERENCE (C.2)
OSTR Training Manual Vol. 2,9 lilA, page 13.
ANSWER (C.3) i c
1 l
REFERENCE - (C.3)
OSTR Training Manual Vol.1, page 106 ANSWER (C.4) b REFERENCE (C.4)
OSTR Training Manual Vol ll, Fig. 2.16 ANSWER (C.5) d' REFERENCE (C.5)
OSTR Training Manual Vol. I, pg.106 ANSWER (C.6).
d' REFERENCE (C.6)
OSTR Training Manual Vol 11, p. 9.
ANSWER (C.7) a REFERENCE (C.7)
OSTR Training Manual, Vol I p. 70.
ANSWER (C.8) c REFERENCE (C.8)
OSTR Training Manual, Vol I, fig.1.40 Standard Rabbit System Schematic, p. 71.
ANSWER (C.9) a
' REFERENCE (C.9)
Oregon State University TRIGA reactor Training Manual, Radiologica/ Protection, p. 21
Section C Plant and Radiation Monitorino Systems Page 28 ANSWER (C.10) b REFERENCE (C.10) l OSTR Training Manual, Vol 1, fig.1.26 TRIGA Control Rod Drive Mecht.nism, p. 48.
ANSWER (C.11) i b
REFERENCE (C.11)
OSTROP 22.0 Emergency Power System, Fig. 22.1 One-Line Schematic Power Distribution ANSWER (C.12) a l
REFERENCE (C.12)
OSTR Training Manual, Vol. I, p. 30 - 32 l
ANSWER (C.13) a REFERENCE (C.13)
OSTR Training Manual, Vol. I, p. 81.
1 ANSWER (C.14) c REFERENCE (C.14)
J OSTR Training Manual Vol. II, fig. 2.17, p. 36 ANSWER (C.15) c REFERENCE (C.15)
OSTR Training Manual, Vol. I p.148.
ANSWER (C.16) c REFERENCE (C.16)
OSTR Training Manual Vol. I, p.116.
ANSWER (C.17) c REFERENCE (C.17)
OSTR Training Manual, Vol. I, p.116 ANSWER (C.18) a REFERENCE (C.18)
OSTR Technical Specifications, TABLE 1, p.12
4 Section C Plant and Radiation Monitorina Svatems Page 29 ANSWER (C.19) d REFERENCE (C.19)
OSTR Technical Specifications, TABLE ll p.12.
4 i
9 f
I b
l i
l I
I I
t l
U. S. NUCLEAR REGULATORY COMMISSION NON-POWER INITIAL REACTOR LICENSE EXAMINATION j
FACILITY:
Oregon State University REACTOR TYPE:
TRIGA (Pulsing)
DATE ADMINISTERED:
1996/10/21
]
REGION:
IV CANDIDATE:
INSTRUCTIONS TO CANDIDATE.
Answers are to be written on the answer sheet provided. Attach the answer sheets to the examination. Points for each question are indicated in paren-theses for each question. A 70%
overall 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 J33.
A.
liEACTOR THEORY, THERMODYNAMICS AND FACILITY OPERATING CHARACTERISTICS 2_0.00 J33.
B.
NORMAL AND EMERGENCY OPERATING PROCEDURES AND RADIOLOGICAL CONTROLS 1
20.00 33.3 C.
PLANT AND RADIATION MONITORING SYSTEMS 60.00 TOTALS FINAL GRA DE All work done on this examination is my own. I have neither given nor received aid.
Candidate's Signature i
i l
l l
i
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.
l 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 gnly to facilitate legible reproductions.
5.
Print your name in the blank provided in the upper right-hand comer of the examination cover sheet and each answer sheet.
t 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 tuming 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.
l l
l 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.
t EQUATION SHEET l
= thc, hT = th hH = UA AT p"",(p-Q)2 l
2a(k)f 4
t' = 1 x 10 seconds ggg, f,
S
-p 1-K,,
A,, = 0.1 seconds ~1 CR,(1 -K,,,) = CR,(1 -K,,,)
CR,(-p,) = CR,(-p,)
A,,p 1 -K,,'
SUR = 26.06 M=
.S-p, 1 - K,,,
f 1
CR, p = p, j osmo y,
1-K,,
CR, SDM = (1 -K,y) t p_p,9 ee T=
P-(1 -p) p p-p 0-p
\\
K'" - K**'
Tf+
~0~P ap,
k,,,x K,,,
p A,p 0.693 (K,,- 1 )
7
=
A p=
K,,
DR =DR, e
- DR,d, = D R d,*
j a
DR = 6C/E(n).
DR - Rem, Ci-curies, g:
E - Mev, R - feet (p2-0)*
(P -0)*
i
- Peak, Peak, 1 Curie = 3.7 x 10 dis /sec 1 kg = 2.21 lbm 8
1 Horsepower = 2.54 x 10 BTU /hr 1 Mw = 3.41 x 10' BTU /hr 1 BTU = 778 ft-lbf
'F = 9/5 *C + 32 i
i l
1 gal (H O) = 8 lbm
- C = 5/9 (*F - 32) 2 l
l c, o 1.0 BTU /hr/lbm/*F c, = 1 cal /sec/gm/*C I
j Section A R Theory. Thermo & Fac. Ooeratina Characteristics Pigs 4 i
ANSWER SHEET A.1 abcd A.11 abcd A.2 abcd A.12 abcd A.3 abcd A.13 abcd l-l A.4 abcd A.14 abcd l
A.5 abcd A.15 abcd I
l A.6 abcd A.16 abcd I
l l
A.7 abcd A.17 abcd A.8 abcd A.18 abcd A.9 abcd A.19 abcd l
A.10 abcd A.20 abcd L
_. -. -.... ~ -.. - =. _.. ~. -.. -... - _...
M Section B Normal /Emera. Procedures & Rad Con Pega 5 ANSWER SHEET B.1 abcd B.6 abcd B.2 abcd-B.7 abcd B.3a 1 2 3 4 B.8 abcd
)
b 1234 B.9 abcd c 1234 B.10 abcd i
l d
1234 B.11 abcd
(
B.4a check test calibration B.12 abcd b check test calibration B.13 abcd c check test calibration B.14 abcd d check test calibration B.15 abcd B.Sa ABC B.16 abcd l.
l' b ABC B.17 abcd l
I L
l l
e Section C Plant and Radiation Monitorina Systems Page 6 ANSWER SHEET C.1a Sys-A Sys-D No-Power C.8 abcd b
Sys-A Sys-B No-Power C.9 abcd c
Sys-A Sys-B No-Power C.10 abcd 1
)
d Sys-A Sys-B No-Power C.11 abcd e
Sys-A Sys-B No-Power C.12 abcd f
Sys-A Sys-B No-Power C.13 abcd C.2 abcd C.14 abcd C.3 abcd C.15 abcd C.4 -
abcd C.16 abcd C.5 abcd C.17 abcd,_
C.6 abcd C.18 abcd i-i.
l C.7 abcd C.19 abcd e
i e
f.
I i
(
. Powsn A'
,'l ri ii
//
//
ff S wire w
/
s 5'
RADIO - CHEMISTRY REACTm h", -
LABORATORY CONTROL ROOM D-102 V COMMUNICATIONS
/
ROOM I '
RADIATION N
0-301 MONITOR p
RA881T n
i RECElWR/
p PNEUMATIC D
SYSTEM enkTRot if
,y n
n L,l L
.]
L A'
8 C
O a
-+-
REACTOR TOP AREA
/
N
/
\\
A0lATION air INTAKE FOR INJECT MONITOR l
AIR INTAKE
~
FOR EJECT p
FILTER REACTOR
[V]l MECHANICAL EQUIPMENT J'
t stoWER ROOM REACTOR CORE EXHAUST TO O-1o6 MAIN STACK Fig. 1.40--Standard Rabbit System Schematic
- _ _ _ _ _ _ - _ _ _ _ _