ML20135C196

NRC Operator Licensing Exam Rept 50-083/OL-97-01 Including Completed & Graded Tests for Tests Administered on 970210-11.Candidate Passed Exam

ML20135C196
Person / Time
Site:	05000083
Issue date:	02/24/1997
From:	Isaac P NRC (Affiliation Not Assigned)
To:
Shared Package
ML20135C197	List: ML20135C196 ML20135C342
References
50-083-OL-97-01, 50-83-OL-97-1, NUDOCS 9703030411
Download: ML20135C196 (42)

Text

~

.- l

U. S. NUCLEAR REGULATORY COMMISSION i OPERATOR LICENSING INITIAL EXAMINATION REPORT -

REPORT NO.: 50-083/OL-97 r ,

' FACILITY DOCKET NO.: 50-083 FACILITY LICENSE NO.: R-56 .

FACILITY: University of Florida  !

EXAMINATION DATES: February 10 - 11,1997 EXAMINER: Patric Isaac, Chief Examiner  !

/

. SUBMITTED BY: '

.- 77 strick Isaaf, Chief Examiner / Da(e 1

SUMMARY

During the week of February 10,1997, the NRC administered Operator Licensing Examinations to one Senior Reactor Operator Instant (SROI) candidate. The candidate passed the i examination.

REPORT DETAILS '  !

.1. Examiners: Patrick Isaac, Chief Examiner

2. Results:

RO PASS / Fall SRO PASS / FAIL TOTAL PASS / FAIL Written l 0/0 1/0 1/0 Operating Tests 0/0 1/0 1/0 -.

Overall 0/0 1/0 110

3. Exit Meeting: #

l Dr. William Vernetson, Director of Nuclear Facilities L Daniel Cronin, Reactor Manager

' Patrick Isaac, NRC, Chief Examiner l The facility examination comments were discussed as noted in Enclosure 2. There were no generic concerns raised by the Chief Examiner.  ;

i ENCLOSURE 1 9703030411 970225 PDR Y ADOCK 05000083 png i

NRC RESOLUTIONS - WRITTEN EXAMINATION QUESTION (A.itl)

The shutdown margin (SDM), upon fullinsertion of all control rods following a reactor scram from full power, is the SDM immediately prior to the scram.

a. Equal to

b. Less than

c. Greater than

d. Indeptadent of Answer: a Facility Comment: ,

The answer key indicates the correct answer to A.18 to be (a); however, we feel the correct answer should be (c) due to the manner in which the facility defines shutdown margin and the manner in wh~n it is applied. The facility makes the distinction between the terms shutdown margin (SDM) and available shutdown margin (ASDM). The full description of the mar.ner in which the trainees and operators are trained on shutdown margin and available shutdown margin is contained in Attachment 1. Additionally, the equation provided on the examination formula sheet for shutdown margin is SDM = (1-k,)/k,. Answer (a) would not be a correct response on the basis of this equation alone because at the critical position, k, would be one and SDM would be zero; at some rod position below critical, the k, would be less than one and the SDM would be greater than zero; therefore the SDM at full insertion would be greater than the SDM at critical.

I NRC Resolution:

Comment Accepted. The answer key will be modified to accept (c) as the correct answer.

ENCLOSURE 2 QUESTION fB.15)

Who (by title) is the lowest level of operations staff who may authorize switching between City Water and Well Water positions.

a. Licensed Reactor Operator at console

b. Licensed Senior Operator at console

c. . Senior Operator on-call

d. The Reactor Manager (or designated alternate).

Answer: c Facility c7 ament:

The answer key indicates the correct answer to B.15 to be (b); however, we feel the correct answer should be either (b) or (c) because our Standard Operating Procedures (SOPS) state i that the authorizing individual is the " designated SRO." The " designated SRO" could be the SRO j on-call or, if the SRO on-call is also the console operator, as frequently occurs, the SRO at the l

)

console.

i NRC Resolution: l Comment accepted. The answer key will be modified to accept both (b) and (c) as correct.. l l

b i

1 i

1 2-

-. . ._ . . - . .~.. - _._-..-...- - ..-.- -_-.-. -.. ._ - - - __ .

i  !

QUESTION (C.01) i

' I During operations you notice that the red and orange lights associated with the secondary l 1

system are extinguished, while the white light is energized. What is the status of the secondary  !

system? (Assume alllights are operable.) l

i

a. The system is secured.

l

b. The system is operating with flow less than 60 gpm. I t

i c. The system is operating with flow greater than 60 gpm but less than 140 gpm. ,

i j d. The system is operating with flow greater than 140 gpm. l

) Answer: c j Facility Comment:

The answer key indicates the correct answer to C. I to be (c); however, we feel the question is l poorly worded and should be discarded from the examination. The conditions indicated in the i

question are confusing and probably not possible as worded. The question makes references to

the " red", " orange", and " white" lights. Our secondary system has three different white lights j associated with it. The condition of the " red" light will be determined by which " white" light is

energized. Additionally, two of the " white" lights will also affect the third " white" light. Finally, the l j secondary flow scram mode in effect will also determine whether or not any or all of the lights are  !

i on or off. ,

2 i

j NRC Resolution: ,

r

Comment accepted. This question will be deleted from the examination.  ;

I iL k

a a

i

'i 4

3 1

1 4

6

~

QUESTION (C.04) I i

On one of the area monitors, the green light is lit, the red light is extinguished, and the amber l light is on. Which ONE of the following conditions is possible? ,

a. The detector is saturated.

b. The monitor is reading a level of 2 mr/hr.

c. The monitor is reading 9 mr/hr. ,

d. The monitor is reading 15 mr/hr.

Answer: d 1 Facility Comment: I

- The answer key indicates the correct answer to C 4 to be (d); however, we feel the correct answer should be (c). The question does not distinguish between the setting required by the j Technical Specifications and the more conservative setting required by the SOPS. The Technical  !

specifications require a " Trip 2" (amber light) at 5 mR/hr and a " Trip 2"(red light) at 25 mR/hr. 1 However, the SOPS require a " Trip 2" (amber light) at 2.5 mR/hr and a " Trip 1" (red light) at 10 i mR/hr. Because of this more conservative setting and the fact that the question does not specify  !

whether the SOPS or Technical Specifications are being used as the basis for the question, the i only possible reading on the monitor given the conditions would be (c) 9 mR/hr. .j i

NRC Resolution:

Comment accepted. The answer key has been corrected to accept "c" as correct.

I

i i

l OPERATOR LICENSING EXAMINATION l ~

! gg REGug

} O

\

\

f. ,,VW m[g\ ,

O .

/

O

+++++

i i.

UNIVERSITY OF FLORIDA

! February 10,1997 j ENN_CJ.OSURE 3

U. S. NUCLEAR REGULATORY COMMISSION NON-POWER INITIAL REACTOR LICENSE EXAMINATION FACILITY: University of Florida REAl r OR TYPE: ARGONAUT DATd ADMINISTERED: 1997/02/10 REGION: 11 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 ir. brackets for each question. A 70% in each section is required to pass the examination. Examinations will be picked up three (3) hours after the examination starts.

% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 20.00 33.3 A. REACTOR THEORY, THERMODYNAMICS AND FACILITY OPERATING CHARACTERISTICS 20.00 33.3 B. NORMAL AND EMERGENCY OPERATING PROCEDURES AND RADIOLOGICAL CONTROLS 20.00 33.3 C. PLANT AND RADIATION MONITORING SYSTEMS 60.00  % TOTALS FINAL GRADE

.All work done on this examination is my own. I have neither given nor received aid.  ;

Candidate's Signature  ;

S

i .

l l

Section A R Theory. Thermo & Fac. Ooeratina Characteristics Pegs 4 ANSWER SHEET i A.1 abcd A.11 abcd i

l A.2 abcd A.12 abcd

! A.3 abcd A.13 abcd-  !

A.4 abcd A.14.- abcd A.5 abcd A.15 abcd l 1

A.6 abcd A.16 abcd l A.7 abcd A.17 abcd A.8 abcd A.18 abcd A.9 abcd A.19 abcd A.10 abcd A.20 abcd i

i I

l

Section B Normal /Emero. Procedures & Rad Con Paga 5 ANSWER SHEET ~ '

i B.1 abcd B.12a check test cal i

B.2 abcd B.12b check test cal ,

i B.3 abcd B.12c check test cal B.4 abcd B.12d check test cal - ,

B.5 -abcd B.13 abcd i B.6 - abcd B.14 abcd B.7 abcd B.15 abcd 8.8 abcd B.16 abcd B.9 abcd B,17 abcd B.10 abcd B.18 abcd B.11 abcd B.19 abcd

~

1 Section C Plant and Radiation Monitorina Systems pig 3 6 .

ANSWER SHEET C.1 abcd C.9 abcd  !

C.2 abcd C.10 abcd 1

, C.3 a FULL ROD-DROP C.11 abcd l i

b FULL ROD-DROP C.12 abcd l

. i

. c FULL ROD-DROP C.13 abcd l.

t l- d FULL ROD-DROP C.14 abcd l

l 4

e FULL ROD-DROP C.15 abcd

• i

i.  :

l i f FULL ROD-DROP C.16a 1 2 3 4 5 t i

C.4' abcd b12345  :

l C.5 ' abcd c12345 C.6 abcd d12345  !

1

i. C.7 abcd e12345 I i

I 1

. C.8 abcd C.17 abcd i i

C.18 abcd t

d 4

EQUATION SHEET -

Q = acpb T = s bH = UA b T.. = ^ (p-S) 2 i' P"'"

2ar (k) (

l' = 5 x 10 ~5 seconds S S SCR - ~ "

-p K ,ff  !

h " = 0.1 seconds '*

CR3 ( 1 -K,,f,) = CR2 I 1 -Eerr,)

CR 3( -p 1 ) = CR 2I ~P 2)

_,S = 0. 001 i

I; 1 - K ,ff*

M= '

A *ff p "

S UR = 2 6. 0 6 1-K.tr 2 '

S-p P=P n10 '"' W 1 CR 3 M= =

1 -K,f f CR 2 e

P=P g e*-.

( 1 -K,f f ) i SDM = '

"arr S (1 -p) .

P= P*

( S-p T= __

p-S -

T= + 'O K,f f* -Kerr

• P A err P bp =

k,ff,x K,ff

( K,f f-1) I p=

T4 - 0.693 K,,,

h DR,d, = DR2d2 2 DR = DR, e ' *

• 2

_ 2 Peak, Peak, I DR = 6CIE(n)

R*

I = I,e~"

1 Curie = 3.7 x 10 dis /sec 1 kg = 2.21 lbm -

1 BTU = 778 ft-lbf "F = 9/5 C + 32

~

l gal (H 2O) = 8 lbm *C = 5/9 ( F - 32) cp = 1.0 BTUIhr/lbml F c, = 1 callsec/gml C

l NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:

l 4

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 p_n_ly n to facilitate legible reproductions.

5. Print your name in the blank provided in the upper right-hand corner of the examination cover sheet and each answer sheet.

6. Mark your answers on the answer sheet provided. USE ONLY THE PAPER PROVIDED AND DO NOT WRITE ON THE BACK SIDE OF THE PAGE.

7. The point value for each question is indicated in (brackets) after the question.

8. If the intent of a question is unclear, ask questions of the examiner only.

9. When turning in your examination, assemble the completed examination with examination questions, examination aids and answer sheets. In addition turn in all scrap paper.

10. Ensure allinformation 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.

I Section A R Theory. Thermo. and Facility Characteristics Page 2 QUESTION (A.1)-[1.0] >

Which ONE of the four factors listed below is the MOST affected by an increase in poison level in the reactor?

a. ~ Fast Fission Factor (c)

b. Fast Non-Leakage Probability (%)

f

c. Thermal Utilization Factor (f) I

d. Reproduction Factor (r))

QUESTION (A.2) [1.0]

Given a control rod worth of 0.1% AK/K/ inch and an cx r of 0.05% AK/K/*F. If temperature i INCREASES by 9'F, how much and in what directior, will the control rod move?

a. 4% inches inward i

b. 4% inches outward

c. 9 inches inward ,

d. 9 inches outward j r

QUESTION (A.3) [1.0]

During startup, you withdraw rods an equal amount (distance). As the reactor approaches  ;

criticality, which ONE of the following statements best describes reactor behavior. (Assume the i reactor remains slightly suberitical.) i

a. Each rod withdrawal will add the same amount of reactivity, i

i

b. Reactor power will increase by the same amount for each rod withdrawal. i

c. The time for power to stabilize will increase for each succeeding rod withdrawal. l

d. Decreasing time between withdrawals will result in a lower critical rod height. l i

t l

l l

l 1

Section A R Theorv. Thermo. and Facility Characteristira Page 3 QUESTION (A.4) [1.0]

Which ONE of the following times would you expect to have the MAXIMUM amount of xenon in the core? (Assume initial condition was in effect for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to 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"4.

QUESTION (A.5) (1.0].

A FAST neutron will lose the MOST energy per collision when interacting with the nucleus of which ONE of the following elements?

a. H'

b. H

c. C'

d. U23e QUESTION (A.6) [1.0]

The reactor had a shutdown margin of 2.5$, and a source range count rate of 15 counts per minute. After placing samples in the reactor the count rate increased to 30 counts per minute.

What is the worth of the sample?

a, =-90p

b. = +90p

c. = -1.25$

d. = +1.25$

Section A R Theorv. Thermo. and Facility Characteristics Page 4 QUESTION (A.7) [1.0)

A THERMAL neutron has the LEAST probability of being absorbed by which ONE of the following elements?

a. H'

b. H

c. C'

d. U238 QUESTION (A.8) [1.0)

D for U235 is 0.0065. D., for the Univ. of Florida ARGONAUT reactor is 0.007. Why is D , larger?

a. The reactor contains U2 8 which has a larger p for fast fission than U235

b. The reactor contains Pu23e which has a larger p for thermal fission than U235,

c. Delayed neutrons are born at a higher average energy than fission neutrons resulting in a greater amount of fast fissioning.

d. De!ayed neutrons are born at a lower average energy than fission neutrons resulting in fewer being lost to fast leakage.

QUESTION (A.9) [1.0)

A few minutes following a scram you note that reactor period is stable, and power level reads 3 x 105 counts. What reading would you expect to see three minutes later?

a. 1 05

b. 3 x 104

c. 1.5 x 10*

d. 10'

)

Section A R Theory. Thermo. and Facility Characteristics Page 5 -

QUESTION (A.10) [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 .

1

a. is impossible as there would be no neutrons available to startup the reactor.

b. wou!d be very slow due to the long time to build up neutron population from so low a level. ,

c. could result in a very short period due to the reactor going critical before neutron l population builds up enough to be read on nuclear instrumentation.

d. cari be compensated for by adjusting the compensating voltage of the startup detector. ,

QUESTION (A.11) [10]

You remove the source from the reactor. Later you reinstall the source and note that reactor power is increasing LINEARLY (i.e. counts increase by 60 count each minute). What was the condition of the reactor just prior to inserting the source? (Assume the source has no reactivity worth, and there are no other changes effecting reactivity). The reactor was .

a. Very subcritical  !

b. slightly suberitical

c. exactly critical

d. slightly supercritical QUESTION (A.12) [1.0)  :

Which ONE of the following is the MAJOR source of 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 A R Theory. Thermo. and Facility Characteristics Page 6 QUESTION (A.13) [1.0]

The term PROMPTJUMP refers to ..

a. the instantaneous change in power due to withdrawal of a control rod.

b. a reactor which has attained criticality on prompt neutrons alone.  ;

c. a reactor which is critical on both prompt and delayed neutrons.

d. a negative reactivity insertion which is less than L.

QUESTION (A.14) [1.0] -

Which ONE of the following evolutions will take the LONGEST time to occur? A reactor power change of ..

a. 5% of rated power, going from 1% to 6% of rated power.

b. 10% of rated power, going from 10% to 20% of rated power.

c. 15% of rated power, going from 20% to 35% of rated power. ,

d. 20% of rated power, going from 40% to 60% of rated power.  ;

QUESTION (A.15) [1.0)

As primary coolant temperature increases, control rod worth:  !

a. increases due to higher reflector efficiency.

b. decreases due to higher neutron absorption in the moderator. ,

c. increases due to the increase in thermal diffusion length. ,

d. remains the same due to constant poison cross-section of the control rods.

t

l I

Section A R Theorv.15ermo. and Facility Characteristics Page 7 QUESTION (A.16) [1.0]

An initial count rate of 100 is doubled five times during startup. Assuming an initial K,=0.950,  :

what is the new K.,7 I

a. 0.957 l 1

b. 0.979 I

c. 0.988 )

l

d. 0.998 QUESTION (A.17) [1.0]

Which ONE of the following describes the MAJOR contribution to the production and depletion of xenon in the reactor? 1

a. Produced from radioactive decay of iodine and depletes by neutron absorption only l

b. Produced from radioactive decay of iodine and depletes by radioactive decay and neutron absorption i

c. Produced directly from fission and depletes by neutron absorption only

d. Produced directly from fission and depletes by radioactive decay and neutron absorption QUESTION (A.18) [1.0)

The shutdown margin (SDM), upon fullinsertion of all control rods following a reactor scram from full power, is the SDM immediately prior to the scram.

a. Equal to

b. Less than

c. Greater than

d. Independent of r

l

Section A R Theory. Thermo. and Facility Characteristics Page 8 QUESTION (A.19) [1.0]

Which one of the following statements is FALSE?

a. The value of an isotope's neutron absorption cross section is independent of a neutron's energy.

b. A U-235 atom can be fissioned by a " fast" neutron.

c. A U-238 atom is less likely to have a " thermal" fission, than a Pu-239 atom,

d. Approximately 210 MeV is released per fission event.

QUESTION (A.20) [1.0]

Which ONE of the following is the reason that reactor indicated power (count rate) stabilizes several hours after a reactor trip? Assume allinstrumentation is operable, and no reactivity changes.

1

a. Subcritical multiplication of source neutrons.

b. Continuing decay of the longest lived delayed neutron precursor.

c. Neutron level dropping below detection threshold, the detector reading is due to a test signal input from Nuclear Instrumentation.

d. Gamma radiation due to decay of fission products below detection threshold, the detector reading is due to a test signal input from Nuclear Instrumentation.

I i

l l

l

. i l

l Section B Normal. Emeraency and Radioloaical Control Procedures Page 9 QUESTION (B.1) [1.0]

A radiation worker works in a room with an average radiation dose of 10 mR/hr, for four hours a day. Which ONE of the following is the MAXIMUM number of days the worker may work without exceeding his 10 CFR 20 limits?

i

a. 250 days ,

d

b. 125 days  !

l

c. 25 days - )

d. 12% days

QUESTION (B.2) [1.0]

Which ONE of the following is the MINIMUM amount of time, according to 10 CFR 55.53.e, that

, a licensed operator must perform his/her licensed duties to maintain proficiency?

a. four hours per calendar month

b. eight hours per calendar month f

c. four hours per calendar quarter

d. eight hours per calendar quarter i

QUESTION (B.3) [1.0)

Per Technical Specifications, a system or component is defined as OPERABLE if . i

a. a channel check has been performed.

b. a functiona! check has been performed.

c. it is capable of performing its intended function.

d. it has no outstanding testing requirements.

I i

- __ - --_ _ .- . . - - . . ... . .. . .. =.- . . - .

4

, Section B Normal. Emeroency and Radioloaical Control Procedures Page 10 QUESTION (B.4) [1.0]

Assuming equal amounts of energy absorbed (Rads), which ONE of the listed types of radiation will have the greatest effect on the body? l

a. Beta

, b. Gamma i ,

c. Alpha j d. Thermal Neutrons QUESTION (B.5) [1.0]

Two point sources emit gamma radiation at the same curie strength. Source A's gammas have an average energy of 1 Mev, while Source B's gammas have an average energy of 2 Mev. You obtain readings for each source using a Geiger Muller Detector at 10 feet with no shielding. '

Which one of the following is the expected relative reading levels for the two sources?

m 1 a. Source B will read four times as high as source A. i

b. Source B will read twice as high as source A. i i

, c. Both sources will read the same. '

d. Source B will read half as high as source A.

l.

t QUESTION (B.6) [1.0] -

A four inch thick steel plate reduces gamma radiation dose from 60 mrem /hr to 6 mrem /hr.

What is the expected dose (at the same distance) if you add another 1 inch of steel plate?

a. 0.56 mrem /hr

b. 1.50 mrem /hr

! c. 2.62 mrem /hr d.- 3.37 mrem /hr 1'

N i

1 l

l i

Section B Normal. Emeraency and Radioloaical Control Procedures Page 11 l QUESTION (B.7) [1.0)

A irradiation sample of Sodium (t.3 = 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />) reads 4.5 R/hr at one meter. How long must the experimenter wait before the sample will read 1 R/hr at one meter?

a. 30 minutes

b. 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> '

c. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

d. 33 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br /> l QUESTION (B.8) [1.0)  :

Crystal River has an accident. You receive 5 REM performing volunteer work in the accident l recovery phase. How is this radiation tracked? l

a. It is tracked at University of Florida as part of your normally allowed 5 REM / year for a i radiation worker. i

b. It is tracked at University of Florida as part of your lifetime Planned Special Exposure l Limit (5 REM / year,15 REM / lifetime). )

c. As an emergency dose, it only tracked at the accident site, it is not tracked at University l of Florida.

d. As an emergency dose, it is only tracked by the NRC, it is not tracked at the University )

of Florida.  !

QUESTION (B.9) [1.0)

Who (by title) may authorize personnel to receive doses in excess of their normal occupational limits?

• I

a. Emergency Director

b. Rad:ation Control Officer

c. Emergency Director with Radiation Control Officer Concurrence

d. Emergency Coordinator with Radiation Control Officer Concurrence I

Section B Normal. Emeroency and Radiolooical Control Procedures Page 12 QUESTION (B.10) [1.0]

As a result of a failed experiment, two out of three of the area radiation monitors are reading 150 mR/hr. What emergency class levelis this?

a. Event less severe than the Lowest Category.

b. Notification of Unusual Event

c. Alert

d. Site Emergency QUESTION (B.11) [1.0]

The reactor trips due to a loss of power. What is the lowest level of management who may authorize restart?

a. Any licensed Reactor Operator.

b. The console operator if licensed as a Senior Reactor Operator.

c. The Senior Reactor Operator on-call.

d. The Reactor Manager (or designated alternate).

QUESTION (B.12) [2.0)

Identify each of the following as either a Channel Check, a Channel Test or a Channel Calibration, as defined by Technical Specifications.

a. Observe overlap between the startup channel and the intermediate range of Nuclear Instrumentation.

b. Replace a resistance temperature detector (RTD) with a precision resistance bridge to check proper circuit operation.

c. Monitor nuclear instrumentation verifying proper shutdown period indication.

d. Based on a heat balance (calorimetric) performed on the primary system, adjust Nuclear instrumentation.

Section B Normal. Emeraency and Radioloaical Control Procedures Page 13 QUESTION (B.13) [1.0]

Which ONE of the following conditions is allowable during reactor operations?

a. Only one air particulate monitor (APD) capable of audibly warning personnel of radioactive particulate airborne contamination in the cell atmosphere.

b. Failure of the fixed stack monitor recorder in the control room.

c. Only one gamma area monitors capable of audibly alarming on high radiation levelin the control room.

d. Failure of the building evacuation alarm.

QUESTION (B.14) [1.0]

What is the radiation dose limit (on contact) for a sample in the rabbit receiver above which Reactor Manager permission is required for renioval?

a. 100 mR/hr

b. 200 mR/hr

c. 500 mR/hr

d. 1000 mR/hr QUESTION (B.15) [1.0]

Who (by title) is the lowest level of operations staff who may authorize switching between City l Water and Well Water positions. i

a. Licensed Reactor Operator at console

b. Licensed Senior Operator at console

c. Senior Operator on-call

d. The Reactor Manager (or designated alternate).

1

Section B Normal. Emeraency and Radiolooical Control Procedures Page 14 QUESTION (8.16) [1.0]

An accessible area with a radiation level of 50 mR/hr should be posted as a:

a. restricted area

b. radiation area

c. high radiation area

d. very high radiation area QUESTION (B.17) [1,0]

Which ONE of the following conditions is the minimum to be met in order to be the Second Person during reactor operations?

a. Read and understand the Emergency Plan.

b. Read and understand the Emergency Procedure-Radiological and Emergencv Procedure-Fire.

c. Read and understand the Emergency Plan, and sign Emergency Plan Qualification  :

Form.

d. Read and understand the Emergency Procedure-Radiological and Emergency Procedure-Fire, and sign Emergency Procedure Qualification Form SOP-B.1A.

QUESTION (B.18) [1.0)

Which ONE of the following is the lowest level of staff who may operate the 3-ton bridge crane during reactor operations?

a. Anyone Nuclear Engineering Student.

b. Any Univ. of Florida Maintenance Technician.

c. Certified Second Person.

d. Licensed Reactor Operator.

l l

I" a

i Section B Normal. Emeroency and Radiolooical Control Procedures

)

Page 15  ;

QUESTION (B.19) [1.0) ,

j A daily startup was completed at 8:15 am, Monday February 10,1997. This startup checklist is i i valid as long as the reactor is started up by ...  !

l l

a. 2

15 pm,2/10/97 l

b. 4:15 pm,2/10/97 l 1

l

c. 8:15 pm, 2/10/97  :

=,

d. 8:15 am,2/11/97 l l

l j

I I

l i

i

)

l i

i I

j i

I

Section C Plant and Radiation Monitorina Systems Page 16 QUESTION (C.1) [1.0] DELETED During operations you notice that the red and orange lights associated with the secondary system are extinguished, while the white light is energized. What is the status of the secondary system? (Assume alllights are operable.)

a. The system is secured.

b. The system is operating with flow less than 60 gpm.

c. The system is operating with flow greater than 60 gpm but less an 140 gpm.

d. The system is operating with flow greater than 140 gpm.

QUESTION (C.2) [1.0]

Which ONE of the listed gases is used as the propellant for the rabbit system?

a. Air

b. Nitrogen

c. CO 2

d. He QUESTION (C.3) [2.0, 0.33 each]

Identify each of the Reactor Trips in Column A, with the correct type of trip FULL or ROD-DROP.

a. Reactor Period less than 3 sec.

b. Reactor Power at 125% of full power.

c. Loss of power to core vent system.

d. Manual scram Bar

e. Loss of chamber high voltage

f. Loss of Secondary Flow

4 i

t a

Section C Plant and Radiation Monitorino Systems Page 17 4 i

QUESTION (C.4) [1.0) _

On one of the area monitors, the green light is lit, the red light is extinguished, and the amber i

! light is on. Which ONE of the following conditions is ponible?

s- ,

1

a. The detector is saturated l

b. The monitor is reading a level of 2 mr/hr. j I c. The monitor is reading 9 mr/hr.

l' d. The monitor is reading 15 mr/hr.

l. QUESTION (C.5) [1.0) i How is primary coolant flow rate adjusted? .

a. Throttle valve on inlet of primary pump.

) b.- Throttle valve on outlet of primary pump.

c. Speed adjust of primary pump.

I

d. Not adjustable, cooling rate controlled by secondary flow rate.

t

.t

. QUESTION (C.6) (1.0)

! Which ONE of the following conditions will NOT cause the Rupture Disk to break?

l

a. Stop and Start Primary Pump.

b. Shut Dump Valve before system completely drained.

- c. Rod-Drop.

d. Steam Production.

4 j

1 l

4 1

d l

'* *--- *- w - ' --

7

Section C Plant and Radiation Monitorina Systems Page 18 QUESTION (C.7) [1.0]

A warning light will energize if secondary flow drops to less than ,

a. 180 gpm

b. 160 gpm

c. 140 gpm

d. 120 gpm QUESTION (C.8) [1.0]

Which ONE of the following conditions is required to clear the secondary flow scram? Power level must drop below .

a. 1000 watts b 700 watts

c. 100 watts "

d. 1 watt QUESTION (C.9) [1.0)

What is the purpose of the Spent fuel pit which DOES NOT have a special controllock?

a. Storage of removable Pu-Be neutron source,

b. Storage of New (Unirradiated Fuel).

c. Storage of Hurricane Rods.

d. No storage allowed, hole used to check water level.

._ . _ . _ . _ . . . . ._ _ _ _ . . _ . _ - _ - - _ - ~ . - _ . - . - .-.---m.

i Section C Plant and Radiation Monitorino Systems Page 19 QUESTION (C.10) [1.0)  ;

The Secondary Cooling System contains two check valves with a drain between. What is the  ;

purpose of this valve combination. l 3

a. To prevent backflow from the city water supply to the Well Water tank. [

b. To prevent backflow from the secondary system to the city water supply.  ;

c. To prevent inadvertent initiation of City Water secondary flow.

i d. To enhance initiation of Well Water secondary flow.

[ QUESTION (C.11) [1.0]

Which ONE of the following is the material the cladding is composed of?

j a. Zircalloy II

b. Zircalloy IV

c. Stainless Steel d Aluminum

QUESTION (C.12) [1.0]

Which ONE of the following is NOT a purpose of the Shield Tank?

a. Experimental Port i

i b. Cooling for experiments

c. Nr4 Shielding of core

d. Vent mechanism for contaminated samples.

(

1

]

l l

Sagtjon C Plant and Radiation Monitorina Systems Page 20 I l

1 QUESTION (C.13) [1.0] i The maximum thermal neutron flux in the UFTR is .

]

a. 1.8 x 10 S n/cmr/sec

b. 1.8 x 10" n/cm 2/sec

c. 1.8 x 10'2 n/cm2/sec

d. 1.8 x 10 n/cm 2/sec QUESTION (C.14) [1.0)

Which of the listed materials (along with reason)is most of the primary system constructed of?

a. Stainless Steel, low activation properties

b. Aluminum, low activation properties  :

c. Stainless Steel, short half-life

d. Aluminum, short half-life  ;

QUESTION (C.15) [1.0]

Why is each control blade clutch light depressed following a reactor shutdown, prior to -

removing the console magnet key?

a. To assure the reactor protection system lower limit switch settings are initiated.

b. To prevent control blade ejection from the core.

I

c. To assure the control blade motors are deenergized. I

d. To prevent a spurious period trip. j i

)

1

{ ,;

l Section C ' Plant and Radiation Monitorina Systems Page 21

! - 'l E QUESTION (C.16) [2.0,0.4 each]

Match the method for gamma compensation given in column b for the detector (s) (and modes) i- listed in column B. Note each item in column A will have only one answer, methods in column B may be used more than once.

Column A Column B

a. Fission Chamber & B-10 (WR) 1. Intrinsic

.- b. Fission Chamber in lon Chamber Mode (WR) 2. None l c. Fission Chamber in Cambelling Mode (WR) Active Gamma Comp.

l_ (Summing Inversion)

d. Uncompensated lon Chamber 4. Pulse Height Discrimination -

e. Col..;cnsated lon Chamber 5. Photomultiplication i

j QUESTION (C.17) [1.0]

} The upper limit switch, period interlock and multiple blade interlock are not in effect for ..

i.

a. any control blade when two clutch lamps are burned out.

b. the regulating blade when power level is less than 1 kW.

c. the regulating blade when power level control is in automatic.

d. any control blade when all control blade down lights are on.

QUESTION (C.18) [1.0)

Which ONF, of the following is NOT a Rod inhibit?

a. 10% drop in the value of the HV power supply to the detector for Ni safety channel #2.

b. 1 cps on the NI Wide Range channels

c. Calibrate switch out of normal for Ni safety channel #2..

d. At 10 seconds period on the Wide Range Channels.

~ ~ --

Page 22 Section A R Theorv. Thermo. and Facility Characteris.ti.cs ANSWER (A.1) c Wesley Publishing, Reading,

/ REFERENCE (A.1)Lamarsh, J.R., /ntroduction 33 p 313to Nuclear

- 3.18. Engineering, Addison I 1983. @ 7.2, p. 300.

Massachusetts, Burn, R., Introduction to Nuclear Reactor Operations, @ 1988, @ . , p . -

ANSWER (A.2) b REFERENCE (A.2) 9"F x 0.0005 0"

0.001 K-inch ANSWER (A.3) c Publishing, Malabar, REFERENCE (A.3)

Glasstone, S. And Sesonske, A, Nuclear5,Reactor pp. 5 5-28 Engineering, Kreige Florida, 1991, @ 3.161 - 3.163, pp.190 - 191.

Burn, R., Introduction to Nuclear Reactor Operations, @ 1988, Chapt.

ANSWER (A.4) d i Publishing, Malabar, REFERENCE (A.4)

Glasstone, S. And Sesonske, A Nuclear81Reactor Engineering, Kre ger 8 4 pp. 8 8-14.

Florida, 1991, @@ 5.56 - 5.80, pp. 250 - 260. . ,

Burn, R., Introduction to Nuclear Reactor Operations, @ 1988, f@ .

ANSWER (A.5) a Publishing, Malabar, REFERENCE (A.5)

Glasstone, S. And Sesonske, A. Nuclear Reactor Engineering, Kreig Florida,1991, @

Burn, R., Introduction to Nuclear Reactor Operations, @ 1988, @

l 1

1 1 1 Section A R Theorv. Thermo. and Facility Characteristics _ Page 23 I

! ANSWER (A.6) d  ;

REFERENCE (A.6)  :

$2.5 = 0.0175 AK/K. K, = 1/(1.0175) = 0.9828  ;

l 1-Ke2 = (1 - K,5) x CR,/CR2 - Km = 1 -[(1 - K,,)CR,/CR 2 "

Ke2 = 1 -[(1 - 0.9828)15/30] = 1 -[ 0.0172 x 0.5 ] = 1 - 0.0086 = 0.9914 p = (0.9914 - 0.9828)/(0.9914 x 0.9828) = 0.008826 = $1.26

' l i

ANSWER (A.7) i b

l REFERENCE (A.7)

REFERENCE:

l Glasstone, S. And Sesonske, A, Nuclear Reactor Engineering, Kreiger Publishing, Malabar, l Florida,1991, 9 2.108 - 2.116, pp. 77 - 81 ,

Lamarsh, J.R., introduction to Nuclear Engineering, Addison - Wesley Publishing, Reading,. '

3 Massachusetts,1983. 3.2, pp. 45 - 51.

i Burn, R., Introduction to Nuclear Reactor Operations, b 1988, @ 2.5, pp. 2 2.44 i

ANSWER (A.8) d

REFERENCE (A.8)

Burn, R., Introduction to Nuclear Reactor Operations, @ 1988, G ANSWER (A.9) i j b REFERENCE (A.9)

Glasstone, S. And Sesonske, A, Nuclear Reactor Engineering, Kreiger Publishing, Malabar, Florida, 1991,5 5.47, p. 246 Lamarsh, J.R., Introduction to Nuclear Engineering, Addison - Wesley Publishing, Reading, Massachusetts, 1983 9 7.1, p. 289 Burn, R., Introduction to Nuclear Reactor Operations, @ 1988, @ 4.6, p. 4-16.

ANSWER ( A 80) c REFERENCE Glasstone, S. And Sesonske, A, Nuclear Reactor Engineering, Kreiger Publishing, Malabar, Florida,1991, @S 2.70 - 2.74, pp. 65 - 66 Burn, R., Irdroduction to Nuclear Reactor Operations, @ 1988, G 5.'2, p. 5-2.

ANSWER  !

(A.11) c REFERENCE (A.11) j Burn, R., Introduction to Nuclear Reactor Operations, @ 1988, 5 5.6, p. 5-25. '

~. . _ . . _ . . _ . _ _ _ . _ . . _ _ _ - _ _ . _ _ _ = . . _ _ . _ . _ . . _ . . - _ _ . _ . _ _ . _ . . . _ -

1 Section A R Theory. Thermo. and Facility Characteristics Page 24 '

l ANSWER (A.12) i b j REFERENCE (A.12)

Glosstone, S. And Sesonske, A, NuclearReactor Engineering, Kreiger Publishing, Malabar,  :

Florida,1991, f 1.52, p.16. . .

l Lamarsh, J.R., Introduction to Nuclear Engineering, Addison - Wesley Publishing, Reading, _

Massachusetts, 1983 9 3.7, Table 3.6, p. 77  !

Burn,' R., /ntroduction to NuclearReactor Operations, @ 1988, @ 3.2.1, p. 3-5. l ANSWER (A.13)  :

a i REFERENCE (A.13) i Glasstone, S. And Sesonske, A, Nuclear Reactor Engineering, Kreiger Publishing, Malabar,-

Florida, 1991, 6 5.31, p. 240. l Lamarsh, J.R., /ntroduction to Nuclear Engineering, Addison - Wesley Publishing, Reading, )

Massachusetts, 1983 6 7.1, pp. 286 - 289.  ;

Burn, R., Introduction to Nuclear Reactor Operations, @ 1988, 9 4.7, p. 4-21

^

ANSWER (A.14)  ;

a ,

REFERENCE (A.14) i Glasstone, S. And Sesonske, A, Nuclear Reactor Engineering, Kreiger Publishing, Malabar, i Florida, 1991, $ 5.18, p. 234  :

I ANSWER. (A.15)  :

c REFERENCE (A.15)  !

Glasstone, S. and Sesonske, A. Nuclear Reactor Engineering, Kreiger Publishing, Malabar, j

- Florida, 1991, 99 5.224 - 5.229, pp. 306 - 307. t ANSWER (A.16) l d -

REFERENCE (A.16)

Glasstone, S. and Sesonske, A, Nuclear Reactor Engineering, Kreiger Publishing, Malabar, i Florida,1991, 3.161 - 3.163, pp.190 - 191.  !

CR i /CR2= (1 - Ke2)/(1 - K,,n) 1/32 (1 - 0.95) = 1 - Ke2 1 - 0.05/32 = Ke2 Ke2 = 0.9984 l 1

Section A R Theorv. Thermo. and Facility Characteristics Page 25 ANSWER (A.17) b PsEFERENCE (A.17)

Glasstone, S. and Sesonske, A, NuclearReactor Engineering, Kreiger Publishing, Malabar, Florida,1991, SS 5.56 - 5.80, pp. 250 - 260.

Lamarsh, J.R., Introduction to NuclearEngineering, Addison-Wesley Publishing, Reading, Massachusetts,1983. 7.4, pp. 316 - 322.

Burn, R., Introduction to Nuclear Reactor Operations, @ 1988, 6g 8.1 -8.4, pp. B 8-14.

ANSWER (A.18) c REFERENCE (A.18)

Burn, R., Introduction to Nuclear Reactor Operations, @ 1982, @ 6.2.3, p. 6-4.

ANSWER (A.19)

REFERENCE (A.19)

Burn, R., Introduction to Nuclear Reactor Operations, @ 1988, Chapt. 2 pp. 2-36 i

ANSWER (A.20) a REFERENCE (A.20)

Glasstone, S. And Sesonske, A, NuclearReactorEngineering, Kreiger Publishing, Malabar, Florida,1991, S 3.161 - 3.163, pp.190 - 191.

Burn, R., Introduction to Nuclear Reactor Operations, @ 1988, Chapt 5, pp. 5 5-28.

Section B Normal. Emeroency and Radioloaical Control Procedures Page 26 ANSWER (B.1)' '

b REFERENCE (B.1) ,

10CFR20.1201(a)(1) 5000 mr x1ht x day = 125 days 10 mr 4 hr

~ ANSWER (B.2) ,

c '

REFERENCE (B.2) 10 CFR 55.53.e. l ANSWER (8.3) c

' REFERENCE (B.3)  ;

Technical Specification 61 Definitions i

ANSWER (B.4) I c

REFERENCE (B.4) 10 CFR 20.1004

. ANSWER (B.5) c REFERENCE (B.5)

A Geiger Mueller detector is not sensitive to energy level.

ANSWER (B.6) d REFERENCE (B.6)

I=lo e ^8 or i = lo (-10**) I = 6 (-10"*) = 3.37 ANSWER (8.7) d REFERENCE (B.7) i I=lo e* where A = 0.693/t.4 In (1/4.5) = -[(0.693/15)

• t) t = 32.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> ANSWER (B.8) b REFERENCE (B.8) 10 CFR 20.1206

)

I l

- - _. - . - ~ - ..- ~ . ,

Section B Normal. Emeraency and Radioloaical Control Procedures Page 27 i

ANSWER (B.9) t c

RnFERENCE (B.9).

University of Florida Emergency Plan,9 3.12 ANSWER (B.10) b REFERENCE (B.10)

University of Florida Emergency Plan, S 7.3.2 l

l ANSWER (B.11) i d

l REFERENCE (B.11) ]

SOP-0.6, S 7.1.2.2 l ANSWER (B.12) a, Check; b, Test; c, Check; d, Calibration l REFERENCE (B.12) l Technical Specifications,61.0 Definitions ANSWER (B.13) l a  !

REFERENCE (B.13)

SOP A.2,9 4.0 ANSWER (B.14) b-REFERENCE (B.14)

SOP A.8,' S 4.6.5 i ANSWER (B.15) b, c .

REFERENCE (B.15)

DESIGN AND OPERATING CHARACTERISTICS OF THE UFTR, p. 22.

ANSWER (B.16) b REFERENCE (B.16) 10CFR20.1003 ANSWER (B.17) d  ;

REFERENCE (B.17)

SOP A.2, p. 3 of 11,9 4.2.2 l

Sgtetion B Normal. Emeroency and Radiolooical Control Procedures Page 28 ANSWER (B.18) d REFERENCE (B.18)

SOP A.2, p. 3 of 11, 9 4.2.4 ANSWER (B.19) b REFERENCE (B.19)

SOP A.2, p. 4 of 11,9 4.4.2 i

l-

l l

I Section C Plant and Radiation Monitorina Systems Page 29 l ANSWER (C.1) DELETED

{

REFERENCE (C.1)

SOP A.6, @@ 4.2 & 4.2.1 ANSWER (C.2) b REFERENCE (C.2) '

SOP A.8 6 4.3.

ANSWER (C.3) a, Full; b, Full; c, Rod-Drop; d, Rod-Drop; e, Full; f, Rod-Drop REFERENCE (C.3)

Table 3.1, Specifications for reactor safety system trips, also similar to Facility Question 5 in Reactor Protection System Section.

ANSWER (C.4) c REFERENCE (C.4)

DESIGN & OPERATING CHARACTERISTICS OF THE UFTR, pp.11,12.

ANSWER (C.5) b REFERENCE (C.5)

DESIGN & OPERATING CHARACTERISTICS OF THE UFTR, p.17 ANSWER (C.6) c REFERENCE (C 6)

DESIGN & OPERATING CHARACTERISTICS OF THE UFTR, pp.18,19

. ANSWER (C.7) c REFERENCE (C.7)

DESIGN & OPERATING CHARACTERISTICS OF THE UFTR, pp. 20 ANSWER (C.8) b REFERENCE (C.8)

DESIGN & OPERATING CHARACTERISTICS OF THE UFTR, p. 20

~

Section C Plant and Radiation Monitorina Systems Page 30 ,

ANSWER (C.9) e REFERENCE (C.9)

DESIGN & OPERATING CHARACTERISTICS OF THE UFTR, p. 30 ANSWER (C.10) b REFERENCE (C.10)

DESIGN & OPERATING CHARACTERISTICS OF THE UFTR, Figure 28.

ANSWER (C.11) d REFERENCE (C.11)

UFTR DESIGN & OPERATING CHARACTERISTICS , Rewording of 2/21/96 requal exam question #3.

ANSWER (C.12) '

d REFERENCE (C.12)

UFTR DESIGN & OPERATING CHARACTERISTICS Rewording of 2/21/96 requal exam, question #11.

ANSWER (C.13) c REFERENCE (C.13)

UFTR DESIGN & OPERATING CHARACTERISTICS Rewording of 2/21/96 requal exam, question #17.

ANSWER (C.14) d REFERENCE (C.14)

UFTR DESIGN & OPERATING CHARACTERISTICS, Rewording of 2/21/96 requal exam, question #23.

ANSWER (C.15) a REFERENCE (C.15)

UFTR DESIGN & OPERATING CHARACTERISTICS 8/16/90 requal exam, question #9.

Section C ~ Plant and Radiation Monitorina Systems Page 31 ANSWER (C.16) a,4; b, 2; c,1; d, 2; e, 3 REFERENCE (C.16) .

i UFTR DESIGN & OPERATING CHARACTERISTICS 82/16/90 requal exam, question #12.  ;

I ANSWER (C.17)  !

c. l REFERENCE (C.17)- l UFTR DESIGN & OPERATING CHARACTERISTICS 8/16/90 requal exam, question #27.

ANSWER (C.18) a-  :

REFERENCE (C.18)

Lecture Notes UFTR instrumentation and Control pp. 5 through 8. l l

i 1

i