ML20149K064

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NRC Operator Licensing Initial Exam Rept 50-228/OL-97-01 on 970602 & 03.Exam Results:Written & Operating Exams Administered to One Candidate for Senior Operator License. Candidate Passed Exams
ML20149K064
Person / Time
Site: Aerotest
Issue date: 07/07/1997
From: Mendonca M
NRC (Affiliation Not Assigned)
To:
Shared Package
ML20149D172 List:
References
50-228-OL-97-01, 50-228-OL-97-1, NUDOCS 9707290232
Download: ML20149K064 (43)
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Text

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i 3 U. S. NUCLEAR REGULATORY COMMISSION

OPERATOR LICENSING INITIAL EXAMINATION RETORT REPORT NO
50-228/OL-97-01

,' FACILITY DOCKET NO.: 50-228 4

l FACILITY LICENSE NO : 'R-98

! FACILITY: Aerotest Operations, Inc.

a.

EXAMINATION DATES: June 2 and 3,1997 1

! EXAMINER: Marvin Mendonca, Chief Examiner

SUBMITTED BY: W~ 7/7/i7 i Marvin Mendonca, Chief Examiner Date l

l

SUMMARY

! The NRC examiner administered written and operating examinations to one candidate for a Senior Operator License (SRO). The candidate passed the written and operating examinations.

I.

i REPORT DETAILS i

i 1. Examiner:

l Marvin Mendonca, Chief Examiner o

i' l 2. Results: l 3

RO PASS /FAll SRO PASS / FAIL . TOTAL PASS / Fall  ;

. Written 0/0 1/0 1/0

Operating Tests 0/G 1/0 1/0 Overall 0/0 1/0 1/0 t
3. Exit Meeting:

i

- Marvin Mendonca, Examiner, NRC j Ray Tsukimura, President, Aerotest Operations Inc. (Aerotest)

[ Alfredo Meren, Manager, Neutron Radiography, Aerotest i Sandy Warren, Radiological Safety Ofncer, Aerotest i

i e

i 4

j. .

D -

i OPERATOR LICENSING EXAMINATION i  !

AB REG O 4

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W. O

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,/ g O @% '

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+++++

i AEROTEST 4, June 02.1997 i

[ ,' U. S. NUCLEAR REGULATORY COMMISSION

l. NON-POWER REACTOR LICENSE EXAMINATION i

FACILITY: AEROTEST

[ REACTOR TYPE: TRIGA p

DATE ADMINISTERED: 06/02/97

- REGION
IV p

CANDIDATE:

f l INSTRUCTIONS TO CANDIDATE:

r Antwers are to be written on the answer sheet provided. Attach the answer sheets to the examination.

l. Pcints for each question are indicated in parentheses for each question. A 70% overall is required to

! pics the examination. Examinations will be picked up three (3) hours after the examina*Jon starts.

% OF i CATEGORY % OF CANDIDATE'S j VALUE TOTAL SCORE CATEGORY l'

20.00 33.3 A. REACTOR THEORY, THERMODYNAMICS AND FACILITY

, OPERATING CHARACTERISTICS 4

l 20.00 33.3 B. NORMAL AND EMERGENCY OPERATING PROCEDURES j AND RADIOLOGICAL CONTROLS 20.00 .22 1, C. PLANT AND RADIATION MONITORING SYSTEMS 1

60.00 TOTALS
FINAL GRADE All work done on this examination is my own. I have neither given nor received aid.

Candidate's Signature

, NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS l-

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
sevsre penalties, c ,

' 2. After the examination has been completed, you must sign the statement on the cover sheet indicating i that the work is your own and you have neither received nor given assistance in comp! sting 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 i- contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.

4

. 4. Use black ink or dark pencil 9.Dlx to facilitate legible reproductions.

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

i~ l

6. The point value for each question is indicated in [ brackets) after the question.

l 1

[ 7. If the intent of a question is unclear, ask questions of the examiner only.  !

I l

! l

8. When tuming in your examination, assemble the completed examination with examination questions, l examination aids and answer sheets. In addition tum in all scrap paper. l 1
9. To pass the examination you must achieve a grade of 70 percent or greater in each category. '
10. There is a time limit of three (3) hours for completion of the examination.
11. When you have completed and tumed 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. '

l 1

- .. - . .. =- . . - . .

. EQUATION SHEET 1

d = rho, AT = th AH = UA A7 p"'" , (p-p)2 l 2a(k)f t' = S x 10 seconds ggg , _ S_ , 5

-p 1 -K,,,.

1 A,, = 0.1 seconds -1 CR,(1 -Kg) = CR,(1 -K,,,)

CR,(-p,) = CR 2(-pa)

A 1 -K.e,

. SUR = 26.06 ,,#- M=

,D-p, 1 -K,,, j y, 1 , GR, p = p,10suRm 1 -K,, CR, e

SDM = (1 -K,y) p ,p ,i en I

T= P = OII ~P) P

, p-) 0-p

~

K,,, - K,,, g. g_p sp , _k,,,x K,,, p A,p 0.693 7, (K,y-1) )

A K,y DR =DR a e* DR,d,* = DR,d,'

M - Rem, G - wies, DR = 6CW(n) g E - Mev, R - feet  ;

I (92-D)* ,

(p,-D)' l Peak, Peak, l

1 l

i 1 Curie = 3.7 x 10" dis /sec 1 kg = 2.21 lbm i l

1 Horsepower = 2.54 x 10' BTU /hr 1 Mw = 3.41 x 10' BTU /hr I 1 BTU = 778 ft-lbf *F = 9/5 *C + 32 l 1 gal (H2 O) = 3 lbm *C = 5/9 (*F - 32) I c, = 1.0 BTU /hr/lbml*F c, = 1 callsec/gmI*C l 1

1

, cw ,_ . . . . . .. .. .. .~ .~ .. . . . . . _ . _. ....

- .. - - - ~ - - ."

1 4 g 10 l -

i Z t -

-ee.e#

.h, 1000 KW 600 KW i

' ~

[' .

~ ~ ~ 300 KW i

- I

.10 2 _ 1

- f ICO Kw

  • - - 60 kW. l l

i 30 KW i

  • x ,

<l '

E 3 I '

O to 8

< N - 10 RW L

, S

  • W I a b Q ]

'M m j $ 4 i

3 .

f f'.

    1. j E

w 10 .

U -((

I KW Y '.f F I";I.81 X 10 BARNS

  • O ,

lo Yo = 1.B 4 X lo FORIRW 10

-5 , J

+

1

  • n

.i

~

l .-

u

-6 1 10 a l I l l \ \

20 40 60 80 .10 0 12 0 14 0 1

TIME (H R) .

Figare'1.6. Calculated reactivity loss due to Xe135 buildup 1-27

l , A. RX THEQRY. THERMO & FAC OP CHARS ~ Pzgo 1  ;

ANSWER SHEET j i

Multiple Choice (Circle or X your choice)

if you change your answer, write your selection in the blank.

001 a b c d

- 002 a b c d i i

003 a b c d

l l 004 a b c d l

005 a b c d l

\

1

. 006 a b . c d
1 007 a b c d I 008 a b c d 009 a b c d )

010 a b c d l 011 a b c d . _

012 a b . c d 013 1 2

3 l

4 '

014 a b c d 015 a b c d 016 a b c d 017 a b c d 018 a b c d 019 a b c d 020 a b c d

(*"" END OF CATEGORY A "*")

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

j l

[ B...NORMAllEMER

G. PROCEDURE

S & RAD CON P:g3 2 i ANSWER SHEET l i

001 m b . c d 1 1

l 002 a b c d i

i. .

003 m b c d I 4

004 1 2 3 4 l i

005 a b c d l e

006 a b c d l 007 a b c d

(

008 a b c d I 009 a b c d I

010 a b c d j 011. a b c d 1

012 a b c d I l

013 1 2 3 4  ;

014 a b c d 015 a b c d 016 a b c d __

4 017 a b c d 018 a b c d 019 a b c d

020 a b c d I 4

4

(***** END OF CATEGORY B *****)

e

  • C. PLANT AND RAD MONITORING SYSTEMS Pcgd 3 ANSWER SHEET 001 a b c d l

002 a b c d __

j- 003_ a b c d 004 a b c d i

005 1 l 2 __

3 4

e 006 a b c d

+

d

007 a b c d 4

008 a b c d i

009 a b c d 010 a b c d 1 011 a b c d i 012 a b c d 013 a b c d i

014 - a b c d 015 ' a b c d 016 a b c d 017 a b c d

' 018 a b c d 019 a b c d 020 a b c d

(~ END OF CATEGORY C "***) ,

(*******" END OF EXAMINATION *"****"*) j l

l 3

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

h Section A Reactor Theorv: Thermodynamics and Facility Characteristics Page 1 i  !

QUESTION (1) [1.0]

i 1

{ Which one of the following results in the most energy release per. fission?

a. The initial . kinetic energy of the fission fragments. i l
b. The initial kinetic energy of. all the resultant neutrons. '

{ c. The energy released by prompt and delayed gamma rays. j

! d. -The energy released by prompt and delayed beta particles.

[ QUESTION (2) [1.0]

i;  ;

l Which one of the following describes the' reason that the delayed neutron fraction is so very important?

l a. It eases reactor control design requirements.

( .

4 l b. It provides a predominant source of energy release. '

[ c. It provides additional fuel through the creation of Plutonium-239 from

! Uranium'238.

I

d. It provides neutrons that fission Plutonium-239 and extend fuel life.

i-QUESTION (3) [1.0]

i Which orie of the following describes the term " Critical Heat Flux?"

a. The heat flux when the reactor is critical.
b. The heat flux above which there would be excessive' fuel clad temperatures.
c. The heat flux above the fluid triple point.
d. The heat flux at which nucleate boiling occurs.

q

i. ..

l* .

S.tqLion A Reactpr Theory. Thermodynamics and Facility characterist.i.s3 page 2 QUESTION (4) [1.0) l Using the following reactivity worths . calculate the Shutdown Margin with the l

Shim rod stuck all the way out. '

Safety Rod: $4.00 Shim Rod: $4.50

Regulating Rod
. $1.75 Core excess reactivity: $2.5 I a. $2.50
b. $3.25 l c. $5.75
d. $7.75 i
QUESTION (5) [1.0]

l Which ONE of the following statements describes Count Rate characteristics after a control rod withdrawal with the reactor subcritical? (Assume the Rx j - remains subtritical.)

., a. Count Rate will ra to a stable value.pidly increase (prompt jump) then gradually increase I b; ' Count. Rate will rapidly increase (prompt jump) then gradually decrease j to the previous value.

! -c. Count Rate will rapidly increase (prompt jump) to a stable value.

[ d. There will be no change in Count Rate until criticality is achieved.

QUESTION (6) [1.0]

The neutrons that are released by fission are fast neutrons because they:

a. allow accurate and quick instrumentation response.
b. are responsible for prompt criticality,
c. decay rapidly to stable levels.
d. are at high kinetic energy levels.

1

_ ._ _ _ ~

i 1

Section A Reactor Theory. Thermodynamics and_ Facility Characteristics Page 3 OVESTION (7) [1.00]

t Which ONE of the following describes the characteristics of e good moderator?

a. High scattering cross section and low absorption cross-section ,
b. Low scattering cross section and high absorptiori cross-section
c. Low scattering cross section and low absorption cross-section
d. High scattering cross section and high absorption cross-section QUESTION (8) [1.0]

What is the count rate in a reactor that has a multiplication factor equal to 0.7 and a neutron source indicating 600 neutrons per second?

a. 1000 i
b. 2000
c. 3000
d. 4000
QUESTION (9) [1.0]

Which ONE of the following describes the difference between reflectors and moderators?

. a. Reflectors decrease core leakage while moderators thermalize neutrons

b. Reflectors shield against neutrons while moderators decrease core leakage
c. Reflectors decrease thermal leakage while moderators decrease fast

. leakage

d. Reflectors thermalize neutrons while moderators decrease core leakage l

'Section A Reactor Theory. Thermodynamics and Facility Characteristics Page 4 b Question (10) .[1.0]

i .

LFor a reactor at 50. kilowatts, which one of the following -is time to reach 250 -

j kilowatts given a steady state reactor period of 40 seconds?

a. 32 seconds-1 i -
b. 64 seconds

!' c. 100' seconds

d. 128 seconds

[

QUESTION (11) [1.0]

l: Given the following:

, 1. Pri ma ry fl ow. . . . . . . . . . . . . . . . . . . . . . . . . . . 180 gpm

2. Pool outlet' temperature.............. ... 115 deg F

-3. Seconda ry f10w. . . . . . . . . . . . . . . . . . . . . . . . . . . 170 gpm

s. 4. HX secondary inlet temperature. . . . . . . . . . . 90 deg F

!; 5. HX secondary outlet temperature........... 95'deg F

! 6. Specific heat for H 0.2 . . . . . . . . . . . . . . . 0.146 kw/gpm-degrees F.

i

[ Which one of the following is approximately the power level?

l a. 31.25 kw l b.~ 62.5 kw j c. 125 kw l

! .d. I250 kw .

i-i.;

1 1-2 l

i i

j. '

l

-e, w -

w

1

~

\

L j

.Section A Reactor Theory. Thermodynamics and Facility Characteristics Page 5 l

[

l: .- QUESTION 12 [1.0] '

1

For a reactor at steady state power at.a uniform moderator temperature of l about 90*C. the fuel temperature reactivity coefficient is -1.0x10"5 AK/K/ C l

,, and the moderator temperature coefficient of reactivity is -1.0x10' AK/K/ C.

L A regulating rod with a uniform rod worth.of 0.05%AK/K/ inch is withdrawn 2.0 inches. A moderator. temperature rise cf a uniform 5*C was measured. Which l

'; one of the following would be the average fuel tem)erature change at the new steady state power level assuming no other feedbact effects other than -i

moderator and fuel reactivity coefficients?

1

[ a. decreased by about 10*C-

[ b. increased by about 10 C

c. . decreased by about 100 C l 1 ,
d. increased by about 100 C
0VESTION.13'[1.0]

y t For the given conditions. reactor Note: Ea:h conditions reactivity and tLAJ.CB the appropriate k,fr condition reactivity may be used more t and k,Ean L once.or not at all. Each correct answer is worth % of a point.

- Reactor condition
1. Subtritical-Critical
2.
3. Supercritical L

Reactivity and k,ff:

l

.a. p<0 and k,rpl.0

( b. p=0 and k,,,=1.0

?

i- c. p<0 and k,ff<1.0 i

j d. p>0 and k,,rl.0 t

i Section A Reactor Theory. Thermodynamics and Facility Characteristics Page 6 r QUESTION 14 Which one of the following figures most closely depicts the reactivity versus time plot for xenon for the.following series of evolutions:

TIME EVOLUTION 1 From clean core conditions, startup to 500 KW startup:

2 0)eration at 500 KW for 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s:

'3 Slutdown for 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />; 4 Startup to 250 Kw and steady state?

a b

  • = '
desie rp

, ]

son -

= , --

. V k

. in . ><-s-><. . > < in ><-->< . .

  • TtWE TIME 6

c d

x.

de rp d*

  • Sa m i sm -

-- - x -- -

I

< in > < - - > , . , <- in ><-3->< .  : l nue ruz l

l i

s l

4

M Section A Reactor Theory. Thermodynamics and Facility Characteristics Page 7 F

QUESTION (15) .[1.00]

o .

During a reactor startup criticality occurred before expected conditions.

Which one of the followings could be the cause?

l a. ~ Experiment a'dding positive reactivity.

l35

b. Xe peaked.

j c. Moderator temperature increased.

d. ' Power defect (Reactor power . increasing).
0VESTION (16) [1.0]

i Which one of the following is the I.ARGEST contribution to~the strongly

negative fuel temperature reactivity coefficient in the ARRR?

i a. With increased fuel temperature, the capture resonances in U"5 are Doppler broadened increasing the capture of thermal neutrons.

b. With increased fuel temperature, core leakage increases and neutron j

captures occur outside the fuel.

c. With increased fuel temperature. the capture resonances in U"8 are  :

Doppler broadened increasing the capture of thermal neutrons.

d. With increased fuel temperature, the zirconium hydride imparts more energy to the thermal neutrons and there is an increased probt '.lity of escape and capture.
QUESTION (17) [1.0)

[ 1411ch'one of the following is the purpose of having a neutron source?

i. a. To provide'a means for allowing reactivity changes to occur in a
subcritical reactor.

L l b. To provide a detectable neutron level for monitoring reactivity changes l

in a shutdown reactor.

c. To generate a sufficient neutron population to start a fission chain i reaction for reactor startup.

j _d'. .To compensate for neutrons absorbed by burnup.

i i

i' l

l

. l

[

Section A Reactor Theory. Thermodynamics and_ Facility Characteristics ~Page 8

. QUESTION (18) -[1.0] l

With the reactor on a constant period, which power change requires the LONGEST l time to occur?
a. 4% of rated power - going from 1% to 5% of rated power
b. 15% of rated power - going from 5% to 20% of rated power
c. 30% of rated power - going from 20% to 50% of rated power
d. 50% of rated power - going from 50% to 100% of rated power 1 OUESTION (19) [1.0]

Which one of the following is the approximate half life of the longest-lived 4

. group of delayed neutron precursors?

a. 80 seconds i
b. 70 seconds 4

[

c. 55 seconds
d. 40 seconds

! . QUESTION (20) [1.0]

i' Which one of the following provides the reason that rod calibrations are performed at power levels between 100 and 3000 watts?

a. Lower power levels could result in a significant contribution from source neutrons & higher power levels could cause significant fuel temperature reactivity contributions.
b. Lower power levels could result in a significant contribution from source neutrons & higher power levels could cause significant bath temperature reactivity contributions.
c. Lower power levels could cause shutdown margin to be exceeded with increased safety rod worth, and higher power levels could raise the temperature to greater than Technical Specification allowed values. pool
d. Lower power levels could result in temperatures falling below Technical Specification limits, and higher )ower levels could cause the shutdown margin to be exceeded as rod wort 1 significantly decreases with increased fuel and moderator temperature.

End of Section A Reactor Theory. Thermodynamics and Facility Characteristics

l l

Section B Normal. Emeraency and Radioloaical Control Procedures Page 9 OUESTION (1) [1.0) l Which one of the following describes the requirement for a temporary change to a procedure that are NOT typographical or spelling corrections?

a. It requires the approval of the Reactor Supervisor.
b. It remains in effect indefinitely withcut further review or approval.
c. It does not change the intent of the previously approved procedure.
d. It does not need to be documented in a Procedure Change Notice.

QUESTION (2) [1.0]

Which one of the following describes allowed operator actions during an emergency?

a. Any Senior Licensed Operator may deviate from procedures as required during an emergency to assure the health and safety of the staff and general public.
b. Deviations from procedures during an emergency need not be reported to the Reactor Supervisor.
c. Deviations from procedures during an emergency should immediately be reported to the Radiation Safety Officer.
d. In accordar,ce with 10 CFR 50.54x. a licensee may NOT deviate from license and Technical Specifications in an emergency even to protect the public health and safety.

QUESTION (3) [1.0]

Which one of the following describes the immediate actions of the Senior Reactor Operator in charge during a fire at the ARRR?

a. Inform the Reactor Supervisor.
b. Contact the San Ramon Fire Department unless the it is obvious that the fire can be controlled by portable fire fighting equipment within the facility.
c. Inform the Radiation Safety Officer.
d. Inform the Reactor Supervisor who will contact the San Ramon Fire Department unless the it is obvious that the fire can be controlled by portable fire fighting equipment within the facility.

Section B Normal . Emeraency and Radioloaical Control Procedures Page 10

QUESTION (4) [0.25 for each correct answer for a total of 1.0]
Match the area of fire to the associate precautions or conditions. (Note:

o M one answer per item. Also, answers may be used more than once or not at i

all).

Area of Fire Precautions and Conditions
1. Area containing a. Even if it is safe to move explosives radioactive material. to a safer location, do NOI move them. *
2. Area containing b. Water may be used to fight the fire.

nuclear fuel.

c. Aerotest personnel will evacuate the area &
3. Area containing fire fighting personnel entering such areas explosives. will use self-contained breathing apparatus.
4. Area containing d. Carbon dioxide and halon are the l l

3 electrical equipment. E extinguishers to be used.

QUESTION (5) [1.0]

1 Which one of the following describes the condition when a Special Work Permit

is required?
a. All work in a high radiation area.
b. All maintenance and other support personnel performing jobs in radiation areas to which they are NOI regularly assigned,
c. All maintenance and other support personnel performing jobs to which they are HQI regularly assigned.
d. 'All activities where radiation monitoring is required.

l l

w- ~ - - - - - - - - . ,

4 l Section B Normal. Emeraency and Radiolooical Control Procedures Page 11 i;

l OUESTION (6) [1.0)

Which one of the following provides the minimum personnel requirements for movement of fuel or graphite elements within the reactor pool?

a. The same as for reactor operations.
i. b. One Senior Reactor Operator, another licensed operator and another individual,
j. c. Three licensed operators.
d. ' One Senior Reactor Operator, and two other licensed operator.

l QUESTION (7) [1.0]

Which one of the following is NOT a suggested method for control rod i calibration?

a. Period measurement and comparison.to In-Hour equation table.
b. Rod drop and ratio long-term beta decay neutron values at various times 4

to the initial value.

I c. Rod drop and measurement of delayed neutron value at one minute to pre-l drop value.

d. Reactivity loss with increasing power correlation to temperature coefficients.

QUESTION-(8) [1.0]

Which one of the following is HQT a suggested method for reactor power calibration?

a. Calorimetric measurement of reactor pool water temperature rise and comparison to preestablished value.
b. Flow and differential temperature measurement across intermediate heat exchanger and comparison to preestablished values.
c. Flow and differential tem)erature measurement across cooling tower and comparison to preestablisied values.
d. Reactivity loss to power measurement.
Section B Normal. Emeroency and Radioloaical Control Procedures Page 12 00ESTION (9) [1.0]

i

Which one of the following is the reason that it is preferable to store all the fuel elements in the reactor tank during defueling and refueling j operations?
a. In the interest of saving time during the operations.
b. In the interest of greater safety and lower personnel expwire.
c. In the interest of reducing the number of radiation alarms.

~

d. In the interest of maintaining k,rr at about 0.9.

i 00ESTION (10) [1.0) l Which one of the following is a requirement for fuel movement?

a. The secuence of fuel movement will' begin in the center of the core and
proceec outward to reduce reactivity quickly, i
b. The highest worth control rod is removed prior to fuel movement to-j demonstrate shutdown margin.

I

c. Only one fuel elenat at a time is allowed to be out of storage or the core.
d. Instrumented elements must be removed last as they provide reactor safety scram signals.

~

QUESTION (11) [1.0)

Which one of the following describes the expected acute effects of 25 to 150
rems of whole-body penetrating ionizing radiation in less than one day to a human being?
a. None.
b. Blood cell changes,
c. Nausea and vomiting within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
d. Nausea and vomiti.ng in under 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and minimal incapacitation.

Section B Normal. Emeroency and Radioloaical Control Procedures Page 13

.0VESTION (12) [1.0]L Which one of the-following describes the reason that graphite elements are added from their storage location to the core at approximately one half the expected allowable number?

a. To ensure that they are not mixed up with fuel elements.

! b. .To ensure that shutdown margin requirements are not exceeded.

L

c. To ensure that excess ~ reactivity requirements are not exceeded.

L d. To ensure that maximum rate of reactivity-insertion requirement is not exceeded.

(

L

'00ESTION (13) [1.0. 0.25 points for each correct answer]

l Match the annual dose limit values to the type of exposure.

-Annual' Dose Limit Value Tvoe of Exoosure

a. 0.1 rem 1. Occupational Total Effective Dose Equivalent (TEDE)
b. 5.0 rem 2. TEDE to.a member of the public-
c. 15.0 rem 3. Lens of the Eye
d. 50.0 rem 4. Extremities c

)

OVESTION (14) [1.0]

. Which one of the following describes the frequency requirement for' checking and calibrating radiation survey instruments, changing themoluminescent dosimeters, and swipe sampling where radioactive material 1s handled?

a. At least once per month.
b. At least once per quarter.
c. At least once per six months
d. At least once per year.

?

s Section B Normal . Emeraency and Radiolooical Control Procedures Page 14 QUESTION.(15) [1.0] -

Which one of the following correctly fills in the blanks for: Weekly cumulative whole body. exposures must be limited to mrem unless higher exposures are specifically approved by the . Employee quarterly exposures > rem and visitor exposures of > mrem must be

investigated and the type and circumstances of exposure documented?

! a. 100. Reactor Supervisor. 1 0, 30

b. 100. Radiological Safety Officer.1.0. 30 I
c. 375. Reactor Supervisor. 3.75, 100

{

l d. 375, Reactor Safeguards Committee. 3.75, 100 i 1

i OUESTION (16) [1.0)

Which one of the following is procedurally required to be trended on a I

]. quarterly frequency? 1 i

a. Shutdown Margin. I

! b. Excess Reactivity at critical, i I \

2 c. Reactor cool .t radiation level. '

i d. Reactor pool water level.

I 3

4 ,

i i

! I 4

4 4

1 h

2 k

v .

Section B Normal. Emeraency and Radiolnaical Control Procedures Page 15 l QUESTION (17) [1.0].

l Which one of the following'is the dose rate at 3 feet from a point source, if

[ the dose rate at 1 foot from the source is 450 mrem /hr?

a. 5 mrem /hr.
b. 15 rr. rem /hr.

j c. 50 mrem /hr.

L d. 150 mrem /hr.

'00ESTION (18) [1~. 0]

~Which.one of the following describes the allowed operating state of the reactor assuming that the demineralizers/ filter string is plugged?

a. The reactor may be operated if the primary coolant conductivity is less than 7.5 micromhos/cm and pH is greater than 5.
b. The reactor can be operated if the primary coolant conductivity is less than 5 micrombos/cm and pH is less than 7.5.
c. Technical Specifications do not allow operation without the primary coolant radioactivity monitor, and the reactor must be shutdown.
d. An automatic scram will occur on loss of demineralizer system flow rate.

QUESTION (19) [1.0]

Fill in the blanks: In order for a reactor operator to stay licensed, the  !

operator must renew the license every , have a physical exam every . .

and stand watch for every . I

a. 2 years. 1 year. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, month.
b. 2 years. 2 years. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, month,
c. 6 years. 2 years. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. 3 months,
d. 6 years. 6 years. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. 6 months.  ;

1

i Section B Normal. Emeraency and Radioloaical Control Procedures Page 16 l

QUESTION.(20) [1.0]

Which one of the following is the MINIMUM dose rate at 30 centimeters from the

! source for a posting of " CAUTION - HIGH RADIATION AREA?"

a. 2 mrem /hr

, b. 5 mrem /hr

c. 30 mrem /hr d 100 mrem /hr

! ****************************************************************************** I

End of Section B Normal. Emeraency and Radiolooical Control Proce 'l
*********************************************************************dures *********

i 4

I i

4 1

4 Section C Plant and Radiation Monitorina Systems Page 17 OUESTION (1) [1.0]

The basic design principle of the reactor building ventilation system for the containment of radioactive material is to:

a. Release radioactivity through filters.
b. Immediately release radioactivity through the gravity ventilators located on the roof of the building.
c. Maintaining activa ventilation systems operable in the reactor area.
d. Contain the airborne material in the reactor bay until it decays or can be controlled and/or decontaminated.

OUESTION (2) [1.0]

Which one of the following is NOT a function of the triangular spacer block that is part of the top end fitting for each fuel element?

a. Provides alignment and position for the top of the fuel element in the top grid plate.
b. Provides passages for cooling water through the grid plate to enhance natural convection cooling of the core.
c. Demonstrates proper installation of the fuel element when it is level with the grid plate, f
d. Provides a difference with the graphite reflector elements in that they do not have these spacers.

QUESTION (3) [1.0]

Which one of the following is the expected light indications when the control rod is being moved by its motor control system?

a. DN light off, UP light on, CONT light on, ON light on.
b. DN light on, UP light off, CONT light on, ON light on.
c. DN light off, UP light off, CONT light on, ON light on.
d. DN light off, UP light on, CONT light off ON light on.

. I Section C Plant and Radiation Monitorina Systems Page 18 I

QUESTION (4) [1.0] l Which one of the following correctly describes the safety system design bases?

The safety system shall be designed such that no single component failure or i circuit fault shall:

a. disable the automatic scram circuits.
b. disable the manual scram circuits.  !
c. simultaneously disable both the automatic and manual scram circuits.
d. simultaneously disable the automatic scram circuits, the manual scram circuits, and the radiation monitoring system.

QUESTION (5) [0.25 for each correct answer for a total of 1.0] l Match the Nuclear Instrumentation Channel No. with the detector type and scram function for the nuclear instrumentation channels. NOTE: Some detectors  !

types may apply to more than one channel or not at all.

Channel Scram Detector Tvoes 1 Short Period & Low Count Rate a. Proportional Counter l 2 Short Period b. Uranium Loaded Fission Chamber  !

3 High and Low Flux Scrams c. Compensated Ion Chamber i 4 High and Low Flux Scrams d. Uncompensated Ion Chamber i

I i

I i

l l

l 1

. . . . . . . . . .- . - . . . . . ~ - - . - - - - - . - - . .- - . . - - . - -

Section C Plant and Radiation Monitorino Systems Page 19 OUESTION (6) [1.0]

Which one of tne following describes the reason that the pool cooling water  ;

system is. activated at crossover (a few watts power)?

l a. To delay the time required for induced radioactivity to reach the top of -

the pool.

b. To ensure that the pool water radioactivity monitors are properly

, functioning.

1

. c. To cool the pool water at higher power levels,

d. To ensure makeup water is provided to maintain pool level.

f

. QUESTION (7) [1.0] >

{ Which one of the following is @I required for safety rod withdrawal?

l a. All four nuclear instrument channels are in the OPERATE mode.

b b. The safety system has been reset. '

{ c. The startup channel count rate is greater than 2 counts per second.

d. The shim rod is fully withdrawn.

l l

4 QUESTION (8) [1.0]

i Which one of the following describes ~ the reason for the location of the intake l for the primary cooling loop?

l

a. To ensure natural convection flow of the core.
b. To ensure forced convection flow of the core.

t c. To sweep away induced radioactivity as it reaches the top of the pool.

d. To prevent pumping water out of the pool below the. license minimum value.

[ . Section C Plant and Radiation Monitorina Systems Page 20  !

QUESTION (9) [1'. 0]

. Which one of the following describes the order of major components in the flow path for the demineralizer system?

a. pool water radiation monitor, temperature switch that turns off the

[1 cooling system, demineralizer, and fluid flow switch.

4

b. temperature switch that turns off the cooling system. 2001 water radiation monitor, demineralizer. and fluid flow switc1.
c. pool water radiation monitor, fluid flow switch, demineralizer, and temperature switch that turns off the cooling system.

i

d. fluid flow switch, pool water radiation monitor, demineralizer, and j- temperature switch that turns off the cooling system.
QUESTION (10) -[1.0]

1 Which one of the following is the correct description for the ARRR reactivity i coefficients?

, a. The bath temperature coefficient and the prompt fuel temperature

, coefficient shall be positive at room temperature and becomes negative at j higher temperatures and the coolant void coefficient shall be negative

across the active core.

i i - b. The bath temperature coefficient and the prompt fuel temperature i coefficient shall be negative at all operating temperatures and the coolant void coefficient shall be negative across the active core.

c. The bath temperature coefficient and the prompt fuel temperature coefficient shall be negative at all operating temperatures and the coolant void coefficient shall be positive across the active core.
d. The bath temperature coefficient shall be negative at all operating s temperatures, the prompt fuel temperature coefficient shall be positive at room temperature and becomes negative at higher temperatures and the coolant void coefficient shall be negative across the active core.

4 Section C Plant and Radiation Monitorino Systems Page 21 QUESTION (11) [1.0]

Which one of the following correctly describes the reactor building ventilation system?

a. All circulating fans and air conditioning systems on be shut off from a single control in the control room.
b. All circulating fans and air conditioning systems which supply air to the control room can be shut off from a single control in the control room.
c. The air pressure in the reactor room is to be more than the air pressure in the control room,
d. The air pressure in the reactor room is to be less than the air pressure in the control rcom.

QUESTION (12) [1.0]

Which one of the following correctly describes the conditions from which maximum core excess reactivity is established?

, a. Cold, clean critical, with or without experiments in place, and with the maximum worth rod fully withdrawn.

b. Cold, clean critical, with or without experiments in place.
c. Cold. clean critical, with experiments in place.
d. Cold, clean full power, with or without experiments in place.

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

L Section C Plant and Radiation Monitorina Systems Page 22 ,

4 00ESTION (13) [1.0]

l Which one of the following most completely describes the design attributes of the fixed gamma monitor employing Geiger tube detector located on the wall connecting the control room and the reactor room? l

. a. The monitor serves as an a failed fuel monitor, and alarms at a remote  !

location.

1

! b. The monitor serves as an area radiation monitor. l

c. The monitor serves as an a criticality alarm, and alarms locally.  ;
d. The monitor ' serves as an area radiation monitor and a criticality alarm.

and alarms both at the facility and at a remote location, j QUESTION (14) [1.0]

i Which one' of the .following is a design feature of the large-component irradiation box?

a. During irradiation experiments are unsecured.
b. The box can encompass the entire core area.
c. C0 is used for purging and to maintain a slight positive pressure relativetothepoolwaterpressure,
d. Components can be removed when the box is in its irradiation position as long as the reactor is shutdown.

QUESTION (15) . [1.0]

Which one of the following is a design feature of the pneumatic transfer facility?

a. The pneumatic transfer facility can only be located in the B-ring.
b. The facility is cperated with air.
c. The facility is exhausted through a filter ventilation system and is monitored for radioactivity.
d. Once the automatic timer control is set, the operator can not change the irradiation time.

~.- . - . - - . .-....- - .. ... - .- -.- ... - .-. . - . - . . . . . . . . . . . ..

t -Section C Plant and Radiation Monitorina Systems Page 23 l . 0VESTION (16) [1.0]

l Which one of the following describes a feature of the glory. hole facility?

a .~ The glory hole accepts capsules to a maximum of 34 square inches.
b. The. glory hole is purged with CO 2 to prevent formation of excessive amounts of nitrogen-16.

f

c. Gas samples are.taken near the pool when the facility is operated without l

a shield plug to insure adequate radioactive gas monitoring.

d. The facility must use a shield plug.

d OUESTION (17) [1.0]

, Which one of the following does NQI describe a feature or condition of the ARRR irradiation facilities?

a. The central 7 fuel elements of the reactor may be removed from the core and a irradiation facility can be installed if the cross section is not greater than 16 square inch.
b. Two triangular exposure facilities are available to allow insertion of circular experiments to a maximum of 2.35 inches diameter or triangular 4

experiments to a maximum of 3 inches on a side.

1

c. Irradiation capsules in the shape of dummy fuel elements shall have a maximum inner void volume of 34 cubic inches in the active fuel region.

Components contacting the pool water can be fabricated of any material.

d.

l v

d 1

y y s - -

.a l Section C Plant and Radiation Monitorinc Systems Page 24

QUESTION-(18) [1.0]

Which one of the following describes an evaluation criteria for' experiments?

a, @LY experiments that provide positive reactivity on insertion need be i

evaluated.

b. Rigidly fixed experiments and those that are not rigidly fixed are

, subject to the same maximum reactivity limits.

c. Consideration must be given to the connection or relation of experiments i to determine if their reactivity should be combined.

- d. Experiments with moving parts are limited to reactivity insertions of $1 per second.

e QUESTION (19) [1.0]

Which one of the following describes an automatic system to reduce potential

reactor coolant system iosses?

, a. Primary and demineralizer pump shutoff on radiation indication in the trench.

b. Primary and demineralizer pump shutoff on moisture indication in the

, trench.

c. Primary and demineralizer system valve isolation on low reactor pool
level indication.

j d. Primary and demineralizer system valve isolation on low loop pressure.

OVESTION (20) L1.0]

l Which one of the following provides the reason that the control rods are

interlocked so that only one may be raised at a time?

l a. Power supply overloads with more than one control rod motor operating.

b. Reactor Operator could not manipulate and observe controls,
c. Reactivity Insertion would be too fast.
d. Axial power profiles would be too peaked.
                              • +********************w*****************************************

End of Section C Plant and Rajiation Monitorina Systems

                          • u*******************************x***************x****************

- .. .- - . . . ~ . - - . - - . - - - . ~ - - - - . - - - . - ~ - ~ .

l Section A Reactor Theory. Thermodynamics and Facility Characteristics
  • ANSWER-(A1) l a .
  • REFERENCE l

}- ARRR REACTOR OPERATOR TRAINING MANUAL. Volume 3. Reactor Theory page RT-4.5.

i section E. Energy Release from Fission

  • ANSWER (A2).

i a i

  • REFERENCE ARRR REACTOR OPERATOR TRAINING MANUAL Volume 3. Reactor Theory, page 1-13.

section 1.2.4. Definition of Dollar t.

!

  • ANSWER (A3)

! b

.

  • REFERENCE
ARRR REACTOR OPERATOR TRAINING MANUAL. Volume 3. Reactor Theory, page 1-31 i L* ANSWER (A4) j~ b

REFERENCE:

?

ARRR Technical Specif ation 5.3.1. i

.4.00 + 4.50 + 1.75 - 10.25 total rod worth '

[ 10.25 - (2.5'+ 4.50) - 3.25 total rod worth minus excess and max rod worth j-

  • ANSWER (A5) <

> a L

REFERENCE:

t ARRR REACTOR OPERATOR TRAINING MANUAL. Volume 4. Reactor Physics and Kinetics.

2 Sections 1.2.1. 1.2.2, 1.2.3. and Appendix A

  • ANSWER (A6) '

d '

  • REFERENCE ARRR REACTOR OPERATOR TRAINING MANUAL Volume 4. Reactor Physics and Kinetics.

-Section 1.1.11, Moderators

  • ANSWER (A7) a
  • REFERENCE ARRR REACTOR OPERATOR TRAINING MANUAL. Volume 3. Reactor Theory, pages RT-5.4 and 5.5
  • ANSWER (A8) b-
  • REFERENCE ARRR REACTOR OPERATOR TRAINING MANUAL, Volume 3. Reactor Theory, page 1-17 SCR - S/(1-Keff) SCR = 600 cps /(1-0.7) = 600 cps /0.3 - 2000 cps

,. H

[ Section A Reactor Theory. Thermodynamics and Facility Characteristics

  • ANSWER (A9)  !

, a -

  • REFERENCE ARRR REACTOR OPERATOR TRAINING MANUAL. Volume 3, Reactor Theory, page 1-5 through 1-8
  • ANSWER (A10)

, b

  • REFERENCE ARRR REACTOR OPERATOR TRAINING MANUAL, Volume 3. Reactor Theory, page 1-5

! through 1-8 n/n, = e**(t/T) 250/50 = e**[t/40 sec]

250/50 = e**[t/40]

in (5) = 1.61 = t/40 t'= 40*1.61 = 64 seconds

  • ANSWER (All) i C
  • REFERENCE

.ARRR ADMINISTRATIVE-PROCEDURE IV CRITICAL ASSEMBLY AND POWER CALIBRATION-PROCEDURES, PAGE 8 P (kw) = F*C*(To-Ti) i P = (170 gpm)*(.146)*(95-90) = 125 kw

  • ANSWER (A12)
b

!

  • REFERENCE ARRR REACTOR OPERATOR TRAINING MANUAL, VOLUME 4. REACTOR PHYSICS AND KINETICS,
Section 1.3.3, " Temperature Coefficients." pages 1-22 through 1-24 Control rod reactivity = +(0.0005 Ak/k/ inch)(2 inches)

Moderator reactivity - ( d 5

K/K/ C)(5*C)

-0.00005Ak/k For critical reactor, l 0 = Control rod reactivity.+ Moderator reactivity + Fuel reactivity, or Fuel reactivity = -(Control rod reactivity + Moderator reactivity)

' = -(0.001 Ak/k - 0.00005Ak/k) = -0.000995Ak/k Fuel temperature change = Fuel reactivity / fuel temperature coefficient
Fuel temperature change - (-0.000995Ak/k)/(-1.0 x 10 ' Ak/k/ C) = 9.95"C.

J Section- A ReactoFTheory. Thermodynamics and Facility Characteristics I

3 L* ANSWER A13

Reactor. condition

[ 1. Subcritical' _c_

c 2. Critica! _b_

a

3. Su)ercritical -~

d E *REFEREiCE-

j. 'ARRR REACTOR OPERATOR TRAINING MANUAL, VOLUME 4. REACTOR PHYSICS AND KINETICS, Section 1.2.1, " Reactivity" pages 1-10 & 1-11 L*ANSWERA14
a
*REFERFNCE
- ARRR REACTOR OPERATOR TRAINING MANUAL, VOLUME 4, REACTOR PHYSICS AND KINETICS,
Section 1.3.7 " Xenon Poisoning" pages 1-25 through 1-29
  • ANSWER A15
' ~a .
  • REFERENCE

- -ARRR REACTOR OPERATOR TRAINING MANUAL, VOLUME 4, REACTOR PHYSICS AND KINETICS,

. Sections 1.3.3, 1.3.4, 1.3.6, & 1.3.7

'* ANSWER A16 1d

  • REFERENCE-ARRR REACTOR OPERATOR TRAINING MANUAL, VOLUME 4. REACTOR PHYSICS AND KINETICS, Section l.337 " Temperature Coefficients" pages 1-22 through 1-24

'* ANSWER A17 b'

  • REFERENCE'-

ARRR REACTOR OPERATOR. TRAINING MANUAL, VOLUME 4. REACTOR PHYSICS AND KINETICS,

Section 1.-2.7 Neutron Source, page 1-17
  • ANSWER A18

.a

  • REFERENCE . .

ARRR REACTOR OPERATOR TRAINING MANUAL, VOLUME 4. REACTOR PHYSICS AND KINETICS,

'Section 1.2.5, Reactor Kinetic Behavior, page'l-14 P/Pg e**(t/T) Ln (5/1)>Ln (20/5)>Ln (50/20)>Ln (100/50)

'* ANSWER A19 c

'* REFERENCE ARRR REACTOR OPERATOR TRAINING MANUAL, VOLUME 4. Page 1-12 Table 1.1 i

Section A Reactor Theory. Thermodynamics and Facility Cha_r_gcteristics

  • ANSWER A20 a

Calibration Procedure Page 5  :

d 4

1 a

f 4

9 e

f i

i

,0 Section B Normal . Emeraency and Radioloaical Control Proce.dg.r.gg i

,

  • ANSWER B-1 c

r

  • REFERENCE 3- 'ARRR Technical Specification 12.2.2 page 12. and ARRR Administrative
- Procedures I.B.6.a, page 3 ,

[

  • ANSWER B-2 L a
  • REFERENCE
ARRR' Administrative Procedures I.B.6.d. page 4 1
  • ANSWER B-3

. b-

-

  • REFERENCE
ARRR Administrative Procedures I.C.6. page 4 1 i
  • ANSWER B-4  !

1 2. _b_ 3. _b_ 4. _d_

ARRR Administrative Procedures I.C.6. page 4 l
\

! '* ANSWER B-5 b

  • REFERENCE

' ARRR Administrative Procedures VI.F.2 page 9

  • ANSWER B-6 4-b i l
  • REFERENCE '
ARRR Administrative Procedures IV.A.3. page 1 l

}

  • ANSWER B-7 'l '

d

  • REFERENCE I ARRR Administrative Procedures IV.C.1. pages 4-7
  • ANSWER B-8

- C' I

  • REFERENCE i ARRR Administrative Procedures IV.C.2. pages 7-8 1 l
  • ANSWER B-9 b

l

  • REFERENCE .

ARRR Administrative Procedures IV.D.8. page 10

  • ANSWER B-10 C l
  • REFERENCE ARRR Technical Specification 11.5 and Administrative Procedures IV.O.9. page 10

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

Sg_t.ign 8 Normal.'Emeraency and Radioloaical , Control Procedures f

i -* ANSWER B-11 b

4

  • REFERENCE j -- ' ARRR REACTOR OPERATOR TRAINING MANUAL. -Volume.5. Radiological Safety, page 6 i
  • ANSWER B-12 C
  • REFERENCE . ..

I ARRR Administrative Procedures IV.E.2 and 6 pages.12 and 13 l

  • ANSWER B-13

! . 1. . . b ' . 2. --

a 3. --

c 4. --

d

[

  • REFERENCE-ARRR Administrative _ Procedures VI.C.1 & 2. pages 2 and 3 and 10CFR20.1201 and.
1301 4

.*ANSWER B-14 i- b U '

  • REFERENCE .

I ARRR Administrative Procedures VI.D.1, 2 & 5. pages 4 & 5 ,

' *ANSWER B-15 b

  • REFERENCE-

.ARRR Administrative Procedures VI.E.1.a. page 5 l

  • ANSWER B-16 b

i

  • REFERENCE j; ARRR Administrative Procedures VIII.A.S. page 3 .

i

  • ANSWER B;17 C <

l

  • REFERENCE' .
- ARRR REACTOR OPERATOR TRAINING MANUAL, Volume 5. Radiological Safety, pages 10

& 11. I/r**2-constant 450 mrem /hr/3ft**2-450/9-50

50*1ft**2-50 mrem /hr

!

  • ANSWER B-18

!. c l'

  • REFERENCE ,

'TS 6.2 and Table 2 2.

i-

  • ANSWER B-19.

! c I

  • ANSWER B-20
d i
  • REFERENCE' l Administrative procedure VI Section F. " Radiation Area Control" page 8 of 13 i

1,

i

)

+

i
n. l j

Section C Plant and Radiation Monitorina Sys1ggi

1

(

  • ANSWER C-1 d
  • REFERENCE

, ARRR REACTOR OPERATOR TRAINING MANUAL,' VOLUME 2. " REACTOR DESIGN FEATURES,"

! PART II, " REACTOR BUILDING AND VENTILATION SYSTEM," Section 1 " General Description," Page 11-1

{

.-

  • ANSWER C-2 l 'd'
  • REFERENCE' .  ;

!. ARRR REACTOR OPERATOR TRAINING MANUAL, VOLUME 2. " REACTOR DESIGN FEATURES " l L PART III, "THE AER0 TEST RADIOGRAPHY AND RESEARCH REACTOR (ARRR)," Section 4.9.

- " Fuel Moderator Elements,"'Page 111-13

-

  • ANSWER C-3
.C
  • REFERENCE
ARRR REACTOR 0PERATOR TRAINING MANUAL, VOLUME 2. " REACTOR DESIGN FEATURES,"

i PART IX, " ROD DRIVE MECHANISM CIRCUITS." Section 3. " Circuit Operation," Page IX-3 l

  • ANSWER C-4

. c

  • ANSWER C-5

! , Channel NL Detector i

1 a  !

2 c 3 d 4 c

!

  • Reference j Technical Specifications Tables 1 and 2 i
  • ANSWER C-6
. a f . *REFERENCE
ARRR REACTOR OPERATOR TRAINING MANUAL, VOLUME 2. " REACTOR DESIGN FEATURES,"
PART V, " REACTOR WATER SYSTEM " Section 2 " Pool Water," Page V-1

$'

  • ANSWER C-7 i; d l'
  • REFERENCE
Technical Specification 6.5, page 4 J

p

- . . - . . ~ - - . .. - . - . - . . . .. . - . - - . - . - . . - - .- .

~.,

t-  :

I

' Section C. Plant and Radiation Monitorina Systems

' i

  • ANSWER C-8' 1 d R
  • REFERENCE-l ARRR REACTOR OPERATOR TRAINING MANUAL. VOLUME 2, " REACTOR DESIGN FEATURES."

PART V. " REACTOR WATER SYSTEM " Section 3.1 " Primary Cooling Leop," Page V-3

  • ANSWER C-9

, b l

  • REFERENCE l- ARRR REACTOR OPERATOR TRAINING MANUAL, VOLUME 2. " REACTOR DESIGN FEATURES."

l PART V. " REACTOR WATER SYSTEM," Section 3.3 "Demineralizer Loop." Page V-5, 4

! Figure V-3a

  • ANSWER C-10 b

d

  • ANS 'ER C-12
  • ANSWER C-13 d
  • ANSWER C-15 c

l-

  • REFERENCE

-Technical Specification 8.2, page 6 L

  • ANSWER C-16 l C
  • REFERENCE Technical Specification 8 3, page 6 i

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

E ,

lo l: . ,

L!

t-

' Section C Plant and Radiation Monitorina Systems l y-

  • ANSWER.C-17 c- d-

! L* REFERENCE-L . Technical Specifications 8.4.2, page 6,.and 8.7, page 7..

  • ANSWER.C-18 c
  • REFERENCE-Technical Specificaticas 9:0, pages 7 & 8 t
  • ANSWER C-19 b-

REFERENCE:

ARRR Administrative Procedures VIII'. page 7

  • ANSWER C-20 l' C l *

- REFERENCE

!  ;TS 5.3.2.

o I

I-l t

{

l, l

r

~

i .

l

{

r i

i

.-.. , . a .. . 2 . - - ~ n _

.~ . . . - - - .- - -. - . . .

  • July 7, 1997 '

Mr. RIy Tsukimura, Presid;ni A$ tot:st Operations, Inc.

J 3455 Fostoria Way

, San Rarnon, Califomia 94583

SUBJECT:

INITIAL EXAMINATION REPORT NO. 50-228/OL-97-01

Dear Mr. Tsukimura:

During the week of June 2,1997, the NRC administered an initial examination to an employee of your facility who had applied for a license to operate your Aerotest Operations, Inc. Reactor. The examination was conducted in accordance with NUREG-1478, "Non-Power Reactor Operator

,i

_ Licensing Examiner Standards," Revision 1. At the conclusion of the examination, the examination questions and preliminary findings were discussed with those members of your staff identified in the enclosed report.

in accordance with 10 CFR 2.790 of the Commission's regulations, a copy of this letter and the enclosures will be placed in the NRC Public Document Room.

Should you have any questions concoming this examination, please contact me on

(301) 415-2170.

Sincerely,

. $ , cting Deputy Director S

Division of Reactor Program Management

! Office of Nuclear Reactor Regulation Docket No. 50-228

Enclosures:

1. Initial Examination Report

' No. 50-228/OL-97-01

2. Examination and answer key (RO/SRO) cc w/ enclosures:

See next page DISTRIBUTION w/ enclosures: DISTRIBUTION w/o enclosures; PUBLIC PDND R/F SWeiss SHolmes, RIV MSlosson MMendonca, PM LTremper, LFDCB (T-9 E-10)

Facility File (DMcCain) Document Control

' Docket File (50 228)

DOCUMENT NAME: G:\SECY/MENDONCA/AEROTEST.!NR i Tm receive a cop <r of thle document, indeste in the boa: *C* = Copy without attachmentlenclosure *E' = Copy with attachment / enclosure *N' = No copy i OFFICE PDND:PM %lE HQQl:lA E PDND:(A)D_Dr la l 1 NAME MMendonca Ddolfai6I SWeiss (44W DATE 07/ 7/97 07M/97 07d/97 ' "

OFFICIAltOPY 3

1

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