ML20217H879

NRC Operator Licensing Exam Rept 50-002/OL-99-01 (Including Completed & Graded Tests) for Tests Administered on 990719- 21.Exam Administered to One RO & One Sro.Ro Failed Written Portion of Exams

ML20217H879
Person / Time
Site:	University of Michigan
Issue date:	08/16/1999
From:	Isaac P NRC (Affiliation Not Assigned)
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ML20217H878	List: ML20217H873 ML20217H879
References
50-002-OL-99-01, NUDOCS 9910220219
Download: ML20217H879 (36)

Text

s U. S. NUCLEAR ~ REGULATORY COMMISSION OPERATOR LICENSING IN;TIAL EXAMINATION REPORT REPORT NO.: 50-002/OL-99 FACILITY DOCKET NO.: 50-002 FACILITY LICENSE NO.: R-28 FACILITY: University of Michigen EXAMINATION DATES July 19 - 21,1999 EXAMINER: Patri ac, Chief Examiner SUBMITTED BY: -

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3- 8 i Pai kisaa@ef Examiner ' bate

SUMMARY

During the week of July 19,1999, NRC administered Operator Licensing Examinations to one Reactor Operator (RO) and one Senior Reactor Operator Instant (SROI) candidate. The RO failed the written portion of the examinations. j i

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I ENCLOSURE 1 l 9910220219 991014 PDR ADOCK 05000002 V PDR

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.2-REPORT DETAILS

1. Examiners:

Patrick Isaac, Chief Examiner

2. Results:

RO PASS / Fall SRO PASS / FAIL TOTAL PASS / Fall Written' 0/1 1/0 ili j Operating Testa 1/0 ' ' 1/0 210 Overall 0/1 1/0 1/1 l

3. Exit Meeting:

Personnel attendingi Mr.. Christopher Walter Becker, Manager of Operations Mr. Patrick Isaac, Chief Examiner -

There were no generic concerns raised by the chief examiner Mr. Isaac expressed his appreciation to the FNR staff for their cooperation in the administration of the examination.

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U. S. NUCLEAR REGULATORY COMMISSION NON-POWER INITIAL REACTOR LICENSE EXAMINATION FACILITY: University of Michigan REACTOR TYPE: Pool DATE ADMINISTERED: 1999/07/19 REGION: lil CANDIDATE:

INSTRUCTIONS TO CANDIDATE:

l Answers are to be written on the answer sheet provided. Attach the answer sheets to the examination. Points for each question are indicated in parentheses 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 l

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. FACILITY 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 ENCLOSURE 2

A RX THEORY. THERMO & FAC OP CHARS ANSWER SHEET 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

'003. a b c d

'004 a b'c d 005 . a b c d 006 a b c d 007 a b c d , , e 008 a b - c d 009 a ~ b c d 010 a b c d 011 abcd 012 a b c d l

013 a ~ b c d i

014 a b c d j t

015 a ' b ' c d I

! I L 016 a b c d I

(- 017 a ' b c . d '

018 a , b c 'd l

' 019 a b c . d l 020 a b c d

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

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F. ,

B. NORMAL /EMERG PROCEDURES & RAD CON l s

ANSWER SHEET ,

Multiple Choice (Circle or X your choice)

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

001 abcd

- 002 . a b c d 003 ~ a b c d

-004 abcd 005 a b c d 006 a b c d 007 a b c d ' ,

008 a b c d 009 a b c d 010 a b c d 011 abcd 012 a b c d

)

013 a - b . c - d 014 ' a b c d '

015 a b c d i

016 a b c d 017 a b c d l 018 a b c d

'019 a b c d  ;

020 a b . c d i

, ("*" END OF CATEGORY B ""*)

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• C PLANT AND RAD MONITORING SYSTEMS l

l ANSWER SH EET Multiple Choice (Circle or X your choice) l- If you change your answer, write your selection in the blank.

001 abcd 002 a b c d 003 a b c d 004 ' a b c d 005 a b cd 006 a b c d 007 a b c d 008 a b c d 009 a b c d 010 a b c d -

011 abcd 012 a b c d 013 a ' b c d 014 abcd 015 a b c d 016 a b c d 017 a b

c.

d l

018 a b c d

} 019 a b c d

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

(""*"*" END OF EXAMINATION ***"*"")

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

1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties, j

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2. After the examination has been completed, you must sign the statement on the cover sheet indic 'q that the work is your own and you have neither received nor given assistance i: *;npleting 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.gnly to facilitate legible reproductions.

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

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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 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.

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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.

EQUATION SHEET o - 612 Q = m cp AT Pm , =(2a(k)f Q = m Ah SCR = S/(1-Kef0 Q = UA AT CRi (1-Kef03 = CR 2(1-Kef02 26.06 (A,rrp) (1-Kef00 SUR = M=

(p - p) (1-Kef03 SUR = 26.06/t M = 1/(1-Kef0 = CR /CRo 3

ERu)

P = Po 10 SDM = (1-Kef0/Keff '

P= Poe " Pwr = W, m p(1-p)

P= Po (* = 1 x 10 5 seconds

@-P

~

t = f*/(p Ip)

T = (f*/p) + [(p-p)/A,rrp]

p = (Keff-1)/Keff A,,, = 0.1 seconds 2 p = AKeff/Keff 0.693 T ,2 =

3

_p = 0.0070 A DR3D32 = DR 2D 22 DR = DRo e*

6CiE(n)

DR = DR = R/hr, Ci a Curies, E e Mev, R s feet 2

R I

1 Curie = 3.7x10 dps 2

1 kg = 2.21 lbm I hp = 2.54x10' BTU /hr 1 Mw = 3.41x106 BTU /hr 1 BTU = 778 ft-lbf *F = 9/5 C + 32 1 gal H2 O = 8 lbm C = 3/9 (*F - 32) l l

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)QUESTlON -(A.1)' [1.0)

A reactor is critical at 1 Watt. Subsequent rod motion causes a power increase at an indicated period of 30 seconds.' Reactor power 2 minutes later will be approximately:

' a. 55 Watts

b. . 35 Watts

c. 15 Watts

d. 5 Watts 1

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J Question (A.2) [1.0) , )

Which one of the following describes the response of the subcritical reactor to eaualinsertions of positive reactivity as the reactor approaches criticality at low power?

a. Each reactivity insertion causes a SMALLER increase in the neutron flux, resulting in a LONGER time to reach equilibrium.

b, Each reactivity insertion causes a LARGER increase in the neutron flux, resulting in a LONGER time to reach equilibrium,

c. Each reactivity insertion causes a SMALLER increase in the neutron flux, resulting in a SHORTER time to reach equilibrium. -

d. Each reactivity insertion causes a LARGER increase in the neutron flux, resulting in a SHORTER time to reach equilibrium.

Question '(A.3) '[1.0)

'Which one of the following is true concerning the differences between prompt and delayed neutrons? i s: Prompt neutrons account for less than one percent of the neutron population while delayed neutrons actount for approximately ninety-nine percent of the neutron population I

. b. Prompt neutrons are released during fast fissions while delayed neutrons are released 1 during thermal fissions j, - c. Prompt neutrons' are relea ,ed during the fission process while delayed neutrons are t

- releaseo during the decay process i 1

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Prompt neutrons are the dominating factor in determining the reactor period while delayed j

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neutrons have little effect on the reactor period  !

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f-I' Question (A.4) [1.0)

Which one of the following statements describes why installed neutron sources are used in reactor cores?

a. To increase the count rate by an amount equal to the source contribution.

b. To increase the count rate by 1/M (M = Suberitical Multiplication Factor).

c. To provide neutrons to initiate the chain reaction.

d. To provide a neutron level high enough to be monitored by instrumentation.

Question (A.5) [1.0)

A reactor contains three safety rods and a control rod. Which one of the following would result in a determination of the excess reactivity of this reactor?

a. The reactor is critical at a low power level, with all safety rods full out and the control rod at some position. The reactivi,ty remaining in the control rod (i.e. its rod worth from its present

)

, position to full out) is the excess reactivity.

b. The reactor is shutdown. Two safety rods are withdrawn until the reactor becomes entical.

The total rod worth withdrawn is the excess reactivity.

c. The reactor is at full power. The total worth of all rods withdrawn is the excess reactivity.

d. . The reactor is at full power. The total worth remaining in all the safety rods and the control rod (i.e. their worth from their present positions to full out) is the excess reactivity.

Question (A.6) [1.0)

Which one of the following statements concerning reactivity values of equilibrium (at power) xenon and peak (after shutdown) xenon is correct? Equilibrium xenon is of power level; peak xenon is of power level.

a. INDEPENDENT INDEPENDENT

b. INDEPENDENT DEPENDENT

c. DEPENDENT INDEPENDENT

d. DEPuNDENT DEPENDENT I

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r Questi:n (A.7)l [1.0).

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. Reactor A increasee power from'10% to 20%'with a period of 50 seconds. Reactor B increases i

power from 20% to 30% with a period of also 50 seconds. Compared to Reactor A, the time required for the power increase of Reactor B is:'

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- a'. longer than A.

b. exactly the same as A.

c. twice that of A.

d. shorter than A.

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. Question (A.8) [1.0).

1 A control rod was withdrawn two (2) inches. The steady reactor period following rod withdrawal is observed to be sixty (60) seconds.

Which one of the following is the differential rod worth?

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a. 1.0 x 10'8 delta k/k per inch -

b. 5.6 x 108 delta k/k per inch

c. 1.12 x 104 delta k/k per inch

d. 5.0 x 10d delta k/k per inch Question (A.9)- [1.0)

A reactor is suberiticalwith a shutdown margin of 0.0526 delta k/k. The addition of a reactcr

- experiment increases the indicated count rate from 10 cps to 20 cps. Which one of the following is the new keff of the reactor?

a. .53

b. .90

c. .975

d. 1.02

o Questi:n (A.10) [1.0)

The reactor is operating at 100 KW. The reactor operator withdraws the control rod allowing power to increase. The operator then inserts the same rod to its original position, decreasing power. In comparison to the ro ndrawal, the rod insertion will result in:

a. a slower period due to long lived delayed neutron precursors.

b. a faster period due to long lived delayed neutron precursors.

c. the same period due to equal amounts of reactivity being added.

d. the same period due to equal reactivity rates from the rod.

Question (A.11) [1.0)

Which one of the following is the principal source of heat in the reactor after a shutdown from extended operation at 100 KW7

a. Production of delayed neutrons

b. . Suberitical reaction of photoneutrons

. c. Spontaneous fission of Um

d. Decay of fission fragments i

Question (A.12) [1.0]

Which one of the following factors is the most significant in determining the differential worth of a control rod?

a. The rod speed,

b. Reactor power.

c. The flux shape.

d. The amount of fuel in the core.

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e Questi:n (A.13) L[1.0)

-2 i Which one of the following is the MAXIMUM amount of reactivity that can be promptly inserted into the reactor WITHOUT causing the reactor to go " Prompt Critical *?

a. 0.10 dollars

'bc 0.46 dollars -

c. 0.75 dollars

d. 1.90 dollars : 1 i

Question (A.14) [1.0) 1 A rod is inserted into a critical reactor with a reactivity of -0.003 AK/K. Which one of the following is the final stable reactor period? f l

a. -100 seconds I

b. -90 seconds

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c. -80 seconds )

'd. -70 seconds l

l Question (A.15). [1.0).

Which one of the following conditions would INCREASE the shutdown margin of a reactor?

l a: . Inserting an expcriment adding positive reactivity.

b. Lowering moderator temperature if the moderator temperature coefficient is negative.

c. Depletion of a burnable poison.

d. Depletion of uranium fuel.

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Question (A.16)- '[1.0)

- A reactor with an initial populaticri of 24000 neutrons is operating with K., = 1.01. Of the

CHANGE in population from the current generation to the next generation, how many are prompt

neutrons?.

a. 24

. b. 238

. c. 240

c. 24240 Question (A.17) [1.0)

Following a significant reactor power increase, the moderator temperature coefficient becomes increasingly more negative. This is because:

a. as moderator density decreases, less thermal ne'utrons are absorbed by the moderator than by the fuel.-

b. the change in the thermal utilization factor dominates the change in the resonance escape probability,

c. a grenier density change per degree F occurs at higher reactor coolant temperatures.

d. the core trs nsitions from an under-moderated condition to an over-moderated condition. l i

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Questbn (A.18) [1.0].

. The following data was obtained during a reactor fuel load.

No. of Elements Detector A (cos) '

O 20-8 30' 16- 50 24 150 28 4000 Which one of the following is the closest number of fuel elements required to make the reactor critical? . (The attached figure may be used to determine the correct response.)

a. 16 i'

b. 28-

c. 32 id. 40

• 1.0 0.9 0.8 0.7 I 1 0.6 M 0.5 c.4 0.3 0.2 0.1 0.0 2 4 6.

8 10 12 14 16 18 20 22 24 26 28 30 32 l

NUMBER oF ELEMENTS INSTALLED

. Question (A 19) [1.0]

Which one of the following will be the resuhing stable reactor period when a $0.25 reactivity  !

insertion is made into an exactly critical reactor core?

a. 18 seconds '

b. 30 seconds 1

! ' c. 38 seconds

d. 50 seconds I

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l. Quistion (A.20) [1.0) i- Which alteration or change to the core will most strongly affect the thermal utilization factor.

, s. Build up of fission products in fuel.

b. Removal'of moderator.. -)

c.. Addition of U"

d. Ren3 oval of a control rod.

(*" End of Section A *")

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. I Section B Normal /Emera. Procedures & Rad Con

' QUESTION (B.1) [1.0]

In order to ensure the health and safety of the public,10CFR50 allows the operator to deviate from Technical Specifications. What is the minimum level of authorization needed to deviate from Tech. Specs?

a. USNRC

b. Reactor Supervisor

c. 1 Licensed Senior Reactor Operator. '

d. Licensed Reactor Operator.

• QUESTION (B.2) [1.0)

Safety Limits are ...

a. limits on very important process variables which are found to be necessary to reasonably protect the integrity of certa!.1 physical barriers which guard against the uncontrolled

• release of radioactivity.

l b. settings for automatic protective devices related to those variable having significant safety l l functions. l

c. settings for ANSI 15.8 suggested reactor scrams and/or alarms which form the protective system for the reactor or provide information which requires manual protective action to be I initiated.

d. the lowest functional capability or performance levels of equipment required for safe j operation of the reactor.

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• QUESTION (B.3) [1.0)

Based on the Requalification Plan for operators, each licensed operator must complete a minimum of reactivity control manipulations during each 2 year cycle.

a. 4

b. 10 20

d. 28

Section B Normal /Emera. Procedures & Rad Con

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• QUESTION (B.4) . [1.0) 1 Which one of the following instruments should you use to survey a gamma source?

a. Thin window ion c'hamber.

b. GM tube.

c. lon chamber (open window).

d. Neutron ball.

• QUESTION (B.5) -[1.0)-

What is the best type of shielding material to protect from a thermal neutron beam?

a. Lead

b. ' Heavy clothing

c. Rubber

d. Boron-10

• QUESTION (B.6) [1.0) 1 Which one of the following is the definition of Committed Dose Equivalent? .j a, The sum of the deep dose equivalent and the committed effective dose equivalent. l l

b. The dose equivalent that the whole body receives from sources outside the body.

c. The sum of the external deep dose equivalent and the organ dose equivalent. )

d.' The 50 year dose equivalent to an organ or tissue resulting from an intake of radioactive material.  !

• QUESTION (B.7) [1.0) -

P = 2.60 Mw (max) T, = 129'F (max) H = 19 ft (min)

The above variables associated with core thermal and hydraulic performance are:

a. ' Safety' limits in the natural convection mode.

b. ' Safety limits in the forced convection mode,

c. ~ LSSS in the natural convection mode.

d. LSSS in the forced convection mode. ,

Section B Normal /Emera. Procedures & Rad Con

• QUESTION (B.8) . [1.0)

Select the MINIMUM amount of time that must be spent performing license activities in order to maintain your license active.

a.' 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per month

b. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per month

c. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per quarter

d. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per quarter

• QUESTION (B.9) [1.0)

When a radioactive material was removed from the core, it read 25 rem /hr gamma at a distance of 5 feet. Five hours later it read 1.5 rem /hr at 5 feet. How long after the material was removed from the core would it take for the sample to decay to 200 mr/hr?

a. 1.23 hr

b. 4.52 hr

c. 6.97 hr

d. 8.57 hr l

• QUESTION (B.10) [1.0) '

Which one of the following lists automatic actions associated with the Technical Specification j Safety Limits for the natural convection mode? d

a. Reactor thermal power, Reactor coolant inlet temperature Height of water above the top of the core.

b. High Power / Header Down. i Reactor coolant exit temperature. I Height of water above the center line of the core.

c. Reactor thermal power.

Height of water above the top of the core.

Pool sater temperature.

'd. Reactor coolant exit temperature.

Height of water above the top of the core.

Primary coolant flow through the core.

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Section B Normal /Emera. Procedures & Rad Con

r.  ? QUESTION (B.11) [1.0]

3 During reactor operation, which one of the following conditions violates confinement integrity?

a. The reactor building ventilation supply fan is off.

b. The access hatch from grade level to the beamport floor is temporarily opened for the

. pass' 2e of personn61.

c. The beamport exhaust damper to Stack 2 is tagged shut.

d. -

The beamport floor door to the Phoenix Memorial Laboratory is clamped closed.

• QUESTION (B.12)L [1,0]

Which one of the following meets the definition of a Channel Check per the Technical Specifications?

a.

Immersing a' temperature detector in an ice bath a' nd verifying a reading of 32' F.

b.' Piacing a source next to a radiation detector and observing meter movement.

c.

Performing a precise determinstion of reactor power, then adjusting reactor power meters to correspond to correct power.-

d. Comparing the NM-1000 to the NP-1000 to verify proper overlap.

- QUESTION (B.13) [1.0]

A room contains a source which, when exposed, results in a general area dose ate of 175 r illirem per hour. This source is scheduled to be exposed continuously for 25 days. Select an acceptable method for controlling radiation exposure from the source within this room.

a. Post the area with words

• Danger-Radiation Area".

b. Equip the room with a device to visually display the current dose rate within the room.

c. Equip the room with a motion detector that will alarm in the control room,

d. - Lock the room to prevent ihadvertent entry into the room.

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Section B Normal /Emera. Procedures & Rad Con-

• QUESTION (B.14) [1.0]

The reactor is in steady-state power at 90% when you, the operator, notice that the Reactor Bridge area radiation monitor is inoperabie. Which one of the following describes the correct

action you should take?

a. Shutdown the reactor. Technical Specifications (T.S.) do not allow operations of the reactor without a fully operating Reactor Bridge radiation monitor, b._ Continue operation.. T.S. allow the unit to be out of service for up to 7 days.

c. Continue operation. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of recognition of failure, replace the unit with a portable gamma-sensitive instrument with alarm.

d. Continue operation as long as a minimum of three other area radiation monitors are operating.

• QUESTION (B.15) [1.0)

Which one of the following statements is TRUE concerning experiments? '

a. An experiment approved for control fuel element irradiation may be irradiated in a polyethylene container if the irradiation is for six hours or less.

b. The reactivity worth of any moveable experiment shall not exceed 0.01 AK/K.

c. Experiments approved for the sample stringer can be loaded or unloaded from the reactor ,

only during reactor shutdown periods.

d. The total reactivity worth of moveable and secured experiments shall not exceed 0.0436 AK/K.

• QUESTION ~(B.16) [1.0)

Which one of the following statements is applicable when fuel handling is in progress at the FNR?

a. During fuel movement, the fission chambers should be moved only when direct communication is established between the control room and the bridge.

b. During fuel movement, all heavy water transfers should be supervised by a SRO other than the Fuel Movement Coordinator.

c. The control console operator grants specific permission to unlatch the fuel tool from elements inserted into the core.

d. The control console operator will be responsible for maintaining the Fuel Logbook.

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Section B Normal /Emera. Procedures & Rad Con 1

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• QUESTION (B.17). [1.0)

A calorimeter is conducted at 100% indication on the Linear Level 1 Mw scale. Actual power based on the calorimeter is determined to be 1.3 Mw. What would be the correct linear level setpoint for 1 Mw.

a. 103%

b. 95.7 %

cc 90.9% -

.d. 76.9 %

• QUESTION (B.18) [1.0)

A fire has occurred at the FNR and a building evacuation is in progress in accordance with EP-101, Reactor Building Emergency. The control room operator has arrived in the lobby and notes that the plant operator has not arrived in the lobby within a reasonable amount of time. Which one of the following actions should the control room operator take?

a. Proceed to the south side entrance door and await the fire department.

b. Attempt to locate the plant operator by retracing the plant operator's prescribed route.

c. Dial 911 and report to Public Safety and Security that the plant operator has not yet reported to the lobby.

d. Await the fire department in the lobby and inform them that the plant operator has not yet reported.

' QUESTION - (B.19) [1.0). . _

A reactor startup is in progress in accordance with OP-101, Reactor Startup. All shim-safety rods i

- have just been withdrawn to the shim range point. Which one of the following indications should be observed on the LCR7 -

' a.' Count rate should have increased by less than two (2) decades, indicating that the reactor i is suberitical.  !

b.- Count rate should have increased by less than two (2) decades indicating that the reactor j is critical on delayed neutrons.

c. Count rate should have increased by greater than two (2) decades indicating that the compensated ion chambers are functioning properly.

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d. Count rate should have increased by greater than two (2) decades indicating the reactor is supercritical with less than a 30 second period.

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i Section B Normal /Emera. Procedures & Rad Con

• QUESTION (B.20) [1.0)

It has jtrt been brought to your attention that a release of I-131 at the reactor stack discharge of 2x104 uCi/ml has been detected before dilution.10CFR20 lbts 10 ' UCi/ml as permissible. What do you do?

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a. Scram the reactor. A T.S. limit has been exceeded. This constitutes a Reportable  !

Occurrence.

b. Shutdown the reactor. The Nuclear Reactor Laboratory Manager shall be notified of the occurrence.

c. Notify the on-call supervisor. The discharge is below 10CFR20 limits.

d. Secure the reactor. Initiate evacuation. Close the stack 2 exhaust damper. j i

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(*** End of Section B *")

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1 Section C Plant and Rad Monitorina Systems QUESTION (C.1) [1.0) {

The Phoenix Laboratory and Reactor Building have experienced a temporary loss of electrical power.

l Which one of the following components must be restarted by first closing the circuit breaker and l then resetting its mercury switch?

a. MAP monitor vacuum pump i

b. Steam recirculation pump

c. Reactor building exhaust fan 1

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d. Laboratory stack exhaust fan

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QUESTION (C.2) [1.0) '

in the event of a reactor scram, the momentum o'f the descending rod is absorbed on the

a. fuel element I

b. lower shock

c. rod guide block

d. holdown QUESTION (C.3) [1.0)

The active fuel meat is composed of

a. Uranium silicide.

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b. Uranium hydride

c. Meta llic uranium

d. Uranium aluminide i

QUESTION (C.4) [1.0)

Which one of the following will cause an auto rundown?

a. Control rod fullin (in auto).

b. Period is 20 seconds.

c. Temperature is 125'F.

- d. Poollevelis 10 inches low.

Section C Plant and Rad Monitorina Systems QUESTION (C.5)_ [1.0]

Which one of the following is NOT a shim-safety rod withdrawalinhibit?

a. Magnet power off,

b. Shim-safety rods in shim range.

c. Automatic rundown.

d. No magnet us.. act.

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QUESTION (C.6) [1.0]

The reactor is operating at 2 MW with Primary Pump number 1 providing forced circulation flow. ,

The inlet butterfly valve to the Holdup Tank is moved from its fully open position.

Which one of the following describes the impact of these events on the plant?

a. The reactor will scram on high power /iow flow due to the Primary Pump tripping on Ipw Net Positive Suction pressure.

b. The reactor will scram due to the inlet butterfly valve being off its fully open seat,

c. The Holdup tank will be slowly pumped dry resulting in a high radiation condition due to insufficient N-16 holdup time.

d. Reactor power will lower slightly due to negative temperature coefficient of reactivity effects resulting from the lower flow conditions.

1 QUESTION (C.7) [1.0]

Which one of the following is the reason for the poollevel auto rundown setpoint?

a. To provide an acceptable safety margin to the maximum fuel cladding temperature,

b. To prevent incipient boiling event if transient power rises to the thermal power trip limit.

c. To assure that an adequate pool volume is available to provide cooling of the core in the event of a loss a coolant accident.

d. To maintain an adequate pool level for the dash-pot action of the control blades in the ever.. of a scam.

Section C Plant and Rad Monitorino Systems

~.

( ,.

- QUESTION . (C 8) [1.0]-

Which one of the following radiation monitors actuates the stack alarm?

~ a. FNR GAD

b. Stack 2 MAPP c.- Pool Floor MAPP d.- Beamport Floor MAPP

- QUESTION . (C.9) -[1.0] _

' The north wall and the northeast column area monitors are unexpectedly alarming, what action should you take in this situation?

.a. . Scram the reactor . -

b. Shutdown the reactor by driving in all rods

- c. Notify the HP for an area survey 1 Enter the building emergency procedure

- d.-

QUESTION (C.10) [1.0]

After the shim safety rocs reach shim rarige, why is the control rod withdrawn to approximately 5 inches?

a. Ensure power stabilizes.

b. Minimize binding due to fuel swell.

L c. Improve reactivity controlin the operating zone.

d .' Stay within T.S. departure from nucleate boiling limits.

QUESTION . (C.11) [1.0]

V"1ich one of the following beampods cannot be drained?

a. J port i

b. E port

c. H port '

i-l d. C port i

i i

r

Section C Plant and Rad Monitorina Systems

) QUESTION (C.12) ' [1.0)

Which one of the following scram signals is defeated by bypassing the interlock scram reset section of the magnet current control?

a. Low poollevel.

b. High power / header down

c. Beamport door open.

d. Building exhaust high radiation level.

QUESTION (C.13) [1.0)

The nitrogen supply system pressure is decreasing slowly. Which one of the following describes how backup supply is initiated?

a. The Bottle Gas System con)es on line automatically when pressure on G-3 decreases

.below 20 psig. - '

b. The Bottle Gas System is manually aligned when the Nitrogen tank pressure decreases below 40 psig.

c. The Bottle Gas System comes on line automatically when the Nitrogen tank pressure decreases below 50 psig.

d. The Bottle Gas System is manually aligned when pressure on G-3 decreases below 20 psig.

! QUESTION (C.14) [1.0)

Which one of the following would be indicative of a secondary heat exchanger tube rupture?

, a. Recurring HIGH WATER LEVEL HOT SUMP alarms; increasing readings on the pool  !

conductivity recorder.

! b. Recurring HIGH WATER LEVEL HOT SUMP alarms; decreasing readings on the pool conductivity recorder,

c. Lowering poollevels, increasing secondary system pH readings,

d. ' Lowering pool levels, decreasing secondary system pH readings.

Section C Plant and Rad Monitorina Systems QUESTION (C.15) [1.0)

Which one of the following will result in a reactor scram and a ventilation system isolation?

a. 5.2 mrem /hr in the fuel vault

b. 1.6 mrem /hr in PML exhaust stack 2

)

c. 0.5 mrem /hr in the building supply header

d. Deenergizing the FNR main exhaust fan QUESTION (C.16) [1.0)

Which one of the following describes the consequences of exceeding 2.5 MW on ONLY safety chrennel A7

a. Only ONE scram signalis generated turning off magnetic power to Channel A.

b. TWO scram signals are generated, one deenergizing magnetic power to Channel A and the other deenergizing magnetic power to all three channels.

c. A SAFETY AMPLIFIER TROUBLE ALARM is generated.

d. A reactor' Auto Rundown signaiis generated.

QUESTION (C.17) [2.0)

Match the Radioactivity Monitors listed in Column I with the type of detector listed in Column 11.

Note: The items in Column 11 may be used once, more than once, or not at all.

Each response is worth 0.5 points.

Column I Column 11 Radioactivity Monitors Detector Type

a. FNR Building Exhaust 1. Proportional counter

b. Beamport Floor MAP 2. Geiger Mueller (GM)

c. PL-4 Hood 3. Compensated ion chamber

d. Fuel Vault 4. Nal Scintillation

r Section C Plant and Rad Monitorina Systems-QUESTION (C.18) _

[1.0]

The bus transfer, which switches the emergency supply from the normal building supply to the emergency generator, is located in:

a. Beamport floor

b. Room 2074

c. Room 1033

d. Room 2077 QUESTION (C.19) [1,0] - -

Which one of the following will limit the loss of coolant in the event of a coolant leak due to rupture f of a beamtube while the beamport is being used for an experiment?

a. Beamport shield door.

b. Flanges at each end of beamport. '

c. Damage control plugs.

d. Collimator.

("* End of Section C *")

4

Section A ' R Theory. Thermo & Fac. Doeratina Characteristics Page 22 ANSWER: (A.01) a REFERENCE P=P,e" = 1e'2 "** = 54.6

' ANSWER: (A.02) .

b REFERENCE I

Giasstone, S. and Sesonske, A, Nuclear Reactor Engineering, Kreiger Publishing, Malabar, . {

Florida, 1991,' $$ 3.161 - 3.163, pp.190 - 191, _ .

Burn, R., ~lntroduction to Nuclear Reactor Operations, C 1988, Chapt. 5, pp. 5 5-28.

l 1

ANSWER: (A.03) c; REFERENCE

' Lamarsh, J.R., Introduction to Nuclear Engineering, - 1983. $ 7.1, pp. 280 - 284 Burn, R.~ Introduction to Nuclear Rec , .or Operations, C 1982, $$ 3.2.2 - 3.2.3, pp. 3 3-12.

)

ANSWER: (A.04) d REFERENCE Standard NRC Question

' Burn, R., @ 1982, $5.2, p. 5 '

ANSWER: (A.05) a-REFERENCE T.S. Definition 1.8 Burn, R., Introduction to Nuclear Reactor Operations, @ 1988, $$ 6.2.1, pp. 6-2.

i

)

' ANSWER:(A.06) d~

l REFERENCE' l l Lamarsh, J.R., Introduction to Nuclear Engineering, 1983. $ 7.4, pp. 316 - 322.  !

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

i i

ANSWER: (A.07) d REFERENCE The power of reactor A increases by a factor of 2, while the power of reactor B increases by a '

factor of 1.5. Since the periods are the same (rate of change is the same), power increase B  ;

takes a shorter time.

Section A- R Theory. Thermo & Fac. Ooeratina Characteristics Page 23 ANSWER: (A:08) ,

d REFERENCE Burn, R., Introduction to Nuclear Reactor Operations,

• 1988. I 7.2 & 7.3, pp. 7 7 9.

T b"~P h" I' AeW p p - (A t) d p.= 0.007/((0.1

• 60) + 1) = 0.007/7 p = 1 x 10-8 delta'k/k p/ inch = 1 x 10 8 delta k/k/2 inches = 5.0 x 10d delta k/k per inch ANSWER: (A.09) ,

c l

REFERENCE:

1 SDM = 1-Ks% - 4 = 1/SDM + 1 -

% = 1/0.0526 + 1 -

4 =.95 CR,/CR2 = (1'. Ke2) / (1 - K ) - 10/20 = (1 - Ke2) / (1 - 0.95) ,

(0.5) x (0.05) = (1 - Km). -

Km = 1 - (0.5)(0.05) = 0.975 ,

'l ANSWER: (A.10) a l

REFERENCE l ' Burn. R., Introduction to Nuclear Reactor Operations,

• 1982, if 3.2.2 - 3.2.3, pp. 3 7 12. )

l- ANSWER: (A.11)

'- d.

REFERENCE Burn, R., Introduction to Nuclear Reactor Operations, @ 1982, G 4.9, pp. 4 4-26.

ANSWER: (A.12) c REFERENCE Lamarsh, J R., Introduction to Nuclear Engineering,1983. $ 7.2, p. 303. ,

Burn, R., Introduction to Nuclear Reactor Operations, @ 1982, 9 7.2 & 7.3, pp. 7 7-9.

- ANSWER:(A 13) e REFERENCE Glasstoni S. and Sesonske, Nuclear Reactor Engineering,1991, p. 264.

ANSWER: (A.14) c REFERENCE =

Intro. to Nuclear Operation, p. 4-16

r 7 .Section A R Theory. Thermo & Fac. Ooeratina Characteristics Page 24 ANSWER: (A.15) d

-REFERENCE 1

Burn, R., Introduction to Nuclear Reactor Operations, @ 1988, G 6.2.3, p. 6-4.

ANSWER: (A.16)'

b REFERENCE ~

' 24000 neutrons in current generation

• 1.01 = 24240 neutrons in next generation 240 neutrons added - 0.7% delayed neutron fraction = 238 prompt neutrons added ANSWER: (A.17) c REFERENCE Burn, R., Introduction to Nuclear Reactor Operations, @ 1982, G 6.4.1, pp. 6-5.

ANSWER: (A.18)'

b REFERENCE Burn, R., introduction to Nuclear Reactor Operations, c 19g2, g 5.5, pp. 51g - 5 25.

ANSWER: (A.19) b' REFERENCE Glasstone, S. and Sesonske, A, Nuclear ReactorEngineering, 1991,$5.18,p.234.

T = (E-p)/Ap T = (.0070 .00175)/.1 x .00175 = 30 seconds ANSWER: (A.20) d REFERENCE

. Lamarsh, J.R., Intro'duction to Nuclear Engineering, - 1983. $ 7 2, p. 300 Burn, R., Introduction to Nuclear Reactor Operations, C 1982, 9 3.3, pp. 3 3-18.

(*" End of Section A *")

l

. s.

P L

i l "

i Section B Normal /Emera Procedures & Rad Con Page 25

'

• ANSWER (B.1)

c.  ;

L REFERENCE ,

l 10CFR50.54(y)

• ANSWER (B.2) a REFERENCE )

TS Definitions pg. 3

• ANSWER (B.3) c REFERENCE-Requalification Plan

• ANSWER (B.4) b REFERENCE , .

Nuclear Power Plant Health Physics and Radiation Protection, Ch.10

'

• ANSWER (B.5) d REFERENCE

]

Nuclear Power Plant Health Physics and Radiatiren Protection, Ch. 9 '

• ANSWER (B.6) l d '

REFERENCE 10CFR20.1003

• ANSWER (B.7) d REFERENCE T.S. 2.2.1

• ANSWER (B.8) c

. REFERENCE 10 CFR 55.53 I

• ANSWER IB.9) d REFERENCE Nuclear Power Plant Health Physics and Radiation Protection, Ch. 4 I

• ANSWER (B.10)

C h REFERENCE l- T.S. 2.2.2

E

l. -

I Section B' Normal /Emera. Procedures & Rad Con. Page 26 l

o . .

• ANSWER (B.11) b REFERENCE T.S. $ 3.5 l

l ' ANSWER . (B.12) b REFERENCE Emergency Plan, Section 2.0

• ANSWER (B.13) d

.. REFERENCE Health Physics Manual pg.13 Sect. 8.5

• ANSWER .- (B.14) a REFERENCE

• T.S. Table 3.2, pg.13

• ANSWER (B.15) c REFERENCE OP-104 Reactor Experiments and Cobalt-60 Irradiations, Sect. 4.7.1 T.S. 3.1

• ANSWER (B.16) c REFERENCE Admin. Proc. #301 - Reactor Fuel; Sect. 6.12.4

• ANSWER (B.17).

d.

REFERENCE (B.11)

Exam Question Section B pg. B-333

' ANSWER (B.18) b REFERENCE EP-101, Reactor Building Emergency, p. 5-6 '

• ANSWER (B.19) a REFERENCE ,

Reactor Startup, OP-iO1, p. 4.

l

• ANSWER (B.20) l c REFERENCE

. Exam. Question section B pg. B-592

(*" End of Section B "*)

rn

'Section C Plant and Rad Monitorina Systems- Page 27

'

• ANSWER (C.1)

-d REFERENCE

~

OP-301, Building Power Failure, p. 2

• ANSWER (C,2) d REFERENCE System Description pg. 3 ' ANSWER (C.3) d REFERENCE FNR System Descriptions, Ch. 2

• ANSWER (C.4) a REFERENCE FNR System Descriptions, Ch.13 *

• ANSWER (C.5) b REFERENCE FNR System Descriptions, Ch.13

• ANSWER (C.6) b

. REFERENCE.

FNR System Descriptions, p. 4-3.

• ANSWER (C.7)

-b REFERENCE SAR; Abnormal Loss of Coolant; pg. 57

• ANSWER' (C.8) b REFERENCE FNR System Descriptions, Ch.13

• ANSWER (C.9) b REFERENCE FNR Exam Question Bank Sect. C pg. C-300

• ANSWER (C.10) e REFERENCE FNR Question Bank Sect. C pg. C 302 I i

m ,

f Section C Plant and Rad Monitorina Systems Page 28 v

• ANSWER . - (C.11)

.d j

REFERENCE Op-108 Step 5.1.2

• ANSWER (C.12) c

' REFERENCE Instrumentation and Control; pg.13-39

• ANSWER (C.13) a REFERENCE

. OP-302," Nitrogen Supply System Operating Procedure", pg. 2

" ANSWER (C.14) a REFERENCE FNR System Descriptions, Chapter 4 and 5.

• ANSWER (C.15) a REFERENCE FNR System Description, p.12-1

• ANSWER (C.16) b REFERENCE NRC Exam administered in 1993

' ANSWER (C.17)

a. 2 b.' 2

c. 2

d. 2 REFERENCE FNR System Description, Chapter 12, p.12-3 i FNR System Description, Chapter 13, p.13-35 l

l

• ANSWER - (C.18)

-b REFERENCE FNR System Description, Chapter 9, p. 9-1 i

• ANSWER (C.19) d REFERENCE -

SAR; Loss of Coolant Analysis: 14.2.4 Beamports l

(*** End of Section C ***)