Text

Exam Rept 50-186/OL-93-01 on 930726-27.Exam Results: Candidate Passed Written & Operating Portions of Exam
	ML20056G058
Person / Time
Site:	University of Missouri-Columbia
Issue date:	08/12/1993
From:	Caldwell J, Doyle P; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20056F883	List: ML20056F882; ML20056G058;
References
	50-186-OL-93-01, 50-186-OL-93-1, NUDOCS 9309010231
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ENCLOSURE 1 i

U. S. NUCLEAR REGULATORY COMMISSION OPERATOR LICENSING INITIAL EXAMINATION REPORT f

REPORT NO.:

50-186/0L-93-01 FACILITY DOCKET NO.:

50-186 i

FACILITY LICENSE NO.:

R-103 i

i FACILITY:

University of Missouri - Columbia EXAMINATION DATES:

July 26-27, 1993 j

EXAMINER:

Paul Doyle, Chief xaminer i

SUBMITTED BY:

2-/2 ~ 9J I

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'oy e, ie Examiner Date j

._ g/ Dtte/d!71 f

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APPROVED BY:

Jarr >s L. Ca l we 1, ' Chief i N -Power Reactor Section erator Licensing Branch pivision of Reactor Controls and Human Factors Office of Nuclear Reactor Regulation

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SUMMARY

i The NRC administered an NRC licensing examination to one Reactor Operator candidate. The candidate passed both the written and operating test portions of the NRC administered examination.

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1 9309010231 930812

.PDR ADOCK'05000186 V

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REPORT DETAILS j

1.

Examiners:

Paul Doyle, Chief Examiner 2.

Results:

1 R0 SRO Total l

(Pass / Fail)

NRC Grading:

1/0 0/0 1/0 3.

Written Examination:

The written examination was administered on July 27, 1993 to one R0 candidate.

The R0 candidate passed this portion of the examination.

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4.

Operating Tests:

i An Operating Test was administered on July 26, 1993 to one R0 candidate.

The R0 candidate passed this portion of the examination.

5.

Exit Meeting:

Personnel Attending:

Paul Doyle, Chief Examiner, NRC

' Walt Meyer, Facility Director Univ. of Missouri Research Reactor Mr. Doyle thanked Mr. Meyer for his support in the administration of the examination, and emphasized the need to promptly submit any comments on the written examination to facilitate final grading of the examination.

The NRC received the facility's comments on August 11, 1993.

See enclosure 2 for a listing of the facility's comments along with NRC resolutions to those comments.

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Facility Comments and NRC Resolution

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Section A

QUESTION (A.2)

[1.0]

l The reactor has been run for a short period of time at 1 Mwatt before being shutdown (Equilibrium Xenon conditions). An experiment worth -2.0 x 10'3 oK/K e

is REMOVED from the reactor. When the reactor is restarted two days later pool temperature is 5'F WARMER. What will be the difference in the rod height when the reactor is returned to 1 Mwatt (equilibrium Xenon conditions)?

a.

0.175% AK/K less rods must be withdrawn b.

0.265% AK/K less rods must be withdrawn

[

t c.

0.056% oK/K more rods must be withdrawn i

t d.

0.119% oK/K more rods must be withdrawn i

ANSWER (A.2) b

-2.0 x 10~3 AK/K, a - 1.3 x 10 AK/K/*F 4

an increase o,f 5'F is equivglent to 5 x (+1.3 x 10 ) = +6.5 x 10" Given:

4

+2 x 10'3

+ 6.5 x 10" - 2.65 x 10' or 0.265

REFERENCE (A.2)

INTRODUCTION TO NUCLEAR REACTOR OPERATIONS, Reed Robert Burn o 1982, 54.4, t

equation 4.15.

Facility Comment The facility comment is to confirm discussion with [the) examiner that for "b" to be correct the pool temperature needs to become 5*C WARMER. The pool has a temperature coefficient.

NRC Resolution The original question had the pool temperature as 5*C COOLER, the examiner had the candidate change COOLER to WARMER durina the administration of the examination.

No change to grading is required.

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oQUESTION (A.6)

[1.0]

The reactor is operating at 9 Mwatts (90%) and the scram setpoint is set at i

175%. What will be the resulting peak power if an experiment inserted into the reactor causes a 100 millisecond period and the scram delay is 0.1 second?

a.

14 Mwatts b.

28 Mwatts c.

34 Mwatts d.

50 Mwatts

ANSWER (A.6) c Standard Power equation 0.1see P=P e * = (1. 2 5Nwa t ts) e 01 o

P = 33.98 Mwatts = 34 Mwatts.

REFERENCE (A.6)

INTRODUCTION TO NUCLEAR REACTOR OPERATIONS, Reed Robert Burn C 1982, k5.5, example 5.5(c)(a).

Facility Comment The facility comment is to confirm with [the] examiner that the question should read: The reactor is operating a 9 Mwatts (90%) and the scram setpoint is set at 125%. What will be the resulting peak power if an experiment inserted into the reactor causes a 100 millisecond period and the scram delay is 0.1 seconds? No change to grading is required.

All possible answers should be a factor of 10 higher.

a.

14 Mwatts t

b.

28 Mwatts c.

34 Mwatts d.

50 Mwatts NRC Resolution The original question had the scram delay as I second and the 90% power level as 900 Kwatts.

Both of these values were corrected by the examiner during the administration of the examination to 0.1 second and 9 Mwatts respectively.

Also_ all four answers were increased by a factor of 10. No change in grading is required.

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oQUESTION (A.9)

[1.0]

Whigh ONE of the following is the time period in which the maximum amount of Xe" will be present in the core?

a.

7 to 11 hours1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months after a startup to 100% power b.

3 to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a power increase from 50% to 100%.

c.

3 to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a power decrease from 100% to 50%.

d.

7 to 11 hours1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months after a scram from 100%

ANSWER (A.9) a

REFERENCE (A.9)

INTRODUCTION TO NUCLEAR REACTOR OPERATIONS, Reed Robert Burn 0 1982, 98.4 i

Table 8.4.

Facility Comment The facility comment is the max'imum Xe"5 will occur 7 to 11 hours1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months after a scram from 100%.

This makes answer "d" the correct response.

NRC Resolution Agree with facility comment. Answer key changed to d.

QUESTION (A.17)

[1.0]

i Which change to the core will most strongly affect the Thermal Utilization Factor?

i a.

Build up of fission products in the fuel.

b.

Removal of moderator. (Thermal expansion) c.

Going from High Enrichment fuel to Low Enrichment fuel.

d.

Removal of a control rod

ANSWER (A.17) d

REFERENCE (A.17)

INTRODUCTION TO NUCLEAR REACTOR OPERATIONS, Reed Robert Burn C 1982, D3.3.1 Neutron Multiplication Facility Comment The facility comment is to confirm discussion with [the] examiner that the answer should read " Removal of a control rod."

This wording change was necessary to distinguish the shimming out of a control rod. The ability to distinguish the best answer between "a" and "d" would have required knowledge of how much the rod was shimned.

NRC Resolution This change was made by the examiner to the candidate during the I

administration of the examination. No change to grading is required.

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Section C

QUESTION (C.13)

[2.0 (0.4 each)]

List the NORMAL (10 Mwatt operation) positions (0 pen, Shut) for the following valves:

l a.

509 (Pool Loop Isolation) b.

545 (Reactor Loop Isolation) c.

527D (2" bypass drain) a.

547 (Reflector Convective Loop Valve)

I e.

527E/F (Reactor Loop Bypass Cleanup)

ANSWER (C.13) a.

Open b.

Open c.

Shut d.

Shut e.

Open

REFERENCE (C.13)

MURR Training Manual pages I-59 through I-65.

l Facility Comment l

The facility comment is to [ confirm] discussion with [the] examiner that the i

valve description for "b." should read " Pressurizer vent." Also that the correct responses for question C.13 should read as follows-t I

a.

Open b.

Shut c.

Shut i

d.

Open e,

Open NRC Resolution Comment accepted. Answers "b" and "d" have been changed to reflect the correct answers as supplied by the facility. Also the valve description of j

choice "b." has been changed to " Pressurizer Vent."

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MASTER U. S. NUCLEAR REGULATORY COMMISSION NON-POWER REACTOR LICENSE EXAMINATION FACILITY:

Univ. of Missouri -- Columbia REACTOR TYPE:

MURR 4

DATE ADMINISTERED:

1993/07/27 REGION:

3 CANDIDATE:

INSTRUCTIONS TO CANDIDATE:

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

Points for each question are indicated in paren-theses for each question.

A 70% 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 t

VALUE TOTAL SCORE VALUE CATEGORY 20.00 34.48 A.

REACTOR THEORY, THERMODYNAMICS AND FACILITY OPERATING CHARACTERISTICS 20.00 34.48 B.

NORMAL AND EMERGENCY OPERATING PROCEDURES AND RADIOLOGICAL CONTROLS 18.00 30.04 C.

PLANT AND RADIATION MONITORING SYSTEMS 58.00__

TOTALS t

FINAL CRADE All work done on this examination is my own.

I have neither given nor received aid.

Candidate's Signature L

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

1. Cheating on the examination means an automatic denial of your application i

and could result in more severe penalties.

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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 not received or given assistance in completing the examination. This must be done after you complete the examination.

3. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.

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

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

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6. Fill in the date on the cover sheet of the examination (if necessary).

7. Print your name in the upper right-hand corner of the first page of each section of your answer sheets.

B. The point value for each question is indicated in parentheses after the question.

9. Partial credit will NOT be given.

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

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II. When you are done and have turned in your examination, leave the examin-ation area as defined by the examiner.

If you are found in this area while the examination is still in progress, your license may be denied or revoked.

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-_-e

EQUATION SHEET (n - 8)2 I

Q = m c AT P.

2a(k)f Nd p

f = 39 y sec k = 96 8 - 0.007 a = 1.26 x 10

t I

Q=mAh SCR = S/(1-Keff)

Q = UA AT CR, (1-Keff), = CR2 (1-Keff)2 26.06 (A,,,p)

(1-Keff)o (8 - p)

(1-Keff)3 SUR = 26.06/7 M = 1/(1-Keff) - CR /CR g

o P=P 10" ")

SDM = (1-Keff)/Keff 1

o P=P e"")

Pwr - W, m o

8(1-p)

P=

P, f* = 1 x 10'5 seconds 8-p (E*/p) + [(IT-p)/A,,,p]

7 = f*/(p-6) 7=

p = (Keff-1)/Keff A,,, = 0. ] seconds

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~

p = AKeff/Keff 0.693 l

I1/2

A i

2 2

DR D

- DR 0 DR = DR e*

f 33 22 o

6CiE(n)

DR =

DR s

, Ci = Curies, E a Mev, R = feet 2

R 1 Curie = 3.7x10'8 dps I kg - 2 21 lbm 3

6 I hp = 2.54x10 BTV/hr 1 Mw = 3.41x10 BTU /hr 1 BTU = 778 ft-lbf

F = 9/5*C + 32 1 gal H O = 8 lbm

'C = 5/9 (*F - 32) 2 t

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.s

Section A R Theory. Thermo & Fac. Operatino Characteristics Page 4

QUESTION (A.1)

[1.0]

You are performing a reactor startup and have just pulled out the source.

If you were to continue the startup by pulling out the regulating rod adding 50.30 worth of reactivity to the reactor, what would be the resulting reactor period?

a.

= 60 seconds b.

= 30 seconds c.

= 20 seconds d.

= 10 seconds 1

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QUESTION (A.2)

[1.0]

The reactor has been run for a short period of time at 1 Mwatt before being 1

shutdown (Equilibrium Xenon conditions). An experiment worth -2.0 x 10'3 oK/K is REMOVED from the reactor. When the reactor is restarted two days later pool temperature is 5'F WARMER. What will be the difference in the rod height when the reactor is returned to 1 Mwatt (equilibrium Xenon conditions)?

a.

0.175% oK/K less rods must be withdrawn j

i b.

0.265% oK/K less rods must be withdraw, c.

0.056% oK/K more rods must be withdrawi.

d.

0.119% oK/K more rods must be withdrawn

QUESTION (A.3)

[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 an 1/M (M = Subcritical Multiplication Factor).

c.

To provide neutrons to initiate the chain reaction.

d.

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

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Section' A FL Theory. Thermo & Fac. Operatino Characteristics Page 5 l

QUESTION (A.4)

[1.0]

What is the K,,, for a reactor. shutdown by 0.0455 AK/K7 a.

0.957 b.

0.855 c.

0.786 d.

0.0455

QUESTION (A.5)

[1.0)

At the beginning of a reactor startup K is 0.90 with a count rate of 30 cps.

Powerisincreasedtoanew, sue #ady-state of 60 cps. The new K,,,

is:

a.

0.91 b.

0.925 c.

0.95 d.

0.974

QUESTION (A.6)

[1.0)

The reactor is operating at 9 Mwatts (90%) and the scram setpoint is set at 125%.

What will be the resulting peak power if an experiment inserted into the reactor causes a 100 millisecond period and the scram delay is 0.1 second7 I

a.

14 Mwatts b.

28 Mwatts c.

34 Mwatts d.

50 Mwatts

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Section A F1 Theory. Thermo & Fac. Operatino Characteristics Page 6

QUESTION (A.7)

[1.0]

Beta (8) and Beta-effective (8,y) both describe the total fraction of delayed neutrons.

The difference between the two is:

1 a.

8,n is smaller than 8 since delayed neutrons are born at a lower energy level than prompt neutrons b.

B is larger than 8 since delayed neutrons are born at a lower energy en level than prompt neutrons c.

8,n is smaller than 8 since delayed neutrons are born at a higher energy level than prompt neutrons d.

8,n is larger than 8 since delayed neutrons are born at a higher energy level than prompt neutrons

QUESTION (A.8)

[1.0]

Which one of the following characteristics of a material would result in the most efficient thermalization of neutrons?

4 a.

LOW atomic mass number and HIGH scattering cross section.

b.

HIGH atomic mass number and LOW scattering cross section.

c.

LOW neutron absorption and LOW scattering cross section.

d.

LOW neutron absorption and HIGH atomic mass number.

QUESTION (A.9)

[1.0]

Which ONE of the following is the time period in which the maximum amount of Xe"5 will be present in the core?

I a.

7 to 11 hours1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months after a startup to 100% power b.

3 to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a power increase from '3% to 100%.

t c.

3 to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a power decrease from 100% to 50%.

d.

7 to 11 hours1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months after a scram from 100%

i Section A R Theory. Thermo & Fac. Operatina Characteristics Page 7 i

QUESTION (A.10)

[1.0]

Each fission event releases approximately 200 Mev of energy. Which one of the i

following contributors to this energy is the largest?

i a.

Beta and gamma radiation (B-y) b.

Prompt and delayed neutrons c.

Kinetic energy of the fission fragments d.

Alpha radiation (a)

QUESTION (4 11)

[1.0)

Which ONE of the following statements describes the Nuclear Instrumentation response for a rod withdrawal while the reactor is subtritical? (Assume the reactor remains subcritical) a.

Count rate will rapidly increase (prompt jump), then gradually increase to a new stable value.

b.

Count rate will rapidly increase (prompt jump), then gradually decrease to the initial value.

c.

Count rate will rapidly increase (prompt jump) to a new stable value.

d.

Count rate will not change until criticality is reached.

QUESTION (A.12)

[1.0]

Integral Rod Worth is defined as the reactivity:

a.

change per unit of rod motion.

b.

due to control rod position.

c.

required for shutdown with the most reactive rod withdrawn.

d.

available for shutdown after control rod withdrawal.

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Sortion A R Theory. Thermo & Fac. Operatino Characteristics Page 8

QUESTION (A.13)

[1.0]

The fuel temperature coefficient of reactivity is -1 x 10 AK/K/*F. When a control rod with an average rod worth of 0.1% oK/K/ inch is withdrawn to 10

-inches, reactor power increases and becomes stable at a higher level.

Assuming the moderator temperature is constant, the fuel temperature has:

l a.

increased by about 100*F b.

decreased by about 100*F c.

increased by about 10*F d.

decreased by about 10*F j

QUESTION (A.14)

[1.0]

Which factor in the six factor formula is represented by the ratio:

number of neutrons that reach thermal enerav f

number of neutrons that start to slow down a.

fast non-leakage probability (L,)

b.

resonance escape probability (p) c.

reproduction factor (g) l d.

thermal utilization factor (f) l l

QUESTION (A.15)

[1.0)

During a fuel loading, as the reactor approaches criticality, the value of 1/M:

I a.

decreases toward zero b.

decreases toward one c.

increases toward infinity d.

increases toward one

?

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Section A R Theory. Thermo & Fac. Ooeratino Characteristics Page.9 l

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QUESTION (A.16)

[1.0]

Which condition below describes a critical reactor?

l a.

K - 1, AK/K - I b.

K - 1, AK/K - O c.

K - 0, AK/K - 1 d.

K = 0, AK/K - O r

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QUESTION (A.17)

[1.0]

Which change to the core will niost strongly affect the Thermal Utilization Factor?

l a.

Build up of fission products in the fuel.

t b.

Removal of moderator. (Thermal expansion) j c.

Going from High Enrichment fuel to i.cw Enrichment fuel.

d.

Removal of control rod

QUESTION (A.18)

[1.0]

i In a reactop at full power, the thermal neutron flux (0) is 2.5 x 10'2 neutrons /cm /sec., and the macroscopic fission cross-section I, is 0.1 cm.

The fission rate is:

I 2.5 x 10" fissions /cm/sec.

a.

b.

2.5 x 10'3 fissions /cm/sec.

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2.5 x 10" fissions /cm /sec.

c.

3 d.

2.5 x 10'3 fissions /cm /sec.

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t Section A R Theory. Thermo & Fac. Operatino Characteristics Page 10 i

. QUESTION (A.19)

[1.0]

{

K,,, differs from K, in that K,,, takes into account:

t a.

leakage from the core l

b.

neutrons from fast fission i

i c.

the effect of poisons l

d.

delayed neutrons I

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QUESTION (A.20)

[1.0]

'Which ONE of the following parameter changes will require a control rod i

INSERTION to maintain reactor power constant following the change?

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a.

Samarium buildup b.

Primary Coolant Temperature Decreases L

c.

Xenon buildup f

d.

U concentration decrease (Fuel Burnup) f 23s

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(*** End of Section A ***)

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i Section B Normal /Emera. Procedures & Red con Page 11

QUESTION (B.1)

[1.0]

What is the minimum amount of primary grade makeup water allowable for reactor operation?

a.

1000 gallons b.

2000 gallons c.

3000 gallons d.

4000 gallons f

QUESTION (B.2)

[1.0]

An experiment is being removed from the pool after irradiation. The Health Physicist has you stop bringing it up four feet below the surface of the pool.

The portable instrument used to monitor the experiment indicates 5 mr/hr over background radiation, at the pools surface. Assuming the " tenth thickness" for water is equal to 24 inches, and the Smr/br is ALL due to the sample, WHICH ONE of the following is the expected dose at one foot after removal of the experiment from the pool?

a.

400 mr/hr l

b.

1600 mr/hr c.

2000 mr/hr i

d.

8000 mr/hr

QUESTION (B.3)

[1.0]

An individual receives 100 mrem of Beta (B), 25 mrem of gamma (y), and 5 mrem

~

of neutron radiation. What is his/her total dose?

a.

275 mrem b.

205 mrem c.

175 mrem I

d.

130 mrem i

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section 8 NormallEmera. Procedures & Rad Con Page 12 l

QUESTION (B.4)

[1.0]

In accordance with 10 CFR 20, (no emergency exists) an individual in a restricted area may receive in excess of 1 25 rem /qtr when three conditions are met. Which ONE of the following is NOT a condition for exceeding the r

limit?

a.

Cumulative dose rate will not exceed 5 (N-18) Rem.

b.

Dose for the quaiter will not exceed 3 REM.

I c.

An updated NRC form 4 is on record.

d.

Personnel dosimeters must be read at double normal frequency.

QUESTION (B.5)

[1.0]

Which ONE of the following is the correct reason for plotting a reciprocal multiplication (1/M) curve during fuel loading?

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a.

Verification of minimum Shutdown Margin.

b.

Measurement of reactivity worth of the Regulating Rod c.

Estimation of a new value of critical mass I

d.

Calibration of in-core instrumentation.

QUESTION (B.6)

[1.0]

The NRC has four standard emergency classifications. Which one of the four given below is NOT applicable for the Univ. of Missouri - Columbia reactor facility?

r a.

Notification of Unusual Event i

b.

General Emergency c.

Site Area Emergency d.

Alert t

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Section B Normal /Emero. Procedures & Rad Con Page 13

QUESTION (B 7)

[1.0]

What is the maximum allowable dose which the facility director can authorize for a volunteer to receive to save the life of someone injured and trapped in the reactor compartment?

a.

125 Rem b.

100 Rem c.

75 Rem i

d.

50 Rem

QUESTION (B.8)

[1.0)

How long after a shutdown of the reactor should you keep the primary system operating to remove decay heat from the reactor?

a.

5 minutes b.

10 minutes c.

15 minutes

+

d.

70 minutes i

QUESTION (B.9)

[1.0]

When must you stop pulling the shim-s;fety blades in gang during a normal reactor startup?

a.

within 5 inches of the ECP position b.

within 2-) inches of the ECP position l

c.

within 2 inches of the ECP position d.

within 1-k inches of the ECP position a

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- Section B Normal /Emero. Procedures & Rad Con Page 14

QUESTION (B.10)

[1.0]

You are in the middle of a week of high power operations. The Shift supervisor informs you that reactor power must be reduced to support necessary work in room 114. How long may you operate at reduced power without shutting down the reactor?

a.

15 minutes r

b.

30 minutes c.

45 minutes e

d.

I hour

)

QUESTION (B.ll)

[1.0]

What is the maximum difference between the highest and the lowest shim blade height allowed for operations above 100 Kwatts power?

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a.

\\ inch i

b.

1 inch i

c.

1-inches d.

2 inches

QUESTION (B.12)

[1.0]

Which ONE of the following rod run-ins is NOT required for Mode III operation with the natural convection flanges and pressure vessel cover removed?

a.

Rod not in contact with magnet b.

Truck Entry Door c.

Reactor Period d.

Anti-siphon line high level t

w.

..a_-a.

a

Saction B Normal /Emera. Procedures & Rad Con Page 15

QUESTION (B.13)

[1.0]

What activity level should samples irradiated using the pneumatic tube system be limited to?

l a.

10 mci b.

25 mci c.

50 mci d.

75 mci

QUESTION (B.14)

[2.0 (0.4 each))

Match the technical specification listed in column A with the correct value listed in column b.

COLUMN A COLUMN B r

a.

Core temperature coefficient (M/*F)

1) -2 x 10'3 b.

Total Regulating blade Worth (M)

2) -6 x 10-5 c.

Maximum shim insertion rate from all blades (M/sec)

3) 3.0 x 10

d.

Core Void Coefficient (M /% void)

4) 6 x 10'3 e.

Maximum secured removable experiment (M)

5) 1 x 10'3

6) 2 x 10'2

QUESTION (B.15)

[1.0]

The sample taken of one of the liquid waste tanks exceeds the limits of 10 CFR 20. Which one of the methods below is the PREFERRED method of reducing activity to dispose of the liquid waste?

a.

Transfer the water to a distillation unit for evaporation of the liquid.

The solid distillate is to be disposed of as solid waste.

b.

Chemically treat the waste so that the radionuclides will form a precipitate. Then pump the water through filters to lower the activity.

c.

Maintain the liquid in the tank (s) until the radioactivity h'as decayed low enough to allow normal pumping.

d.

Add domestic cold water (DCW) to the tank to r6 duce the concentration low enough to allow pumping.

Section B Normal /Emera. Procedures & Rad Coq Page 16

QUESTION (B.16)

[1.0]

What level ~ of MURR management is responsible for analyzing Reactor Utilization

' Requests (RURs) to determine whether the proposed experiment represents a change to an existing experiment or represents a change to an existing experiment which has safety significance?

a.

The Health Physics Manager b.

The Reactor Manager c.

The Reactor Safety Subcommittee (RSSC) d.

The Reactor Advisory Committee (RAC)

QUESTION (B.17)

[1.0]

How often is the operability of the Pool Fill system required to be tested (in accordance with Technical Specifications)?

a.

Daily b.

Weekly c.

Monthly d.

Semiannually

QUESTION (B.18)

[1.0]

The Primary System Fuel Failure Monitor is secured due to an electrical problem. How often must the primary coolant' be sampled to continue reactor operation?

a.

every hour b.

every two hours c.

every four hours d.

every eight hours

'w,.<

+

-.wem,,

-~.am-

Section B Normal /Emera. Procedures & Rad con Page 17

QUESTION (B.19)

[1.0)

What is the minimum allowable reactor period during a normal reactor startup a h reaching 1 Mwatt?

a.

10 seconds I

b.

30 seconds c.

100 seconds i

d.

300 seconds

.i l

l

{

)

i i

i

(*** End of Section B ***)

Sect.jon_C__ Plant and Rad Monitorino Systems Page 18 f

QUESTION (C.1)

[1.0]

Where does waste water' from the primary and pool sampling station discharge

(

to?

r a.

Containment Hot Sump' j

b.

Liquid Waste Tank i

c.

Labyrinth Sump

{

d.

Drain Collection Tank I

QUESTION (C.2)

[1.0]

Which ONE of the following conditions will prevent placing the Regulating t

Blade in the "AUT0" mode?

l a.

a reactor period of 43 seconds I

b.

The power trace pointer on the Wide Range Meter recorder reading greater I

than the auto control prohibit.

c.

The regulating blade position at 45%.

I d.

The Wide Range Meter switch in the 5Kw red scale position.

t

QUESTION (C.3)

[1.0]

The scintillation used by the fission product monitor detects radiation from-fission products collected in the:

a.

Filter Column.

l b.

Cation Column.

~

c.

Anion Column.

i d.

Holdup Tank.

i k

t I

5 6

ms

Section C Plaat and Rad Monitorina Systems Page 19

QUESTION (C.4)

[2.0 (0.5 each)]

- Match the parameter ' listed in column A with the appropriate value listed in column B.

a.

Temperature Coefficient of Reactivity 1.

-7 x 10~5 (reactor) b.

Temperature Coefficient of Reactivity 2.

+1.34 x 10

(pool) i c.

Effective delayed neutron precursor 3.

+7.38 x 10'3 fraction (Ben) l d.

Reactivity worth of equilibrium Xenon 4.

-2.73 x 10-2 (10 Mwatt)

QUESTI6N (C.5)

[1.0]

Which ONE of the following electrical loads can be supplied from the emergency generator?

a.

Containment Building Personnel Elevator j

b.

Supply Ventilation Fan SF2 c.

Truck Entry Door Motor d.

Primary Pump P501A l

i

QUESTION (C.6)

[1.0]

Which ONE of the following conditions will result in a reactor scram?

a.

Source Range Monitor Channel 1 Inoperative b.

Thermal Column Door Open c.

High Off-gas Activity d.

Low Pool Flow

Section C Plant and Rad Monitorina Systems Page 20 i

'i

QUESTION (C.7)

[2.0 (0.5 each)]

Match the Nuclear Instrumentation channel in Column A with the type of i

detector listed in column B.

Answer in column B may be used once, more than once or not at all.

1

[

Column A Column B i

a.

Channel I (Source Range) 1.

Uncompensated Ion Chamber j

b.

Channels 2 & 3 (Intermediate Range) 2.

Compensated Ion Chamber c.

Channel 4 (Cide Range) 3.

Proportinnal Counter

{

d.

Channels 5 & 6 (Power Range) 4.

Fission Chamber i

I

QUESTION (C.8)

[1.0]

Which ONE of the following signals will result in a Rod Run-In?

t a.

Low Pressurizer Pressure i

b.

Short Period c.

Anti-Siphon System Pressure High d.

Thermal Column Door Open

QUESTION (C.9)

[1.0]

I Which ONE of the following components can be supplied by the Emergency Backup Air Compressor?

I a.

Pool Loop Isolation Valve 509 j

b.

Reactor Loop Isolation Valve 507A 7

c.

Containment Building Exhaust Valve 16A d.

Emergency Generator

[

i i

}

I

Section C Plant and Rad Monitorino Systems Page 21

QUESTION (C.10)

[1.0]

How is deicing of the cooling tower accomplished?

i a.

Open the basin steam supply valve.

f b.

Secure the cooling tower fan while maintaining normal CT flow.

c.

Run the cooling tower far. in reverse while maintaining normal CT flow.

d.

Auxiliary operator using heat guns.

QUESTION (C.11)

[1.0)

As part of his/her duties in the immediate actions for a Reactor Isolation, who has responsibility for ensuring all personnel are cleared from all levels of the containment building?

a.

The Reactor Director b.

The Duty Health Physics person c.

The Shift Supervisor d.

The Duty Reactor Operator

QUESTION (C.12)

[1.0) i Using the figure supplied, which valve lineup is correct for returning a rabbit from the reactor?

a.

Wind gates 1 and 4 open, 2 and 3 shut.

b.

Wind gates 1 and 4 shut, 2 and 3 open.

c.

Wind gates 1 and 2 open, 3 and 4 shut.

d.

Wind gates 1 and 2 shut, 3 and 4 open.

l r

_ ~ - - ~~

Saction C Plant and Red Monitorino Systems Page 22

QUESTION (C.13)

[2.0.(0.4 each)]

4 List the NORMAL (10 Mwatt operation) positions (0 pen, Shut) for the following valves:

a.

509 (Pool Loop Isolation) b.

545 (Pressurizer Vent) l c.

5270 (2" bypass drain) d.

547 (Reflector Convective Loop Valve) j i

e.

527E/F (Reactor Loop Bypass Cleanup) i I

)

QUESTION (C.14)

[1.0]

How is gamma radiation compensated for in the Startup Channel (Channel #1)?

j A compensating current equal and opposite to the signal due to gammas is a.

sent to the detector.

b.

The detector is positioned in toward and out away from the core to compensate for gammas, The output of the detector is put through a discriminator circuit which c.

passes only pulses caused by neutron interactions.

j d.

Lead shielding around the detector decreases the signal due to gammas low enough.such that compensation is not required.

?

f i

i i

c l

l y.

.. =

I Section C Plant and Rad Monitorino Systems Page 23 i

6 I

QUESTION (C.15)

[1.0]

What is the ratio between the amount of heat dissipated by the primary cooling system and the pool cooling system?

a.

9 Mwatt to 1 Mwatt I

b.

8 Mwatt to 2 Mwatt l

c.

7 Mwatt to 3 Mwatt t

d.

6 Mwatt to 4 Mwatt l

i i

4 t

i f

4 J

i r

r i

l i

i i

i a

e r

l i

i t

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

i 4

.~.

k

..... = -,,

Section A R Theory. Thermo & Fac. Operatina Characteristics Page 24 i

ANSWER (A.1) c l

Given 51.00 - 8====> p = 0.3(8) l l

r __

_ p,g - 0.3(p,,f) _

0.7(p,g) l p,g - p _

i A,fp 1,f 0.3(p,f)

A,f 0.3(p,f) j Assume A*" - 0.125====>

T = 0.7 + (0.125 x 0.3) = 23 i

REFERENCE (A.1)

INTRODUCTION TO NUCLEAR REACTOR OPERATIONS, Reed Robert Burn C 1982, 54.4, i

equation 4.15.

l l

ANSWER (A.2) l b

t

-2.0 x 10~3 AK/K, a - 1.3 x 10 AK/K/* F an increase o,f 5'F is equivglent to 5 x (+1.3 x 10) = +6.5 x 10

Given:

i

+2 x 10'3

+ 6. 5 x 10 = 2.65 x 10' or 0.265

REFERENCE (A.2)

INTRODUCTION TO NUCLEAR REACTOR OPERATIONS, Reed Robert Burn C 1982, 54.4, t

i equation 4.15.

i

ANSWER (A.3) d

REFERENCE (A.3) i INTRODUCIION TO NUCLEAR REACTOR OPERATIONS, Reed Robert Burn C 1982, 55.2, p.

i 5-1 l

ANSWER (A.4) l a

REFERENCE (A.4) i INTRODUCTION TO NUCLEAR REACTOR OPERATIONS, Reed Robert Burn C 1982, 93.3.4, equation 3.8, p. 3-21.

l 0.0455 K = K - 1 i

1 = [1 -(-0.0455)]K

{

K,n = 1/(1-p) = 1/[1(-0.0455)] = 0.9565 i

l l

I i

j

)

Section A R Theory. Thermo & Fac. Operatina Characteristics Page 25

ANSWER (A.5) l c

REFERENCE (A.5)

INTRODUCTION TO NUCLEAR REACTOR OPERATIONS, Reed Robert Burn o 1982, 55.5.

example 5.5(c)(a).

CR I ~E,

60 _

0.90 2

r5 30 1-K,f~,

CR 1-K 1

ef2 K,,,2 = 0.95

ANSWER (A.6) c Standard Power equation o.lsec i

P = P e'I'=(1.25Mwatts) e

1 l

o P = 33.98 Mwatts = 34 Mwatts.

l

REFERENCE (A.6)

INTRODUCTION TO NUCLEAR REACTOR OPERATIONS, Reed Robert Bu:n o 1982, 55.5, l

example 5.5(c)(a).

i k

ANSWER (A.7) b

REFERENCE (A.7)

INTRODUCTION TO NUCLEAR REACTOR OPERATIONS, Reed Robert Burn o 1982, 5

ANSWER (A.8) a

REFERENCE (A.8)

INTRODUCTION TO NUCLEAR REACTOR OPERATIONS, Reed Robert Burn o 1982, 6 i

ANSWER (A.9)

J e

d Answer changed per facility comment.

}

REFERENCE (A.9)

I!ITRODUCTION TO NUCLEAR REACTOR' OPERATIONS, Reed Robert Burn o 1982, 58.4 l

Table 8.4.

j

ANSWER (A.10) c

REFERENCE (A.10)

INTRODUCTION TO NUCLEAR REACTOR OPERATIONS, Reed Robert Burn o 1982, %3.2.1, Table 3.2.

t Section A R Theory. Thermo & Fac. Operatino Characteristics Page 26

ANSWER (A.ll) a

REFERENCE (A.11)

INTRODUCTION TO NUCLEAR REACTOR OPERATIONS, Reed Robert Burn o 1982, 5 4

ANSWER (A.12) l b

l

REFERENCE (A.12)

INTRODUCTION TO NUCLEAR REACTOR OPERATIONS, Reed Robert Burn 0 1982, 57.3.

l I

ANSWER (A.13) a i

REFERENCE (A.13) i Reactivity added by control rod = +(0.00 lok /K/ inch) (10 inches) = 0.01M/K.

i Fuel temperature change = - Reactivity added by rod + fuel temp coeff.

Fuel temp. change = (-0.016K/K) + (-l x 10"AK/K/'F = 100*F I

ANSWER (A.14) b

-j

REFERENCE (A.14) i INTRODUCTION TO NUCLEAR REACTOR OPERATIONS, Reed Robert Burn 0 1982, 53.3.1,

p. 3-16.

i 1

j

ANSWER (A.15) l a

i'

REFERENCE (A.15) 4 INTRODUCTION TO NUCLEAR REACTOR OPERATIONS, Reed Robert Burn 0 1982, 55.4, pp.

5-15 through 5-18.

l 4

ANSWER (A.16) b l

i

REFERENCE (A.16)

INTRODUCTION TO NUCLEAR REACTOR OPERATIONS, Reed Robert Burn 0 1982, 53.3.4, l

example 3.3.4(b) p. 3-23 1

i

ANSWER (A.17) d

REFERENCE (A.17)

INTRODUCTION TO NUCLEAR REACTOR OPERATIONS, Reed Robert Burn o 1982, 53.3.1 Neutron Multiplication

ANSWER (A.18) l c

REFERENCE (A.18)

INTRODUCTION TO NUCLEAR REACTOR OPERATIONS, Reed Robert Burn o 1982, 52.6.2, equation 2.15, p. 2-50.

l R = 0 I, = (2.5 x 10 ) x 0.1 = 2.5 x 10" 12 t

4

,e%..

i

.T.*~*71"'

g 7

Section A P, Theory. Thermo & Fac. Operatina Characteristics Page 27

}

l

ANSWER (A.19) a

REFERENCE (A.19)

INTRODUCTION TO NUCLEAR REACTOR OPERATIONS, Reed Robert Burn o 1982, 53.3.1

ANSWER (A.20) b

REFERENCE (A.20)

INTRODUCTION TO NUCLEAR REACTOR OPERATIONS, Reed Robert Burn o 1982, 57.7.4.

4 1

i i

1 I

l

)

i 4

j

.i l

l t

a 4

I l

r i

f

(*** End of Section A ***)

i i

3y r -

+

Section B Normal /Emero. Procedures & Rad Con Page 28 4

ANSWER (B.1) b

REFERENCE (B.1)

Technical Specification 3.10

ANSWER (B.2) d

= D R,

d DR2 d2 Where Initial Activity 2500 mr/hr (100 x 5 mr/hr) 2 Activity = 500 mr/hr (4 /1 ) = 500 x 16 Activity - 8000 mr/hr

REFERENCE (B.2)

ANSWER (B.3) d

REFERENCE (B.3) a rem is a rem is a rem

ANSWER (B.4) d

REFERENCE (B.4) 10 CFR 20.101.b(1), (2) and (3)

ANSWER (B.5)

C

REFERENCE (B.5)

ANSWER (B.6) b

REFERENCE (B.6)

MURR Emergency plan, 53.0, Classification of Emergency Conditions

ANSWER (B.7) c

REFERENCE (B.7)

Site Emergency Procedure SEP-ll, Monitoring Planned Exposures in Excess of Limits in 10 CFR 20, also Emergency Plan for the University of Missouri Research Reactor Facility G 5.0.1 Protective Actions for All Classes, page 12.

ANSWER (B.8) c

REFERENCE (B.8) 50P IV-2, Shutdown of Primary System 7

~

Section B Normal /Emero. Procedures & Rad Con Page 29

ANSWER (B.9) u c

REFERENCE (B.9)

SOP 11-1, and 50P l-1.

ANSWER (B.10)

C

REFERENCE (B.10) l SDP 11-7, Reductions in Power t

t

ANSWER (B.ll) b

REFERENCE (B.11)

Technical Specifications 53.2, Control Blade Operation.

ANSWER (B.12) d i

REFERENCE (B.12)

~

Technical Specifications, 5 3.4.c Reactor Instrumentation

ANSWER (B.13) b e

REFERENCE (B.13)

SDP VIII.3, Pneumatic Tube (P-tube) System Irradiations, 6 3.2, Sample Limitations l

ANSWER (B.14) a 2

i b

4 c

3 i

d 1

l e

4

REFERENCE (B.14)

ANSWER (B.15)

C-

REFERENCE (B.15) l S0P s Vll.B.3.B page VII-23

ANSWER (B.16) l b

REFERENCE (B.16)

S0P f Vill.1.1.B.3 page VIII-2 l

t t

e

-e.,

w e,y

- - - - - s wn. r w -,

n!

t t

Section B Normal /Emera. Procedures & Rad Con Page 30 l

ANSWER (B.17) d i

REFERENCE (B.17)

I Technical Specifications E 5.6 Auxi7iary Systems.

t

-I

ANSWER (B.18) i C

i

REFERENCE (B.18) l Technical Specifications, f 3.9, Coolant System, page 1 of 3.

t

ANSWER (B.19)

[

c REFERENCE (B.19)

SDP 51I.1.1 Procedure for Reactor Startup j

i f

i I

)

l r

l 1

i

(

)

i.

f 2i-y. _

-^

..~

Sgetion C Plant and Rad Monitorina Systems Page 31

-j

ANSWER' (C.1) d

REFERENCE (C.1)

Training Manual for Reactor Operators, Page I-75

ANSWER (C.2) c

REFERENCE (C.2)

Univ. of Missouri Research Reactor, Training for Reactor Operators, f 2.15 Rod Control System, page 79.

l

ANSWER (C.3) f c

REFERENCE (C.3) l Hazards Summary and Training Manual i

ANSWER (C.4)

{

a. I

b. 2

c. 3 i

d. 4.

l

REFERENCE (C.4)

ANSWER (C.5) c

REFERENCE (C.5)

{

Training Manual pages III-10 through III-12 i

4

ANSWER (C.6) i d

t 4

REFERENCE (C.6) 5 l

ANSWER (C.7) l a.

4 l

b.

2 c.

2 1

d.

1 i

j

REFERENCE (C.7)

Training Manual for Reactor Operators, pages 11-1 through II-11.

ANSWER (C.8) b.

T

REFERENCE (C.8)

Training Manual for Reactor Operators, page 11-65.

p e e.

Sectica C Plant and Rad Monitorina Systems Page 32 i

ANSWER (C.9)

C

REFERENCE (C.9)

Training Manual, p. IV-5 i

ANSWER (C.10) l C

l

REFERENCE (C.10) i S0P G VI.3.3 page VI-6.

ANSWER (C.II) d

REFERENCE (C.ll)

FEP-2, page 2 of 4.

ANSWER. (C.12) d

REFERENCE (C.12)

Schematic diagram of Pneumatic Tube System.

ANSWER (C.13) a.

Open t

b.

Open Shut c.

Shut d.

6 hut Open e.

Open Answers "b" and "d" changed per facility comment.

t 4

REFERENCE (C.13)

MURR Training Manual pages I-59 through I-65.

ANSWER (C.14) c l

REFERENCE (C.14) l Training for Reactor Operators, Page 11-11.

[

k i

ANSWER (C.15) j a

REFERENCE (C.15) t Training for Reactor Operators, pages I-22 and I-38, Also facility supplied _

question j

h 5

r i

i-

.m n.n.-

1 Section A Rr Theory. Thermo & Fac. Operatina Characteristics Page 1 l

l 1.

a b c d II.

a b c d i

l J

2.

a b c d 12.

a b c d i

3.

a b c d 13.

a b c d i

4.

a b c d 14.

a b c d i

5.

a b c d 15.

a b c d l

t 6.

a b c d 16.

a b c d 7.

a b c d 17.

a b c d 8

a b c d 18.

a b c d 9.

a b c d 19.

a b c d j

10.

a b c d 20.

a b c d r

i i

t

?

i i

t

.~

f i

t i

t s

^

D Section B Normal /Emero Procedures & Rad Con Page 2 j

r r

1.

a b c d 13.

a b c d

{

i 2.

a b c d

14. a.

1 2 3 4 5 6 i

3.

a b c d b.

1 2 3 4 5 6 4.

a b c d c.

I 2 3 4 5 6

}

-5.

a b c d d.

1 2 3 4 5 6 t

i 6.

a b c d e.

1 2 3 4 5 6 l

i 7.

a b c d 15.

a b c d 16.

a b c d 8.

a b c d

)

9.

a b c d 17.

a b c d I

10.

a b c d 18.

a b c d I

11.

a b c d 19.

a b c d i

i 12.

a b c d l

t d

i i

P l

1 l

1 n

1 e

2-

t i

Section C Plant and Rad Monitorina Systems Page 3 1.

a b c d 8.

a b c d i

2.

a b c d 9.

a b c d i

3.

a b c d 10.

a b c d i

4. a.

1 2 3 4

11. a b c d l

b.

1 2 3 4 12.

a b c d i

c.

I 2 3 4 13.

a Open Shut 3

d.

1 2 3 4 b Open Shut i

5.

a b c d c Open Shut l

l i

4 6.

a b c d d Open Shut

7. a.

1 2 3 4 e Open Shut f

b.

1 2 3 4 14.

a b c d l

c.

1 2 3 4 15.

a b c d j

P 9

d.

I 2 3 4 i

i 1

J i

J F

.1 l

Section A F1 Theory. Thermo & Fac. Operatino Characteristics Page 1 4

ANSWER KEY A. I c A. 2 b A. 3 d A. 4 a A. 5 a A. 6 c A. 7 b A. 8 a A. 9 e d

Answer changed per facility comment.

A.10 c A.ll b A.12 b A.13 a A.14 b A.15 a A.16 b A.17 d

(

A.18 c A.19 a l

A.20 b l

l i

Section B Normal /Emera Procedures & Rad Con Page 2 ANSWER KEY t

B. I b

B. 2 d

B. 3 d

B. 4 d

B. 5 c

B. 6 b

B. 7 c

B. 8 c

B. 9 c

4 B.10 c

B.ll b

B.12 d

8.13 b

B.14 a,2 b, 4 c, 3 d,1 e, 4 B.15 c

B.16 b

B.17 d

B.18 c

B.19 c

j l

l

)

1 i

_a

Section C Plant ar.d Rad Monitorina Systems Page 3 i

ANSWER KEY C. I d

C. 2 c

C. 3 c

a,1 b,2 c,3 d,4 C. 5 c

C. 6 d

C. 7 a, 4 b, 2 c, 2 d,1 C. 8 b

C. 9 c

C.10 c

C.11 d

C.12 d

C.13 a, Open b, Open Shut c, Shut d, Open Shut e, Shut Aaswers "b" and "d" were changed per facility comment C.14 c

C.15 a

.

ML20056G058

Text

SUMMARY

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