ML19323H023

MD Univ Training Reactor Requalification Program.

ML19323H023
Person / Time
Site:	University of Maryland
Issue date:	06/30/1980
From:	MARYLAND, UNIV. OF, COLLEGE PARK, MD
To:
References
NUDOCS 8006110195
Download: ML19323H023 (13)

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7 COLLEGE PARK MARYLAND,20742 L,

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REQUALIFICATION PROGRAM

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FOR

( LICENSED OPERATORS

{ Pursuant to 10CFR, Part 55 Section 55.33 which requires that each licensed individual demonstrate his

[ continued competence every two years in order for his license to be renewed, the following outline of the

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University of Maryland Operator requalification program

{ is submitted.

1. Schedule: "The requalification program shall be

[ conducted for a continuous period not to exceed two years."

[

Licensees at the University of thryland and generally

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staff members of students participating in the nuclear program. Thus, during the period of their license they

( are either teaching or taking courses related to reactor theory or operation. Such duties require frequent

[ operation of the training and research reactor.

2. Lecture: "The requalification program shall include preplanned lectures on a regular and continuing basis

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throughout the license period."

{ During the period of his licenses at the MUTR the l licensee is either teaching or taking the following

[ courses in our nuclear program. They are:

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ENNU 215: Introduction to Nuclear Technology

( ENNU 320: Nuclear Reactor Operations ENNU 440: Nuclear Technology Laboratory

[ ENNU 450, 455: Nuclear Reactor Engineering I & II ENNU 460: Nuclear Heat Transport These courses are taught at least once per year.

( Operator may be required to participate in the laboratory sections of ENNU 320 which involves teaching the practice of Nuclear Reactor Operations.

3. On-the-job training: "The requalification program shall include on the job training."

( Each license is required to perform at least ten start-ups and shut-downs during each calendar year. At least one of these is under the observation / direction of the Reactor Director. In addition each licensee parti-t cipates in the preventive maintenance program involving such areas as,

(

a) Control rod inspection b) Fuel inspection c) Rod drop time measurement.

(

Licensees are also encouraged to attend the quarterly

( meeting of the Reactor Safety Committee.

4. Evaluation: The annual enduction of the licensee is based upon a written examination, and discussion and actual start-up and reactor operations.

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[ The written examination covers the following areas:

a) Principles of Reactor Operations

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b) Features of Facility Design

( c) General Operating Characteristics d) Instruments and Controls e) Safety and Emergency Systems f) Standard and Emergency Operating Procedures g) Radiation Control and Safety h) Administrative Procedures, Controls and Limitations.

A typical examination is shown in Appendix A.

5. Records: Annual examinati.on records and expired operators' licenses are kept en file for a period of at least two years. Current operators' licenses are

{ prominently displayed in the reactor console room.

In accordance with 10CFR, Part 55.60 Section 6 of Appendix A, the Reactor Director prepares the annual examination, reviews it with the licensee and maintains records of these examination. He certifies that each

[ licensee has a minimum of ten operations wherein the controls are manipulated. In accordance with past policy, the Reactor Director is exempted from taking the written requalification examination.

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MARYLAND UNIVERSITY TRAINING REACTOR OPERATOR REOUALIFICATION EXAMINATION

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A. Principles of Reactor Operation y .l. Discuss and explain the role of the moderator and reflector in

[ a reactor. -

l P . 2. What is' meant by stable reactor period? Explain.

[

} 3. '

A source of neutrons from an external source is usually used in connection with start-up of a reactor. E); plain why this

( source is used. Is it always needed?

M 4. What is the source of delayed. neutron, and what are their role ,

( in reactors?

5. How does the critical control rod position. vary with power level?

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Explain your answer.

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6. Define the following: '

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( (a) Multiplication f actor, -(b) . Reactivity, .

(d) % &./k , (c) $, (f) cent. I

( 't 7. What is the value of excess reactivity for MUTR (2 50hW) Whc.t " . 't does it mean? How can.it be determined? Explain your answer.

(, 8. Xe 135 and Sm 149 are very strong neutron.absorbersy therefore, the accumulation of these fission products in the fuel region of t

j a reactor will have a large (positive), (negative) reactivity -

(

offect. The strongest poison of the two fission products Xe 135 and Sm 149 is ( ).

9. Explain briefly the follouing:

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( (c) Criticality.

(d ) Neutron flux.

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(e ) Nuclear cross section. '

• (f ) Promot critical.

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10.

j As operator, you are bringing the reactor to power on a stable i period of thirty seconds. At a certain time, the nuclear instrumentation indicates a level of ten watts. What level will iili be indicated at the following times: ti t

(a) two minutes- t

'r (b) Three minutes-e -

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[ D. Features of Facility Design

2. Name and draw positions of each neutron detector in the core.

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3. What are the control rods . (MUTR) made of?

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• 4. How many exhaust fans are.there at the MUTR facility? ,

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. hey are located at .

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' .5 . Lh.stthefollowinginformation.

(a) Active core dimensions. ,

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(b) Number of fuel elements.

(c) Fuel enrichment.

. . .i For the following console readouts indicate:

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(a) The name of the detectar supplying the signal.

(b) The trip setting (s) associated with the Channel.

(1) Log Power Meter (2) Linear Chart Recorder (3) Safety Channel 1 .

(4) Safety Channel 2

[- (5) Period Meter l

- (c) For each detector noted above, indicate.how it distingui hes )

[- between neutron and gamma induced pulses.

7. Where in the core is the instrumented fuel element located?

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Can it be relocated without changing the console setting for the upper limit temperature? If so, Why?

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K 8. Li'st the materials composing the thermal column, starting at the face of the reactor shielding and proceeding toward the

[ . reactor core.

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General Operating Characteristics OPERATION AND PROCEDURE -

1 Why is an. instrument check-out required before start-up? Explain.

If the building monitor alarm .sarest the pool goes off scale

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2 and sounds a high radiation leve alarm while the reactor,is operating steadily, what would yoa 's the reactor operator do?

r- Consider possible causes of this event. How would one establish L . what the real cause is,. and what action would be taken?

3. Should you ever operate the reactor when one.of the ion chambers

[ has failed? Explain. .

.4 What would b'e tile consequences if all of the pool water should

[ be lost so that the core is uncovered?

5 You are the duty operator, and. the reactor is operating at full

[ power when an experimenter asks if he.may lower a small sample L- dovm against the core.. What procedure would you follow in assisting him? \

[ 6. Give values for.each: .

(a) Average thermal flux (b) Void coefficient

[ - (c) Temperature coefficient .

. (d) , critical mass

b. 7 ?3 What would be the effect of xenon buildup after three days of

. operation at full power?

y ~8 Give method for determining.when the reactor is critical.

9. ' What actions should be taken if, . during the MUTR start-Up, you

[ . believe that-(a) The detectors do not indicate.the true situation in the IcaCtor.

[' (b) Source drive motor is rot working properly.

(c) Source "in" and "out" indications not working.

(d) In caditinn to all above, you couldn't locate and get in

[. touch with the senior operator.

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.D. Instruments and Controls I 7 1 Why is there an interlock which prevents you from liiting the I safety rods if you have less than 2 CPS on the Log count rate motor?

I 2.

Describe b. afly each ci ennel in the control system, and explain its primary purpose.

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3.

Draw recorder power. traces on changing power levels and going up to I 4.

Draw the two safety channels, and.. discuss each circuit. '

.5. Give values for each:

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(a) Regulating rod speed. ~

1 ' (b) Control (Shim) rod speed.

l (c) Rod drop time. ~ '

(d) Minimum power level for automatic. operation.

6.

How often is the instrument. check done? How often the startup?

T l 7.

What would happen if CIC were overcompensated, would the counts increase or decrease?

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.y 8. W 1 ,is hatitislocated?

the purpose of the siphon break, and on which pipe I- }. 9 . been List threedriven indicators which will determine if'the sonrce has into the core.

-E. Safety and Emergency System '

1.

List the six scram indicators and give the appropriate trips for each; what three conditions.will give: a MANUAL; an EXTERNAL scram?

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3. Inklication for gross release of radioactivity.

Why circuitsis the pressure in the different in the primary and secondary heat exchanger?

4.

How would a leaky fuel element in the core be detected?

Explain in detail.

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Why is the difusser necessary for operation at 250 kW?

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is the principle governing its use? What l t A 6.. i 1

What useful information is obtained from the " Bulk Water Meter?"

7.

1 Which of the three detectors, fission chamber, CIC, and IC i 8.

has its individual power supply? Why?

U Will the build-up of Argon-41 in the beam tube and through

tubes be a problem when operating at 250 kW. If so, .

I what method has been proposed to alleviate such a problem? i F. Standard and Emergency Operating Procedures r 1. t Explain building.the Duties procedure for emergency evacuation of the reactor L

2 of the reactor operator.

State the rules and regulations for:

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H 7 (a) Disposal of liquid waste. l l (b) Removing samples from the Egm51n .- - - - - - - - - - - -

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' 3. What would happen if all water suddenly were lost?

Reactivity would be what? How about heating? ,

- 4 While you are operating the MUTR at 10 kW automatic, you have

- noticed that the power is steadily increasing. What do you do?

Explain.

5. While you are operating the MUTR.at 0.5 watts. What happens '

if the Shim I drops into the core. Specify the indications.

6. How do you insure that the reactor is scrammed when electric power failed (for about 0.5 hr) during the reactor operation.

[ Explain. ,

'7. " Describe the procedural requirements which must be met before you, as an operator, would insert a sample into or near the

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reactor for irradiation.

[ G. Radiation Control and Safety ,

1. What i the acceptable emergency maximum permissible whole body

[ radiation dose for once in a. lifetime exposure? Under what circumstances would you take this amount or instruct another to do so? Explain.

.2. When you pull a sample out of the pool, you have available a

.. portable geiger counter and an ionization chamber type survey

, meter. Which would you use and why? Explain.

3. Explain what each sign means and what level or dose cach implies:

(a) Caution radiation area.

(b) caution high radiation area. -

(c) C0ution radioactive materials.

4.~ (a)- When <:an the console be left unattended?

(b) When can the console operator leave the console? -

{, (c) When can an unlicensed operator handle the controls?

5. Discuss the purpose and use of the two counts per second typass.

[ .6. What is N16 and A41 Are these big , problems for tlic MUTR? Explain.

1 7. Give the values of RBE for. Alpha, Beta, Gamma, Fast Neutrons, and I Thermal neutrcnc. -

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y 8.

A 1.7 millicurie Co60 scurce, and a 1 curie Pu-Be source, are both E housed in a Rqd Waste Container. The gamma dose rate at 20 cm was L measured and found to be 80 milli-Roentgens /hr. The thermal neutron dose rate at 40 cm was found to be 40 millirads/hr, and the fast c neutron dose rate at the same point, was found to be 4 millirads/hr.

L (a) What will be the totcl dose rate in millirens/hr at a point

- 100 cm from the container.

1 (b) How long could one spend at that point without" exceeding the AEC hourly dose limite -

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9. (c) Write down the reaction resulting .in neutron production

- for the Pu-Be source.

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H. Reactor Theory ,

{ X1 What is meant by a reactor period of 10 sec.? ^

2 What do you mean by " critical", and what criterion may we use to determine that this condition exists in the MUTR facility?'

. 3.

Assume a small step insertion of reactivity. What will be the reactor response after the initial transient.

t y 4.

Among the products of fission, there exist particles which are.

larger, also gamma and some rays and which heat.areIn more highly ionizing, than neutrons,

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• view of this, why do we select to measure the neutron population? .

5.

{ Of what use is a control-rod calibration (integral and differential) curves? .

6'.

( What is the difference between an integral and a differential rod worth curves. How do we obtain this at the MUTR?

X7

( If we'have total of 10 fast fission neutrons ( = nx ) available for slowing down, and one.is absorbed while slowing down, so that nine (9),, escape (10),' absorption, (1.11), . (0. 9) .

the resonance escape' probability, p, is

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Radioactive Materials, Handling, Disposal and Hazards

( . 1.

Tell operator's radioactivity in role air or in water.

case of a reactor malfunction, release of 2

[ What level?design features do we have to hold Ar 41 down to permissable ' ~

/0 3.

[ If you stay in a " Radiation Area" g for.120 minutes, what will be the madam radiation dose you enbeM receive -

d. What is the air stmpling procedure at the MUTR? -

5. If 2 curies of I(131) to 10CFR20, Appendix B,is mixed with 20 liters of water, according l

{ released to the regular outlet?after how much time can this mixture be 6.

r A man is exposed to 10 mrem /hr. o'f gar.tma, 5 mr/hr of beta and 1,400 thermal neutrons /cm2 -soc.

L an hour in rem. What dosage does he pick up in half picks up maximum weeklyAnd how long can he stay in this field before he exposure?

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7. . What is the approximate attenuation to 1 Mev gamma rays afforded by:

[ (a) one inch of lead shielding.

(b) .one foot of water shielding. .,

. y of lead = 2.3/ inch

( p-of water = 0.19/ inch *

8. An operator in his haste to run from a suddenly exposed gamma source, trips and strikes his head. Assuming that he lay

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unconccious three feet from the source, calculate the following:

(a) Average whole body d'se o received in one hour.

[. (b) If the man were not removed for eight hours, discuss the probable state of his health after the exposure.

( NOTE: Source strength is 20 curies, emitting gamma energies of 1.7 and 1.4 MEV.

9.

.{ Assume there is a release of radioactive material within the

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reactor building and it is necessary that your operators enter the building area without waiting for a health physicist to arrive.

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(a) What kind of monitoring equipment would you require?

Xb) What~ precautions would you take prior to entry?

{ (c) Explain in detail how you would determine how long the operators could stay in the vessel.

fJ. Specific Operating Characteristics 1.

If, when you are operating the reactor under regular conditions,,

( someone reports that a beam tuba is open, and a high intensity beam is escaping, what action would you take?. Explain.

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2. What would operator

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. compensating voltage? do upon loss o5 h igh voltage, low voltage, and

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3. The reactor is critical at 50 watts of power and increasing en 'a stable reactor period. The. start-aup channels show a 100 second

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period and the intermediate channels show a 65 second period.

Which period do you consider is the more accurate? Explain your

, answer.' -

4 Outline a method that may be' used to determine the cold minimum core shutdown margin during low reactor power operation.

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K. Fuel Handlino and Core Parameters ,

i 1.

{ What effect does the thermal column have on the reactor? EAplain.

2. What precautions are taken so that there is no unauthorized handling of fuel? j

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3 Neutrons produced near the outer edge of the core could travel away from the core and have no opportunity to produce fission.

The reflector is placed around the core to scatter many of ,

( these neutrons back into the core where they can produce (steam), (hea t) , (fission).

{- 4 Let us assume that a reactor has a temperature coefficient of

• -0.002% Jk /0F and the operating temperature is increased from .

k 0

100 F to 1400F. Which is equivalent to a eT= The total

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reactivity (loss), (gain) will be equal to .

[ '5 It is genera'lly true thaIt the "end positions" of control-rod travel are not as effective as the." middle. positions." Why?

6.. Define the " reflector savings"

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[ 7. A start-up detector is placed near a reactor core to monitor .

a new core loading. As the loading proceeds, the following data are-obtained.

Step y'No. of. Elements -. . CPS

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4

. 10 12 2 "8 '15 3 10 20

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12 30

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13 60 .'

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6~ 14. 150 (a)

Estimate the number of fuel elements that will make the reactor critical and show how.you arrived at this figure.

[ (b) Do you think the detector is optimumly placed, too close or too far away from the source? Explain.

[.

L. Administrative Procedures, Conditions and Limitations .

(, 1 What is the responsibility of. licensed operator and senior operator?

2, Outline the steps you would normally follow to irradiate a

{ sample for a few minutes.

safety answers. ) '(Include administrative and technical e

3.

[ List the kinds of records that must be kept according to the pro-

. visions of the facility license and the regulations.

4.

[ Assume a core loading that you are the senior operator on duty and are supervising change.

into the core (it is halfway An operator is lowering a fuel element inserted) when the reactor top fixed radiation monitor alarms. What should be your actions? ,

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. 5. On your answer sheet, complete the following table by listing I the personnel that must_be in each status for the event given:

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In ,

Event -

Control At Facility on Call I

• Replacement of fuel in core Reactor re-start after scram from high flux.

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Reactor power escalation after a 10% power drop,* -

( cause unknown. Conditions

. appear norma 2.

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Normal reactor;shutdoun .

6 List the responsibilities of the Reactor _ Safety Committee.

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7 Who can legally operate the MUTR? .

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