ML063540127

Initial Examination Report No. 50-166/OL-07-01, University of Maryland

ML063540127
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Site:	University of Maryland
Issue date:	01/05/2007
From:	Johnny Eads NRC/NRR/ADRA/DPR/PRTB
To:	Al-Sheikhly M Univ of Maryland, Univ of Maryland - College Park
Eads J, NRR/ADRA/DPR/PRTB, 415-1471
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Report No. 50-166/OL-07-01
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January 5, 2007 Dr. Mohamad Al-Sheikhly, Director Radiation Facilities 2309A Chemical and Nuclear Engineering Building University of Maryland College Park, MD 20742-2115

SUBJECT:

INITIAL EXAMINATION REPORT NO. 50-166/OL-07-01, UNIVERSITY OF MARYLAND

Dear Dr. Al-Sheikhly:

During the week of December 4, 2006, the NRC administered operator licensing examinations at your Maryland University Training Reactor. The examination was conducted in accordance with NUREG-1478, "Non-Power Reactor Operator Licensing Examiner Standards," Revision 1.

In accordance with 10 CFR 2.390 of the Commission's regulations, a copy of this letter and the enclosures will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at (the Public Electronic Reading Room) http://www.nrc.gov/NRC/ADAMS/index.html. The NRC is forwarding the individual grades to you in a separate letter which will not be released publicly. Should you have any questions concerning this examination, please contact Mr. Phillip T. Young at 301-415-4094 or internet e-mail pty@nrc.gov.

Sincerely,

/RA/

Johnny Eads, Chief Research and Test Reactors Branch B Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Docket No. 50-166

Enclosures:

1. Initial Examination Report No. 50-166/OL-07-01

2. Facility Comments with NRC Resolution

3. Examination and answer key cc w/enclosures:

Please see next page

University of Maryland Docket No. 50-166 cc:

Director, Dept. of Natural Resources Power Plant Siting Program Energy & Coastal Zone Administration Tawes State Office Building Annapolis, MD 21401 Mr. Roland Fletcher, Director Center for Radiological Health Maryland Department of Environment 201 West Preston Street 7th Floor Mail Room Baltimore, MD 21201 Mr. Vincent G. Adams Associate Director-Reactor Facility Department of Materials and Nuclear Engineering University of Maryland College Park, MD 20742

January 5, 2007 Dr. Mohamad Al-Sheikhly, Director Radiation Facilities 2309A Chemical and Nuclear Engineering Building University of Maryland College Park, MD 20742-2115

SUBJECT:

INITIAL EXAMINATION REPORT NO. 50-166/OL-07-01, UNIVERSITY OF MARYLAND

Dear Dr. Al-Sheikhly:

During the week of December 4, 2006, the NRC administered operator licensing examinations at your Maryland University Training Reactor. The examination was conducted in accordance with NUREG-1478, "Non-Power Reactor Operator Licensing Examiner Standards," Revision 1.

In accordance with 10 CFR 2.390 of the Commission's regulations, a copy of this letter and the enclosures will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at (the Public Electronic Reading Room) http://www.nrc.gov/NRC/ADAMS/index.html. The NRC is forwarding the individual grades to you in a separate letter which will not be released publicly. Should you have any questions concerning this examination, please contact Mr. Phillip T. Young at 301-415-4094 or internet e-mail pty@nrc.gov.

Sincerely,

/RA/

Johnny Eads, Chief Research and Test Reactors Branch B Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Docket No. 50-166

Enclosures:

1. Initial Examination Report No. 50-166/OL-07-01

2. Facility Comments with NRC Resolution

3. Examination and answer key cc w/enclosures:

Please see next page DISTRIBUTION w/ encls.:

PUBLIC RNRP\R&TR r/f JEads Facility File (EBarnhill) O-6 F-2 ADAMS ACCESSION #: ML063540127 TEMPLATE #:NRR-074 Package No.: ML062410173 OFFICE PRTB:CE IOLB:LA PRTB:SC NAME PYoung:tls* EBarnhill* JEads:tls*

DATE 01/03/2007 12/28/2006 01/05/2007 OFFICIAL RECORD COPY

U. S. NUCLEAR REGULATORY COMMISSION OPERATOR LICENSING INITIAL EXAMINATION REPORT REPORT NO.: 50-166/OL-07-01 FACILITY DOCKET NO.: 50-166 FACILITY LICENSE NO.: R-70 FACILITY: University of Maryland EXAMINATION DATES: 12/04 - 05/2006 EXAMINER: Phillip T. Young, Chief Examiner SUBMITTED BY: /RA/ 01/03/2007 Phillip T. Young, Chief Examiner Date

SUMMARY

During the week of December 04, 2006, NRC administered Operator Licensing examinations to three Reactor Operators (RO) applicants. The two RO candidates passed the examinations.

ENCLOSURE 1

REPORT DETAILS

1. Examiners:

Phillip T. Young, Chief Examiner

2. Results:

RO PASS/FAIL SRO PASS/FAIL TOTAL PASS/FAIL Written 2/1 N/A 2/1 Operating Tests 3/3 N/A 3/3 Overall 2/1 N/A 2/1

3. Exit Meeting:

Personnel attending:

Dr. Mohamad Al-Sheikhly, Director Radiation Facilities Vince Adams, Operations Manager, Maryland University Training Reactor Phillip T. Young, NRC There were no generic concerns raised by the examiner. The examiner thanked the facility for their support in conducting the examinations.

Re-exited on December 15, 2006, via telephone.

Personnel on line:

Dr. Mohamad Al-Sheikhly, Director Radiation Facilities Vince Adams, Operations Manager, Maryland University Training Reactor Phillip T. Young, NRC Discussed issues with one applicants written examination results and medical form.

Facility Comments with NRC Resolution Facility Comment:

Question C.017 The answer key identifies D as the correct answer and that is the case. But according to OP101, "A" is also an appropriate answer. The reference is vague as turning the trim pot on the fuel temperature in fact does not cause a scram but placing the select switch in any other position than "OPERATE" does in fact cause the reactor to scram.

NRC Resolution:

Comments accepted, both D and A will be allowed as correct answers.

Enclosure 3

UNIVERSITY OF MARYLAND WRITTEN EXAMINATION ANSWER KEY OPERATOR LICENSING EXAMINATION December 04, 2006 Enclosure 3

Section A L Theory, Thermo & Fac. Operating Characteristics Page 1 of 20 Question A.001 1.0 point (1.0)

The reactor is operating at 100 KW. The reactor operator withdraws the Regulating Rod allowing power to increase. The operator then inserts the same rod to its original position, decreasing power.

In comparison to the rod withdrawal, 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.

Answer: a.

Reference:

ENNU 320, section 9.1.1. Outline (A.11 and A.10)

Question A.002 1.0 point (2.0)

The reactor is shutdown with a keff of 0.965. Initial power on the linear recorder is 300 mW. The operator withdraws the first rod, Shim I, to its required position; thereby adding 220 cents of reactivity.

DETERMINE the new power level.

a. 400 mW

b. 525 mW

c. 750 mW

d. 900 mW Answer: b.

Reference:

ENNU 320, section 7.4; 8.2 1. delta k/k = (keff -1)/keff = (.965 -1) / .965 = -0.0363

2. delta k/k [shim I] = $

• Beta = (220 cents) (.0070) = .0154

3. delta k/k [new] = delta k/k [initial] + delta k/k [shim I] = - .0363 + .0154 = - .0209

4. keff [new] = 1 / 1 - delta k/k [new] = 1 / 1 - (- .0209) = .980

5. (1 - keff [initial]) (CR [initial] = (1 - keff [new]) (CR [new]) (1 - .965) (300 mW) = (1 - .980) (CR [new])

525 = CR [new]

Question A.003 1.0 point (2.0)

Positive reactivity has been added to the reactor. Which of the following conditions concerning delayed neutrons causes the reactor to be controllable?

a. Delayed neutrons are born at lower energy levels than prompt neutrons.

b. The slowing down time for delayed neutrons is less than that of prompt neutrons.

c. The diffusion time for delayed neutrons is less than that of prompt neutrons.

d. The fission release time for delayed neutrons is longer for delayed neutrons than for prompt neutrons.

Answer: d.

Reference:

ENNU 320, section 7.2 and figures 7-1 and 7-2.

Section A L Theory, Thermo & Fac. Operating Characteristics Page 2 of 20 Question A.004 1.0 point (4.0)

Following a scram, the value of the stable reactor period is:

a. infinity, since neutron production has been terminated.

b. approximately 50 seconds, because the rate of negative reactivity insertion rapidly approaches zero.

c. approximately -10 seconds, as determined by the rate of decay of the shortest lived delayed neutron precursors.

d. approximately -80 seconds, as determined by the rate of decay of the longest lived delayed neutron precursors.

Answer: d.

Reference:

ENNU 320, Vol.1, Sect. 9.1.1 Question A.005 1.0 point (5.0)

Following a one week shutdown, the reactor is started up to 235 kw with rod control in automatic. If the reactor remained in automatic for the next five hours what would be the change in xenon concentration be during hours two through five?

Xenon concentration would:

a. decrease due to the absorption of neutrons by xenon.

b. decrease due to the beta decay of xenon to cesium.

c. increase due to the decay of iodine to xenon.

d. increase due to the fission yield of xenon from fission.

Answer: c.

Reference:

ENNU 320, section 9.4.

Question A.006 1.0 point (6.0)

The MUTR startup neutron source produces neutrons by which of the following methods?

a. Spontaneous Fission

b. Gamma Ray Absorption

c. Photofission

d. Alpha Particle Absorption Answer: d.

Reference:

ENNU 320, section 6.3.1.

Section A L Theory, Thermo & Fac. Operating Characteristics Page 3 of 20 Question A.007 1.0 point (7.0)

Which of the following would result in a decrease in excess reactivity?

a. Burnout of xenon.

b. Replacing the shim rod with a new rod.

c. Replacing four fuel bundles with new fuel.

d. Placement of an experiment containing graphite in the beam tube.

Answer: b.

Reference:

ENNU 320, Section 7.5. Outline (A.15)

Question A.008 1.0 point (8.0)

During performance of reactor power calibration which of the following conditions would result in calculated core power being greater than actual core power?

a. Pool lights are left on during the calibration.

b. The diffuser pump is left on during the calibration.

c. The primary pump is left on during the calibration.

d. Pool water level is increased during the calibration.

Answer: a.

Reference:

SP 202, Reactor Power Calibration. Outline (A.19)

Question A.009 2.0 points, 0.5 each (10.0)

Match each term in column A with the correct definition in column B.

Column A Column B

a. Prompt Neutron 1. A neutron in equilibrium with its surroundings.

b. Fast Neutron 2. A neutron born directly from fission.

c. Thermal Neutron 3. A neutron born due to decay of a fission product.

d. Delayed Neutron 4. A neutron at an energy level greater than its surroundings.

Answer: a. = 2; b. = 4; c. = 1; d. = 3

Reference:

Burn, R., Introduction to Nuclear Reactor Operations, © 1982, § 2.5, p. 2-36.

Section A L Theory, Thermo & Fac. Operating Characteristics Page 4 of 20 Question A.010 1.0 point (11.0)

You enter the control room and note that all nuclear instrumentation show a steady neutron level, and no rods are in motion. Which ONE of the following conditions CANNOT be true?

a. The reactor is critical.

b. The reactor is subcritical.

c. The reactor is supercritical.

d. The neutron source has been removed from the core.

Answer: c.

Reference:

Standard NRC Question Question A.011 1.0 point (12.0)

Which ONE of the following describes the MAJOR contributor to the production and depletion of Xenon respectively in a STEADY-STATE OPERATING reactor?

Production Depletion

a. Radioactive decay of Iodine Radioactive Decay

b. Radioactive decay of Iodine Neutron Absorption

c. Directly from fission Radioactive Decay

d. Directly from fission Neutron Absorption Answer: b

Reference:

Burn, R., Introduction to Nuclear Reactor Operations, © 1988, §§ 8.1 8.4, pp. 8-3 8-14.

Question A.012 1.0 point (13.0)

Which factor of the Six Factor formula is most easily varied by the reactor operator?

a. Thermal Utilization Factor (f)

b. Reproduction Factor ()

c. Fast Fission Factor ()

d. Fast Non-Leakage Factor (Lf)

Answer: a

Reference:

Burn, R., Introduction to Nuclear Reactor Operations, © 1982, § 3.2, pp. 3-13 3-18.

Section A L Theory, Thermo & Fac. Operating Characteristics Page 5 of 20 Question A.013 1.0 point (14.0)

Which of the following does NOT affect the Effective Multiplication Factor (Keff)?

a. The moderator-to-fuel ratio.

b. The physical dimensions of the core.

c. The strength of installed neutron sources.

d. The current time in core life.

Answer: c

Reference:

Burn, R., Introduction to Nuclear Reactor Operations, © 1982, § 3.3.4, p. 3-21.

Question A.014 1.0 point (15.0)

You perform two initial startups a week apart. Each of the startups has the same starting conditions, (core burnup, pool and fuel temperature, and count rate are the same). The only difference between the two startups is that during the SECOND one you stop for 10 minutes to answer the phone. For the second startup compare the critical rod height and count rate to the first startup.

Rod Height Count Rate

a. Higher Same

b. Lower Same

c. Same Lower

d. Same Higher Answer: d

Reference:

Burn, R., Introduction to Nuclear Reactor Operations, © 1982, § 5.7, pp. 5-28 5-38.

Question A.015 1.0 point (16.0)

Several processes occur that may increase or decrease the available number of neutrons. SELECT from the following the six-factor formula term that describes an INCREASE in the number of neutrons during the cycle.

a. Thermal utilization factor (f).

b. Resonance escape probability (p).

c. Thermal non-leakage probability (th).

d. Reproduction factor ().

Answer: d

Reference:

Burn, R., Introduction to Nuclear Reactor Operations, © 1982, § 3.2, pp. 3-13 3-18.

Section A L Theory, Thermo & Fac. Operating Characteristics Page 6 of 20 Question A.016 1.0 point (17.0)

Which ONE of the following atoms will cause a neutron to lose the most energy in an elastic collision?

a. Uranium238

b. Carbon12

c. Hydrogen2

d. Hydrogen1 Answer: d

Reference:

Burn, R., Introduction to Nuclear Reactor Operations, © 1988, § 2.5.3 p. 2-45.

Question A.017 1.0 point (18.0)

Keff for the reactor is 0.98. If you place an experiment worth +$1.00 into the core, what will the new Keff be?

a. 0.982

b. 0.987

c. 1.013

d. 1.018 Answer: b

Reference:

SDM = (1-keff)/keff = (1-0.98)/0.98 = 0.02/0.99 = 0.02041 or 0.02041/.0075 = $2.72, or a reactivity worth () of -$2.72. Adding +$1.00 reactivity will result in a SDM of $2.72 -

$1.00 = $1.72, or .0129081 K/K Keff = 1/(1+SDM) = 1/(1 + 0.0129081) = 0.987 Question A.018 1.0 point (19.0)

About two minutes following a reactor scram, period has stabilized, and is decreasing at a CONSTANT rate. If reactor power is 10-5 % full power what will the power be in three minutes.

a. 5 x 10-6 % full power

b. 2 x 10-6 % full power

c. 10-6 % full power

d. 5 x 10-7 % full power Answer:10 c

Reference:

P = P0 e-T/ = 10-5 x e(-180sec/80sec) = 10-5 x e-2.25 = 0.1054 x 10-5 = 1.054 x 10-6

Section A L Theory, Thermo & Fac. Operating Characteristics Page 7 of 20 Question A.019 1.0 point (20.0)

Core excess reactivity changes with

a. Fuel burnup

b. Control Rod Height

c. Neutron Level

d. Reactor Power Level Answer: a

Reference:

Burn, R., Introduction to Nuclear Reactor Operations, © 1982, § 6.2 p. 6-1 6-4.

Section B Normal/Emergency Procedures and Radiological Controls Page 8 of 20 Question B.001 1.0 point (1.0)

A safety limit is established on fuel element temperature to prevent a loss of fuel element cladding integrity which could arise from:

a. excessive pressure between cladding and fuel.

b. exceeding the melting point for cladding.

c. exceeding the melting point for fuel.

d. excessive swelling of the fuel.

Answer: a.

Reference:

Technical Specification 2.1 Basis, pg. 5.

Question B.002 1.0 point (2.0)

Experiments containing materials corrosive to reactor components shall:

a. only be irradiated in experimental facilities not in the reactor.

b. not be irradiated in the reactor or experimental facilities.

c. be encapsulated prior to being placed in the reactor.

d. be double encapsulated prior to being placed in the reactor.

Answer: d.

Reference:

Technical Specification 3.5, pg. 13.

Question B.003 1.0 point (3.0)

Which of the following conditions COMPLETELY satisfies the technical specification definition of "Reactor Secured?"

a. When the reactor contains insufficient fissile material or moderator present in the reactor and adjacent experiments to attain criticality under optimum available conditions of moderation and reflection.

b. When all scrammable rods have been fully inserted and verified down and the console key has been removed from the console.

c. When no work is in progress involving core fuel, core structure, installed control rods, or control rod drives unless they are physically decoupled from the control rods and no experiments are in the reactor.

d. When sufficient control rods are inserted to assure that the reactor is subcritical by at least 1.00 dollar of reactivity, with the fuel and moderator at ambient temperature.

Answer: a.

Reference:

Technical Specification Definition 1.20, pg. 3.

Section B Normal/Emergency Procedures and Radiological Controls Page 9 of 20 Question B.004 1.0 point (4.0)

An area where there is posted a "Caution - Radioactive Materials" sign is by definition a:

a. Radioactive Material Storage Area.

b. Potentially Contaminated Area.

c. High Radiation Area.

d. Radiation Area.

Answer: b.

Reference:

ENNU 320 Manual, Nuclear Reactor Operations, section 1.2.2, pg. 1-2.

Question B.005 1.0 point (5.0)

When responding to a dry spill, the spill should be:

a. covered with dry rags and then plastic.

b. cleaned up using a filtered vacuum cleaner.

c. carefully swept up and placed in a sealed plastic bag.

d. dampened with water and confined with absorbent material.

Answer: d.

Reference:

EP 404, Release Of Radioactivity, Step 3.2, pg. 2.

Question B.006 1.0 point (6.0)

The annual limit for total effective dose equivalent per 10CFR20 is:

a. 0.5 rem

b. 1.25 rem

c. 3 rem

d. 5 rem Answer: d.

Reference:

10CFR20.1201(a)(1)

Question B.007 1.0 point (7.0)

Following evacuation of the reactor building, who must authorize reentry into the Reactor Building?

a. Emergency Director

b. Radiation Safety Office

c. Emergency Coordinator

d. Reactor Support Coordinator Answer: a.

Reference:

EP 401, Reactor Building Evacuation, Step 2.7, pg. 1. EP 406, Responsibilities and Instructions of the MUTR Emergency Organization, Section 3.0, 4.0, 5.0, 6.0, & 7.0, pgs.

1 - 4.

Section B Normal/Emergency Procedures and Radiological Controls Page 10 Question B.008 1.0 point (8.0)

A REACTOR RUN is to be conducted after the initial reactor startup of the same day.

This run can be considered part of the initial run if the reactor is being started up following:

a. a valid reactor scram for which the cause has been corrected.

b. an interruption of operation for training purposes.

c. an interruption of operation for minor maintenance.

d. a reactor scram which occurred inadvertently.

Answer: b.

Reference:

OP 101, Reactor Startup Checkout, Definition 2.1, pg. 1.

Question B.009 1.0 point (9.0)

Which ONE of the following statements define the MUTR Technical Specification term "Channel Test?"

a. The introduction of a signal into a channel for verification of the operability of the channel

b. The qualitative verification of acceptable performance by observation of channel behavior

c. The combination of sensors, electronic circuits and output devices connected to measure and display the value of a parameter

d. The adjustment of a channel such that its output corresponds with acceptable accuracy to known values of the parameter which the channel measures Answer: a.

Reference:

Technical Specifications Section 1.0, Page 1 Question B.010 2.0 points, 0.5 each (11.0)

Match the type of radiation in column A with its associated Quality Factor (10CFR20) from column B.

Column A Column B

a. alpha 1

b. beta 2

c. gamma 5

d. neutron (unknown energy) 10 20 Answer: a. = 20; b. = 1; c. = 1; d. = 10

Reference:

10CFR20.100x

Section B Normal/Emergency Procedures and Radiological Controls Page 11 Question B.011 2.0 points, 0.5 each (13.0)

Match the radiation reading from column A with its corresponding radiation area classification (per 10 CFR 20) listed in column B.

COLUMN A COLUMN B

a. 10 mRem/hr 1. Unrestricted Area

b. 150 mRem/hr 2. Radiation Area

c. 10 Rem/hr 3. High Radiation Area

d. 550 Rem/hr 4. Very High Radiation Area Answer: a. = 2; b. = 3; c. = 3; d. = 4

Reference:

10 CFR 20.1003, Definitions Question B.012 1.0 point (14.0)

Per Technical Specifications, for a period of time not to exceed _______, the Exhaust Radiation Monitor may be taken out of service for maintenance if it is replaced with a portable gamma sensitive instrument observable by the reactor operator.

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

b. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

c. 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />

d. 7 days Answer: c.

Reference:

Technical Specifications 3.6 Question B.013 1.0 point (15.0)

Which ONE of the following is the 10 CFR 20 definition of TOTAL EFFECTIVE DOSE EQUIVALENT (TEDE)?

a. The sum of the deep does equivalent and the committed effective dose equivalent.

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

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

d. The dose to a specific organ or tissue resulting from an intake of radioactive material.

Answer: a.

Reference:

10 CFR 20.1003 Definititions

Section B Normal/Emergency Procedures and Radiological Controls Page 12 Question B.014 1.0 point (16.0)

Two inches of shielding reduce the gamma exposure in a beam of radiation from 400 mR/hr to 200 mR/hr. If you add an additional four inches of shielding what will be the new radiation level? (Assume all reading are the same distance from the source.)

a. 25 mR/hr

b. 50 mR/hr

c. 75 mr/hr

d. 100 mr/hr Answer: b.

Reference:

Nuclear Power Plant Health Physics and Radiation Protection, Research Reactor Version©1988, § 9.2.3 "Half-Thickness and Tenth-Thickness" Question B.015 1.0 point (17.0)

Your Reactor Operator license expires after _____ years.

a. 2

b. 4

c. 6

d. 8 Answer: c.

Reference:

10CFR55.55(a)

Question B.016 1.0 point (18.0)

Which one of the following is the basis for maintaining the reactivity coefficients for the reactor within Technical Specification Safety Limits?

a. To assure that the reactor can be shutdown from any operating condition.

b. To ensure that the net reactivity feedback is negative.

c. To ensure that the operable core is similar to the analyzed core in the FSAR.

d. To provide protection against the reactor operating outside the safety limits.

Answer: b.

Reference:

Tech. Specs 3.2

Section B Normal/Emergency Procedures and Radiological Controls Page 13 Question B.017 1.0 point (19.0)

In accordance with Technical Specifications, which One of the following statements is TRUE?

a. Explosive experiments shall be doubly encapsulated

b. The reactivity worth of any individual in-core experiment shall not exceed $3.00.

c. Experiments containing materials corrosive to Rx components shall not be irradiated in the Rx.

d. Each fuel experiment shall be controlled such that the total inventory of Iodine isotopes 131 thru 135 in the experiment is no greater than 1.5 curies.

Answer: a.

Reference:

Tech Specs 3.5 Question B.018 1.0 point (20.0)

Which one of the following is a Reportable Occurrence per Technical Specifications?

a. The safety-system setting (LSSS) for reactor fuel temperature is set at 365EC.

b. An unexpected reactivity change of $0.75

c. One of the Bridge Radiation Monitor is inoperable while the reactor is a power.

d. The interlock for the Beam Port is disabled by the Senior Reactor Operator while the reactor is at power.

Answer: a.

Reference:

Technical Specifications 1.27, Definitions Technical Specifications 3.0

Section C Facility and Radiation Monitoring Systems Page 14 of 20 Question C.001 1.0 Point (1.0)

A break has occurred in the reactor coolant system on the pool inlet line to the pool. The break is outside of the pool. Select the level that the pool will drain to and why?

a. Approximately to the height of the elbow located on the pool inlet pipe.

b. Two feet above the top of the core because this is the discharge point of the pool inlet line.

c. The top of the core because all reactor coolant piping external to the core is above the top of the core.

d. Two feet below the top of the reactor pool due to the siphon break located on the pool water outlet pipe.

Answer: a.

Reference:

FSAR section 4.1.

Question C.002 1.0 Point (2.0)

Select the component where differential pressure is measured in order to determine flow through the primary water system.

a. Primary Coolant Pump.

b. 200 Kw Heat Exchanger.

c. Restricting orifice.

d. 40 Kw Heat Exchanger.

Answer: c.

Reference:

FSAR section 4.1.

Question C.003 1.0 Point (3.0)

Select the setpoint for the high radiation scram that is initiated by the bridge radiation monitor.

a. 10 mrem/hr

b. 20 mrem/hr

c. 50 mrem/hr

d. 70 mrem/hr Answer: c.

Reference:

FSAR Table 6-1.

Section C Facility and Radiation Monitoring Systems Page 15 of 20 Question C.004 1.0 Point (4.0)

Which one of the following channels of power monitoring is supplied by a compensated ion chamber?

a. Multirange Linear Channel

b. Wide Range Log Power Channel

c. Safety Channel I

d. Safety Channel II Answer: a.

Reference:

FSAR section 6.1.2.

Question C.005 1.0 Point (5.0)

The driving force for the pneumatic transfer system is pressurized carbon dioxide. Carbon dioxide is used to minimize the production of:

a. moisture.

b. hydrogen.

c. argon-41.

d. nitrogen-16.

Answer: c.

Reference:

FSAR section 5.4.

Question C.006 1.0 Point (6.0)

As part of an approved experiment the reactor is operated without the plexiglass cover on the Through Tube. Which of the following would increase?

a. Argon 41.

b. Nitrogen 16.

c. Beta Radiation.

d. Alpha Radiation.

Answer: a.

Reference:

ENNU 320, Volume II, section 8.3.

Section C Facility and Radiation Monitoring Systems Page 16 of 20 Question C.007 1.0 point (7.0)

The regulating rod is operating in automatic. Select the response if the operator attempts to withdraw Shim I while the Regulating Rod is being withdrawn by the servoamplifier.

a. The Shim I rod will withdraw but the Regulating Rod will be interlocked to prevent further withdrawal.

b. The Regulating Rod will withdraw but Shim I will be interlocked to prevent movement.

c. The Shim I rod will withdraw but the Regulating Rod will be shifted to manual mode.

d. Both rods will withdraw simultaneously.

Answer: d.

Reference:

FSAR Section 6.1.3.

Question C.008 1.0 point (8.0)

The ion chamber power indications are correlated to the heat balance calculated thermal power by:

a. adjusting the detector high voltage.

b. adjusting the circuit comparator voltage.

c. moving the graphite reflectors to change the neutron flux near the detectors.

d. physically adjusting the height of the detectors in the support assembly.

Answer: d.

Reference:

ENNU 320, Vol. 2, Sect. 7.3; SP-202, Step 6.2 Question C.009 1.0 point (9.0)

In the automatic power regulation mode, reactor response to a large increase in demand is limited by:

a. the speed of the rod drive motor.

b. the maximum voltage on the demand potentiometer.

c. a period limiter in the input to the rate error pre-amplifier.

d. the decay constant of the longest lived delayed neutron precursor group.

Answer: c.

Reference:

FSAR section 6.3.2.

Section C Facility and Radiation Monitoring Systems Page 17 of 20 Question C.010 1.0 point (10.0)

What is achieved by use of the diffuser above the core during operation?

a. Better distribution of heat throughout the pool.

b. Better heat transfer across all fuel elements in the core.

c. Reduced dose rate at the pool surface.

d. Consistent water chemistry in the core.

Answer: c .

Reference:

FSAR section 8.3 Question C.011 1.0 point (11.0)

When is the replaceable demineralizer cartridge in the primary system required to be replaced?

a. An annunciator light on the console is illuminated during primary system operation.

b. The differential pressure across the orifice indicates a flow rate of less than 10 gpm.

c. The differential pressure across the filter exceeds 5 psid.

d. Conductivity on the outlet exceeds 1 x 10 -6 mhos/cm.

Answer: a.

Reference:

FSAR section 4.1.

Question C.012 1.0 point (12.0)

A control rod interlock is applied when the trip test switch for Safety Channel I is placed on.

This control rod interlock is required because the:

a. period scram signal is bypassed.

b. minimum source count rate interlock can be bypassed.

c. gain of the instrument is affected by operation of the trip test knob.

d. the input signal to the automatic-servo is affected by the trip test knob.

Answer: b.

Reference:

ENNU 320, Volume 2, section 6.1.2.1.

Section C Facility and Radiation Monitoring Systems Page 18 of 20 Question C.013 1.0 point (13.0)

What feature of the area radiation monitors allows the operator to check that the channels are functional when the reactor has been shutdown for an extended period and the background radiation levels are very low?

a. A low level source is installed in the detector to keep the instrument on scale.

b. The self check circuitry in the instrument channel illuminates the yellow light if readings are below the range of the indicator.

c. The self check circuit maintains an artificial input signal at a level just above the instrument minimum sensitivity so that it is never below scale.

d. A portable Co-60 source is provided for positioning near the detector and verifying, or adjusting, the channel linearization to within 10% of known radiation levels.

Answer: a.

Reference:

ENNU 320, Vol. 2, Sect. 6.3; SP-205, Sect. 5.0 Question C.014 1.0 point (14.0)

In addition to the control room, where can the exhaust fan radiation level be read?

a. Left hand side of the entrance to the west balcony laboratories from reactor bridge.

b. Opposite the door into the reception room, from the outside hallway.

c. Opposite the door into the west balcony, from the outside hallway.

d. In the Nuclear Engineering Program office.

Answer: a.

Reference:

ENNU 320, Volume 2, section 6.3 Question C.015 1.0 point (15.0)

The instrumented fuel rod will measure core temperature that is:

a. equal to the average of all fuel rod temperatures.

b. the highest fuel rod temperature during normal conditions.

c. the highest fuel rod temperature during accident conditions.

d. at least 50% of the temperature of the hottest fuel rod during accident conditions.

Answer: d.

Reference:

ENNU 320, Volume 2, section 6.1.3.2.

Section C Facility and Radiation Monitoring Systems Page 19 of 20 Question C.016 1.0 point (16.0)

If an operator were to continuously withdraw a shim rod from the core approximately how long would it take for the rod to go from the bottom to the top of the core?

a. 30 seconds.

b. 45 seconds.

c. 75 seconds.

d. 120 seconds.

Answer: b.

Reference:

ENNU 320, Volume II, Appendix A.

Question C.017 1.0 point (17.0)

Operation of which of the following Calibrate switches will result in a scram?

a. Fuel temperature

b. Wide range log power channel

c. Multirange linear channel

d. Safety Channel I Answer: d. and a. per facility comment

Reference:

ENNU 320, Volume II, section 6.1.4.

Question C.018 1.0 point (18.0)

Select the location where the ventilation system can be secured.

a. Entrance to the west balcony from the hallway.

b. Inside the water room.

c. Pneumatic transfer system laboratory.

d. West wall of the ground level.

Answer: d.

Reference:

ENNU 320, Volume II, Figure 2.5.

Section C Facility and Radiation Monitoring Systems Page 20 of 20 Question C.019 1.0 point (19.0)

During pre-startup checkout, the reactor operator lines up the makeup water system to add water to the pool and neglects to check the level later. Overflow from the pool will go:

a. directly into the water handling room sump through the pool overflow piping.

b. into the holdup tank in the water handling room sump through the pool overflow piping.

c. directly into the sewer system through the floor drain gratings around the base of the reactor.

d. directly into the water handling room sump through the floor drain gratings around the base of the reactor.

Answer: a.

Reference:

ENNU 320, Vol.2, Sect. 8.1 Question C.020 1.0 point (20.0)

The thermocouples in the instrumented fuel bundle measure temperature at the:

a. interior surface of the cladding

b. center of the zirconium rod

c. outer surface of the fuel

d. interior of the fuel Answer: d.

Reference:

ENNU 320 MANUAL VOL. 2, Page 3-1