Text

Initial Examination Report No. 50-274/OL-08-01, U.S. Geological Survey
	ML080810447
Person / Time
Site:	U.S. Geological Survey
Issue date:	03/31/2008
From:	Johnny Eads; NRC/NRR/ADRA/DPR
To:	Timothy Debey; US Dept of Interior, Geological Survey (USGS)
	Young P, NRC/NRR/ADRA/DPR, 415-4094
Shared Package
ML080080551	List: ML080810420; ML080810447;
References
	50-274/OL-08-01
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March 31, 2008 Mr. Timothy DeBey Reactor Director U.S. Geological Survey Box 25046 - Mail Stop 424 Denver Federal Center Denver, CO 80225

SUBJECT:

INITIAL EXAMINATION REPORT NO. 50-274/OL-08-01, U.S. GEOLOGICAL SURVEY

Dear Mr. DeBey:

During the week of March 10, 2008, the NRC administered an operator licensing examination at your GSTR Facility. The examination was conducted according to NUREG-1478, "Operator Licensing Examiner Standards for Research and Test Reactors," Revision 2. Examination questions and preliminary findings were discussed with those members of your staff identified in the enclosed report at the conclusion of the examination.

In accordance with Title 10 of the Code of Federal Regulations Section 2.390, 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 NRC's Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.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 via 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. 50-274

Enclosures:

1. Initial Examination Report No. 50-274/OL-08-01

2. Written examination with facility comments incorporated cc without enclosures:

See next page

March 31, 2008 Mr. Timothy DeBey Reactor Director U.S. Geological Survey Box 25046 - Mail Stop 424 Denver Federal Center Denver, CO 80225

SUBJECT:

INITIAL EXAMINATION REPORT NO. 50-274/OL-08-01, U.S. GEOLOGICAL SURVEY

Dear Mr. DeBey:

During the week of March 10, 2008, the NRC administered an operator licensing examination at your GSTR Facility. The examination was conducted according to NUREG-1478, "Operator Licensing Examiner Standards for Research and Test Reactors," Revision 2. Examination questions and preliminary findings were discussed with those members of your staff identified in the enclosed report at the conclusion of the examination.

In accordance with Title 10 of the Code of Federal Regulations Section 2.390, 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 NRC's Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.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 via 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. 50-274

Enclosures:

1. Initial Examination Report No. 50-274/OL-08-01

2. Written examination with facility comments incorporated cc without enclosures:

See next page DISTRIBUTION w/ encls.:

PUBLIC PRTB r/f RidsNRRDPRPRTA RidsNRRDPRPRTB Facility File (CHart) O-12 G-15 ADAMS ACCESSION #: ML080810447 TEMPLATE #:NRR-074 OFFICE PRTB:CE IOLB:LA E PRTB:SC NAME PYoung pty CHart cah JEads jhe DATE 03/19/2008 3/28/2008 3/31/2008 OFFICIAL RECORD COPY

U.S. Geological Survey Docket No. 50-274 cc:

Mr. Brian Nielsen Environmental Services Manager 480 S. Allison Pkwy.

Lakewood, CO 80226 Mr. Eugene W. Potter State of Colorado Radiation Management Program HMWM-RM-B2 4300 Cherry Creek Drive South Denver, CO 80246 Mr. Timothy DeBey Reactor Director U.S. Geological Survey Box 25046 - Mail Stop 424 Denver Federal Center Denver, CO 80225 Test, Research, and Training Reactor Newsletter University of Florida 202 Nuclear Sciences Center Gainesville, FL 32611

U. S. NUCLEAR REGULATORY COMMISSION OPERATOR LICENSING INITIAL EXAMINATION REPORT REPORT NO.: 50-274/OL-08-01 FACILITY DOCKET NO.: 50-274 FACILITY LICENSE NO.: R-113 FACILITY: UNITED STATES GEOLOGICAL SURVEY EXAMINATION DATES: 3/10 - 12/2008 SUBMITTED BY: ___________/RA/___________ 3/14/2008 Phillip T. Young, Chief Examiner Date

SUMMARY

Two Reactor Operator applicants took the examination given the week of March 10, 2008. Both applicantss passed all portions of the examination.

REPORT DETAILS

1. Examiners: Phillip T. Young, Chief Examiner, NRC

2. Results:

RO PASS/FAIL SRO PASS/FAIL TOTAL PASS/FAIL Written 2/0 0/0 2/0 Operating Tests 2/0 0/0 2/0 Overall 2/0 0/0 2/0

3. Exit Meeting:

Phillip T. Young, NRC, Examiner Timothy DeBey, Reactor Director The examiner thanked Mr. DeBey for the facility support during the examinations. There were no generic issues identified during the examinations. The facility provided the following comments on the written examination:

ENCLOSURE 1

Hi Phillip, Here are the comments that I have about four of the exam questions. Please let me know if you have any questions.

Thanks, Tim Question A.006 I believe the answer should be (a) An immediate increase in the prompt neutron population.

I do not have your Reference, but DOE-HDBK-1019/2-93, Reactor Theory, page 14 states, The positive reactivity insertion is followed immediately by a small immediate power increase called the prompt jump. This power increase occurs because the rate of production of prompt neutrons changes abruptly as the reactivity is added. Recall from an earlier module that the generation time for prompt neutrons is on the order of 10-13 seconds. The effect can be seen in Figure 2. After the prompt jump, the rate of change of power cannot increase any more rapidly than the built-in time delay the precursor half-lives allow. Therefore, the power rise is controllable, and the reactor can be operated safely.

Question B.010 I believe the answer should be (a) Both dose rates are the same.

This is because G-M tubes are not sensitive to energy, as given in the Reference. Since the curie level of both sources are the same and a G-M tube is used as the detector, the dose rates will be the same.

Question C.001 I believe the answers should be: a=0, b=1, c=0, d=3, e= 4&5, f= 2&6; which is a change of answers to c, e, and f compared to the answer sheet.

The answers for c and f are related because the detectors for the two safety channels are uncompensated ion chamber, as indicated in the text file that accompanied the I & C system block diagram. The answers for e should be 4&5 because these are the two fuel temperature channels of the instrumentation system and they both have thermocouples for sensors.

Question C.011 I believe this answer should be (c) Drive completely up, rod is down, no magnet contact.

The Reference, GA Control Console Operators Manual, p 1-5, states, For each shim rod and the regulator rod, the small square box at the bottom of the drive mechanism indicates the status of the rod magnet power. If the box is yellow, then magnet power is ON. If the box is black, then magnet power is OFF. If the magnet power is OFF, then the control rod will be dropped (down) and the magnet will not be making contact.

NRC Response:

All comments are accepted as stated.

US NRC LICENSE EXAMINATION U.S. GEOLOGICAL SURVEY, NO. 50-274 WRITTEN EXAMINATION with ANSWER KEY - OL-08-01 MARCH 10, 2008

A. REACTOR THEORY, THERMODYNAMICS & FACILITY OPERATING CHARACTERISTICS Page 7 of 26 Question: A.001 [1.0 point] (1.0)

The effective neutron multiplication factor, Keff , is defined as:

a. Keff = production/(absorption + leakage).

b. Keff = (production + leakage)/absorption.

c. Keff = (absorption + leakage)/production.

d. Keff = absorption/(production + leakage).

Answer: A.001 a.

Reference:

Glasstone & Sesonske, Nuclear Reactor Engineering, 3rd Edition, pg.

149.

Question: A.002 [1.0 point] (2.0)

As the moderator temperature increases, the resonance escape probability:

a. increases, since the moderator becomes less dense.

b. decreases, since the time required for a neutron to reach thermal energy increases.

c. remains constant, since the effect of moderator temperature change is relatively small.

d. increases, since the moderator-to-fuel ratio increases.

Answer: A.002 b.

Reference:

Glasstone & Sesonske, Nuclear Reactor Engineering, 3rd Edition, pg.

264.

Question: A.003 [1.0 point] (3.0)

Which one of the following is the definition of the FAST FISSION FACTOR?

a. The ratio of the number of neutrons produced by fast fission to the number produced by thermal fission

b. The ratio of the number of neutrons produced by thermal fission to the number produced by fast fission

c. The ratio of the number of neutrons produced by fast and thermal fission to the number produced by thermal fission

d. The ratio of the number of neutrons produced by fast fission to the number produced by fast and thermal fission Answer: A.003 c.

Reference:

p. 3-16.

A. REACTOR THEORY, THERMODYNAMICS & FACILITY OPERATING CHARACTERISTICS Page 8 of 26 Question: A.004 [1.0 point] (4.0)

Two different neutron sources were used during two reactor startups. The source used in the first startup emits ten times as many neutrons per second as the source used in the second startup.

Assume all other factors are the same for the second startup. Which ONE of the following states the expected result at criticality?

a. Neutron flux will be higher for the first startup.

b. Neutron flux will be higher for the second startup.

c. The first startup will result in a higher rod position (rods further out of the core).

d. The second startup will result in a higher rod position (rods further out of the core).

Answer: A.004 a.

Reference:

Which ONE of the following is the principal source of energy (heat generation) in the reactor 15 minutes following a reactor shutdown from extended operation at full power?

a. Production of delayed neutrons.

b. Subcritical multiplication of neutrons.

c. Spontaneous fission of U-238.

d. Decay of fission products.

Answer: A. 005 d.

Reference:

A step insertion of positive reactivity in a critical reactor causes a momentary rapid increase in the neutron population, known as a prompt jump. Which ONE of the following describes the cause of this increase?

a. An immediate increase in the prompt neutron population.

b. A shortening of the delayed neutron generation when power increases.

c. The positive reactivity insertion due to the rapid fuel temperature coefficient feedback.

d. The step insertion produces a rate of reactivity addition which exceeds the delayed neutron fraction, eff .

Answer: A.006 a b. Per facility comment.

Reference:

Glasstone & Sesonske, Nuclear Reactor Engineering, 3rd Edition, pg. 240.

A. REACTOR THEORY, THERMODYNAMICS & FACILITY OPERATING CHARACTERISTICS Page 9 of 26 Question: A.007 [1.0 point] (7.0)

A reactor is subcritical with a Keff of 0.984. and a count rate of 1500 cps on the startup instrumentation.

Rods are withdrawn until the count rate is 6000 cps. At this point, the value of Keff is:

a. 0.992

b. 0.994

c. 0.996

d. 0.998 Answer: A.007 c.

Reference:

Which ONE of the following is the reason for operating with thermal neutrons rather than fast neutrons?

a. As neutron energy increases, neutron absorption in non-fuel materials increases exponentially.

b. Probability of fission is increased since thermal neutrons are less likely to leak out of the core.

c. The absorption cross-section of U-235 is much higher for thermal neutrons.

d. The fuel temperature coefficient becomes positive as neutron energy increases.

Answer: A.008 c.

Reference:

The equations which describe the operation of the installed neutron source in the reactor are:

a. Am-241 -> alpha + Np-237 Be-9 + alpha -> C-12 + neutron

b. Am-241 -> alpha + Np-237 B-10 + alpha -> N-13 + neutron

c. Am-241 -> beta + Cm-241 Be-9 + beta -> Li-8 + neutron

d. Am-241 -> beta + Cm-241 B-10 + beta -> Be-9 + neutron Answer: A.009 a.

Reference:

Chart of Nuclides

A. REACTOR THEORY, THERMODYNAMICS & FACILITY OPERATING CHARACTERISTICS Page 10 of 26 Question: A.010 [1.0 point] (10.0)

Which ONE of the following conditions describes a critical reactor?

a. Keff = 1; k/k () = 1

b. Keff = 1; k/k () = 0

c. Keff = 0; k/k () = 1

d. Keff = 0; k/k () = 0 Answer: A.010 b.

Reference:

Question: A.011 [1.0 point] (11.0)

In choosing a Fuel-Moderator material ZrH ratio of 1.7, the limiting effect or factor verses other ratios is the:

a. relative widely spaced cracks that occur, in higher ratios, due to the elevated temperatures produced by pulsing

b. large volume changes associated with the with the phase transformations that occur in higher ratios.

c. zirconium hydride chemical reactivity rate with water associated with lower ratios.

d. hydrogen gas over pressure at higher fuel temperatures.

Answer: A.011 d.

Reference:

A. REACTOR THEORY, THERMODYNAMICS & FACILITY OPERATING CHARACTERISTICS Page 11 of 26 Question: A.012 [1.0 point] (12.0)

The reactor supervisor tells you the reactor is shutdown with a shutdown margin of 12%. An experimenter inserts an experiment in the core and nuclear instrumentation increases from 100 counts per minute to 200 counts per minute. What is the new Keff of the reactor?

a. 0.920

b. 0.946

c. 0.973

d. 1.000 Answer: A.012 b.

Reference:

Standard NRC Question:

1 1 K eff1 = = = 0.892857 1+ SDM 1+ 0.12 CR1(1 K eff1 ) = CR 2 (1 K eff2 );

100 1 K eff2 = (1 0.892857)= (0.0535715) 200 K eff2 = 0.9464285 Question: A.013 [1.0 point] (13.0)

Two reactors are identical except that Reactor 1 has a beta fraction of 0.0072 and Reactor 2 has a beta fraction of 0.0060. An equal amount of positive reactivity is inserted into both reactors. Which ONE of the following will be the response of Reactor 2 compared to Reactor 1?

a. The resulting power level is lower.

b. The resulting power level is higher.

c. The period of the power increase is shorter.

d. The period of the power increase is longer.

Answer: A.013 c.

Reference:

Glasstone & Sesonske, Nuclear Reactor Engineering, 3rd Edition, pg. 237.

= ( /) + [(-)/eff]. The reactor with the smaller beta fraction will have the shortest period.

Question: A.014 [1.0 point] (14.0)

Inelastic scattering can be described as a process whereby a neutron collides with a nucleus and:

a. recoils with the same kinetic energy it had prior to the collision.

b. is absorbed by the nucleus, with the nucleus emitting a gamma ray.

c. recoils with a lower kinetic energy, with the nucleus emitting a gamma ray.

d. recoils with a higher kinetic energy, with the nucleus absorbing a gamma ray.

Answer: A.014 c.

Reference:

Glasstone & Sesonske, Nuclear Reactor Engineering, 3rd Edition, pg. 70.

A. REACTOR THEORY, THERMODYNAMICS & FACILITY OPERATING CHARACTERISTICS Page 12 of 26 Question: A.015 [1.0 point] (15.0)

A reactor with an initial population of 1x108 neutrons is operating with Keff = 1.001. Considering only the increase in neutron population, how many neutrons (of the increase) will be prompt when the neutron population changes from the current generation to the next? Assume eff = 0.007.

a. 700.

b. 7,000.

c. 99,300.

d. 100,000.

Answer: A.015 c.

Reference:

Glasstone & Sesonske, Nuclear Reactor Engineering, 3rd Edition, pg. 230.

Increase = 1.001x108 - 1x108 = 1x105. Prompt neutron population = 0.993x1x105 = 99,300.

Question: A.016 [1.0 point] (16.0)

Starting with a critical reactor at low power, a control rod is withdrawn from position X and reactor power starts to increase. Neglecting any temperature effects, in order to terminate the increase with the reactor again critical but at a higher power, the control rod must be:

a. inserted back to position X.

b. inserted deeper than position X.

c. inserted, but not as far as position X.

d. inserted, but exact position depends on power level.

Answer: A.016 a.

Reference:

In the absence of any temperature effects, core reactivity returns to zero when the Rods are returned to their original positions.

Question: A.017 [1.0 point] (17.0)

Which ONE of the following is the description of a thermal neutron?

a. A neutron which possesses thermal rather than kinetic energy.

b. The primary source of thermal energy increase in the reactor coolant during reactor operation.

c. A neutron that has been produced in a significant time (on the order of seconds) after its initiating fission took place.

d. A neutron that experiences no net change in kinetic energy after several collisions with atoms of the diffusing media.

Answer: A.017 d.

Reference:

Glasstone & Sesonske, Nuclear Reactor Engineering, 3rd Edition, pg. 13.

A. REACTOR THEORY, THERMODYNAMICS & FACILITY OPERATING CHARACTERISTICS Page 13 of 26 Question: A.018 [1.0 point] (17.0)

In a subcritical reactor with Keff = 0.861, +0.104 delta k/k reactivity is added. As a result, the new Keff is:

a. 0.899

b. 0.946

c. 0.989

d. 1.0574 Answer: A.018 b.

Reference:

Glasstone & Sesonske, Nuclear Reactor Engineering, 3rd Edition, pg. 231.

Initial reactivity = (K - 1)/K = (0.861 - 1)/0.861 = -0.1614 delta k/k. Final reactivity = -0.1614 + 0.104 = -

0.0574 delta k/k; k = 1/(1 + 0.0574) = 0.946 Question: A.019 [1.0 point] (19.0)

Which ONE of the following is responsible for the constant rate of power change several minutes after a reactor scram from full power?

a. The decay of the longest-lived delayed neutron precursors.

b. The decay of the shortest-lived delayed neutron precursors.

c. The mean average decay of the delayed neutron precursors.

d. The decay of fission product gammas producing photoneutrons.

Answer: A.019 a.

Reference:

Glasstone & Sesonske, Nuclear Reactor Engineering, 3rd Edition, pg. 120.

Question: A.020 [1.0 point] (20.0)

As a reactor continues to operate over time, for a constant power level, the thermal neutron flux:

a. decreases, due to the increase in fission product poisons.

b. increases, in order to compensate for fuel depletion.

c. decreases, because fuel is being depleted.

d. remains the same.

Answer: A.020 b.

Reference:

Glasstone & Sesonske, Nuclear Reactor Engineering, 3rd Edition, pg. 105.

B. NORMAL/EMERGENCY PROCEDURES & RADIOLOGICAL CONTROLS Page 14 of 26 Question: B.001 [1.0 point] (1.00)

In accordance with the Technical Specifications, which ONE situation below is permissible when the reactor is operating?

a. A pool water level of 17 feet.

b. A bulk pool temperature of 70 degrees C.

c. A total, absolute reactivity worth of in-core experiments of $3.00.

d. A pulse reactivity insertion of 2.9% delta k/k.

Answer: B.001 c.

Reference:

Technical Specifications,Section I.4.

Question: B.002 [2.0 points, 1/2 each] (3.00)

Match type of radiation (a thru d) with the proper penetrating power (1 thru 4)

a. Gamma 1. Stopped by thin sheet of paper

b. Beta 2. Stopped by thin sheet of metal

c. Alpha 3. Best shielded by light material

d. Neutron 4. Best shielded by dense material Answer: B.002 a. = 4; b. = 2; c. = 1 d. = 3

Reference:

Standard Question:

Question: B.003 [1.0 point] (4.00)

A system or component is defined as operable by Technical Specifications if:

a. a channel check has been performed.

b. a functional test has been performed.

c. it has no outstanding testing requirements.

d. it is capable of performing its intended function.

Answer: B.003 d.

Reference:

Technical Specifications, Section A.5.

B. NORMAL/EMERGENCY PROCEDURES & RADIOLOGICAL CONTROLS Page 15 of 26 Question: B.004 [1.0 point] (5.00)

In accordance with the Emergency Plan, the emergency planning zone is:

a. the area within the operations boundary.

b. the area within the site boundary.

c. the Nuclear Science building.

d. the assembly area.

Answer: B.004 a.

Reference:

USGS Emergency Plan, Definitions.

Question: B.005 [1.0 point] (6.00)

An automatic scram signal required by the Technical Specifications when operating in the square-wave mode is:

a. short period.

b. high power level.

c. high fuel temperature.

d. loss of percent power channel high voltage.

Answer: B.005 b.

Reference:

Technical Specifications, Table I.

Question: B.006 [1.0 point] (7.00)

In accordance with the Emergency Plan, emergency action levels are:

a. documented instructions that detail the implementation actions and methods required to achieve the objectives of the emergency plan.

b. projected radiological doses or dose commitment values to individuals that warrant protective action following a release of radioactive material.

c. conditions which call for immediate actions, beyond the scope of normal operating procedures, to avoid an accident or to mitigate the consequences of one.

d. specific instrument readings, observations, radiological dose or dose rates, etc., which may be used as thresholds for establishing emergency classes and initiating appropriate emergency measures.

Answer: B.006 d.

Reference:

USGS Emergency Plan, Definitions.

B. NORMAL/EMERGENCY PROCEDURES & RADIOLOGICAL CONTROLS Page 16 of 26 Question: B.007 [1.0 point] (8.00)

Which ONE of the following activities requires the direct presence (supervision) of a senior licensed operator?

a. Reactor power calibration.

b. Removal of one (1) fuel element.

c. Control rod drop time measurement.

d. Removal of control rod for inspection.

Answer: B.007 b.

Reference:

PROCEDURE FOR FUEL LOADING AND UNLOADING Question: B.008 [2.0 points 0.5 each] (10.00)

Research and Test reactors primarily worry about two isotopes N16 and Ar41. Identify the approximate half-life and gamma energy for each. (Each item has only one answer.)

Isotope Radiological Parameters 41

a. Ar half-life 1) 1.8 sec 2) 1.8 min 3) 1.8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 4) 1.8 day

b. Ar41 gamma energy 1) 10 KeV 2) 100 KeV 3) 1 MeV 4) 10 MeV

c. N16 half-life 1) 7 sec 2) 7 min 3) 7 hour8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months 4) 7 day 16

d. N gamma energy 1) 6 kev 2) 60 keV 3) 600 keV 4) 6 Mev Answer B.008 a. = 3; b. = 3; c. = 1; d. = 4

Reference:

Standard NRC Question Question: B.009 [1.0 point] (11.00)

The continuous air monitor is out of service waiting on back ordered parts for repair. As a result:

a. the reactor cannot be operated.

b. the reactor can continue to operate.

c. the reactor can continue to operate only if the alarm setpoints of the area radiation monitors are lowered.

d. the reactor can continue to operate only if the monitor is replaced by a portable monitor with a read-out and capable of alarming.

Answer: B.009 a.

Reference:

Technical Specifications, Section F.2.

B. NORMAL/EMERGENCY PROCEDURES & RADIOLOGICAL CONTROLS Page 17 of 26 Question: B.010 [1.0 point] (12.00)

Two different gamma point sources have the same curie strength. The gammas from Source A have an energy of 1 Mev, and the gammas from Source B have an energy of 2 Mev. The dose rate from each source is measured at a distance of 10 feet using a G-M tube. Which ONE of the following statements is correct?

a. Both dose rates are the same.

b. The dose rate of Source B is half that of Source A.

c. The dose rate of Source B is two times that of Source A.

d. The dose rate of Source B is four times that of Source A.

Answer: B.010 a c. Per facility comment.

Reference:

G-M tubes are not sensitive to energy.

Question: B.011 [1.0 point] (13.00)

Two centimeters of lead placed in a beam of gamma rays reduces the radiation level from 400 mR/hr to 200 mR/hr. Which ONE of the following is the total thickness of lead that would reduce the gamma radiation level from 400 mR/hr to 50 mR/hr?

a. 3 cm.

b. 4 cm.

c. 6 cm

d. 8 cm.

Answer: B.011 c.

Reference:

Each 2 cm. of lead reduces the radiation level by a factor of 2. 6 cm reduces It by a factor of 8.

Question: B.012 [1.0 point] (14.00)

In accordance with the Technical Specifications, given the control rod worths and excess reactivity below, calculate the minimum shutdown margin.

Shim rod = 1.8% delta k/k Regulating rod = 2.5% delta k/k Safety rod = 2.0% delta k/k Transient rod = 2.1% delta k/k Excess reactivity = 4% delta k/k

a. 1.8% delta k/k.

b. 1.9% delta k/k.

c. 4.4% delta k/k.

d. 8.4% delta k/k.

Answer: B.012 b.

Reference:

Technical Specifications, Section E.5.

Shutdown margin + Excess reactivity = Rod worth (excluding the most reactive rod)

B. NORMAL/EMERGENCY PROCEDURES & RADIOLOGICAL CONTROLS Page 18 of 26 Question: B.013 [1.0 point] (15.00)

In accordance with GSTR Procedure No. 2, Reactor Power Calibration, a percent power indicating channel which is incorrect by more than 1.5% power is adjusted to agree with the true power by:

a. mechanically adjusting the pointers on the meters to give the proper indication.

b. adjusting the compensating voltage to the ion chamber to give the proper indication.

c. adjusting the high voltage to the ion chamber to give the proper indication.

d. changing its position to give the proper indication.

Answer: B.013 d.

Reference:

GSTR Procedure No. 2, pg. 2.

Question: B.014 [1.0 point] (16.00)

Which of the following is not considered to cause a whole body exposure?

a. Ar-41

b. I-131

c. Xe-133

d. Kr-88 Answer: B.014 b.

Reference:

Glasstone/Sesonske - Chapter 9 Question: B.015 [1.0 point] (17.00)

Your Reactor Operator license expires after _____ years.

a. 2

b. 4

c. 6

d. 8 Answer B.015 c.

Reference:

10CFR55.55(a)

B. NORMAL/EMERGENCY PROCEDURES & RADIOLOGICAL CONTROLS Page 19 of 26 Question: B.016 [1.0 point] (18.00)

A Reactor Operator has not been actively performing functions and duties for over 6 months. To resume functions authorized by the license, issued under part 50.51, what is the minimum hours required to perform shift functions under the direction of an operator or senior operator as appropriate and in the position to which the individual will be assigned?

a. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

b. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months

c. 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months

d. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Answer: B.016 d.

Reference:

10 CFR 55.53(f)(2)

Question: B.017 [1.0 point] (19.00)

According to the Pocket Dosimeter Drift Check Procedure, after a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months drift check period, what is the maximum drift allowable if the dosimeter is to remain in service?

a. <1% from the initial reading

b. <5% from the initial reading

c. <1% of meter full scale

d. <2% of meter full scale Answer: B.017 d.

Reference:

USGS Pocket Dosimeter Drift Check Procedure Question: B.018 [1.0 point] (20.00)

A pocket dosimeter is being calibrated in accordance with the POCKET DOSIMETER CALIBRATION PROCEDURE. At the end of the predetermined exposure period (typically 25 minutes), what should the meter be reading?

a. 110 mr

b. 125 mr/hr

c. 180 mr

d. 250 mr Answer: B.018 a.

Reference:

USGS POCKET DOSIMETER CALIBRATION PROCEDURE

C. FACILITY AND RADIATION MONITORING SYSTEMS Page 20 of 26 Question: C.001 [2.0 points, 0.333 points each] (2.00)

Match the measuring channel listed in Column A with the correct detector position on the GSTR Instrument Diagram attached. If an item in Column A is not used select 0. from Column B. Items listed in Column A may be used more than once or not at all.

Column A Column B

a. Proportional counter 0.

b. Fission chamber 1.

c. Compensated Ion Chamber 2.

d. RTD 3.

e. Thermocouple 4.

f. Uncompensated Ion Chamber 5.

6.

Answer: C.001 a. = 0; b = 1; c. = 0 2 & 6; d. = 3; e = 4 & 5; f. = 0 2 & 6 Changes to correct answers per facility comment.

Reference:

Control System Block Diagram.

Question: C.002 [2.0 points, 0.4 points each] (4.00)

Match the item listed in Column A with the correct position on the GSTR TRIGA Cross Section attached. Items listed in Column A may be used more than once or not at all.

Column A Column B

a. Pneumatic Transfer System 1.

b. Isotope Removal Tube 2.

c. Control rod Drive 3.

d. Reflector 4.

e. Rotary Specimen Rack 5.

6.

7.

8.

9.

10.

Answer: C.002 a. = 2; b. = 3; c. = 1; d. = 7; e. = 4 .

Reference:

Hazard Summary Report Figure 5 - 1

C. FACILITY AND RADIATION MONITORING SYSTEMS Page 21 of 26 Question: C.003 [2.0 points, 0.4 points each] (6.00)

Match the item listed in Column A with the correct position on the GSTR Pneumatic - electrical Transient - Rod Drive diagram attached. Items listed in Column A may be used more than once or not at all.

Column A Column B

a. Drive Motor 1.

b. Potentiometer Drive Gear 2.

c. Shock Absorber 3.

d. Damper 4.

e. Piston 5.

6.

7.

8.

9.

Answer: C.003 a. = 7; b. = 6; c. = 1; d. = 9; e. = 5 or 8 .

Reference:

Hazard Summary Report Figure 5 - 7.

Question: C.004 [1.0 point] (7.00)

Which ONE of the following conditions is NOT a requirement to enter the Square Wave mode of operation?

a. Power less than 1 kW.

b. Period greater than 26 seconds.

c. Reactor in Steady State mode.

d. All control rods and transient rod above the down limit.

Answer: C.004 d.

Reference:

GA Control Console Operators Manual.

C. FACILITY AND RADIATION MONITORING SYSTEMS Page 22 of 26 Question: C.005 [1.0 point] (8.00)

Which ONE of the following radiation monitors is interlocked with the ventilation system?

a. CAM.

b. Reactor console.

c. Top of the reactor.

d. Entrance to reactor laboratory.

Answer: C.005 a.

Reference:

Hazards Summary Report, Section 9.3.1.

Question: C.006 [1.0 point] (9.00)

Which ONE of the following will NOT automatically activate Building 15 evacuation alarm?

a. Fire alarm.

b. Air particulate monitor.

c. Gaseous stack monitor.

d. Continuous air monitor alarm.

Answer: C.006 c.

Reference:

Emergency Procedures, Section 7.3.3.

Question: C.007 [1.0 point] (10.00)

The neutron source used in the reactor is a:

a. plutonium-beryllium source.

b. polonium-americium source.

c. polonium-beryllium source.

d. antimony-beryllium source.

Answer: C.007 c.

Reference:

Hazards Summary Report, Section 5.6.

C. FACILITY AND RADIATION MONITORING SYSTEMS Page 23 of 26 Question: C.008 [1.0 point] (11.00)

At the GSTR total primary loop flow equals ________ , and flow through the purification loop is normally ______ .

a. 250 gpm, 25 gpm

b. 300 gpm, 25 gpm

c. 350 gpm, 20 gpm

d. 400 gpm, 20 gpm Answer: C.0 08 c.

Reference:

Hazard Summary Report Figure 5 - 12 Question: C.009 [1.0 point] (12.00)

Which ONE of the following describes the purpose of the Draw Tube in a control rod drive assembly?

a. Actuates the rod down microswitch.

b. Provides rod full out position indication.

c. Connects to the electromagnet to move the rod.

d. Automatically engages the control rod on a withdraw signal.

Answer: C.009 c.

Reference:

Hazards Summary Report, Section 5.4.1.

Question: C.010 [1.0 point] (13.00)

Which ONE is NOT an input to the Regulating Rod Servo?

a. % demand potentiometer

b. Linear power channel

c. Rod raising interlock

d. Period channel Answer: C.010 c.

Reference:

GA TRIGA Instrumentation Maintenance Manual

C. FACILITY AND RADIATION MONITORING SYSTEMS Page 24 of 26 Question: C.011 [1.0 point] (14.00)

Limit switches mounted on each drive assembly provide switching for console lights. What is the significance of a "MAGENTA" rod color and a "BLACK" magnet box?

a. Reactor scram, control rod drive down.

b. Drive between limits, rod down, no magnet current.

c. Drive completely up, rod is down, no magnet contact.

d. Rod and drive completely withdrawn, magnet making contact.

Answer: C.011 c d. Per facility comment.

Reference:

GA Control Console Operator's Manual pg. 1-5 Question: C.012 [1.0 point] (15.00)

One of the differences between the transient control rod and the other control rods is that the transient control rod:

a. contains an air void while the other rods do not.

b. is 43 inches long while the other rods are 37 inches long.

c. has a reactivity rod worth of about twice that of each of the other rods.

d. has 21 inches of borated graphite in the upper section while the other rods have 15 inches of borated graphite in the upper section.

Answer: C.012 a.

Reference:

Hazards Summary Report, Section 5.3.1.

Question: C.013 [1.0 point] (16.00)

Which ONE of the following statements is NOT a reason for maintaining a minimum reactor pool level during reactor operation?

a. Provide Net Positive Suction Head (NPSH) to the reactor water pump

b. Proper "Dash Pot" action for the control rods during a scram

c. Ensure proper operation of the pool skimmer

d. Provide shielding from the core Answer: C.013 a.

Reference:

GA TRIGA Mechanical Maintenance and Operating Manual pg. 4-13

C. FACILITY AND RADIATION MONITORING SYSTEMS Page 25 of 26 Question: C.014 [1.0 point] (17.00)

An important design feature of the fuel-moderator elements is:

a. an enrichment of 90%, allowing for prolonged operations before refueling.

b. the reduced neutron absorption provided by the zirconium cladding.

c. the prompt negative temperature coefficient of reactivity.

d. the inclusion of the burnable poison erbium.

Answer: C.014 c.

Reference:

Hazards Summary Report, Section 5.1.

Question: C.015 [1.0 point] (18.00)

Which ONE of the following statements describes the moderating properties of Zirconium Hydride?

a. Elevation of the hydride temperature increases the probability that a thermal neutron will escape the fuel-moderator element before being captured.

b. The probability that a neutron will return to the fuel element before being captured elsewhere is a function of the temperature of the hydride.

c. The ratio of hydrogen atoms to zirconium atoms affects the moderating effectiveness for slow neutrons.

d. The hydride mixture is very effective in slowing down neutrons with energies below 0.025 eV.

Answer: C.015 a.

Reference:

GA TRIGA Mark I Reactor Hazards Analysis.

Question: C.016 [1.0 point] (19.00)

Maximum fuel temperatures are expected to occur in the:

a. B-ring fuel elements.

b. D-ring fuel elements.

c. F-ring fuel elements.

d. G-ring fuel elements.

Answer: C.016 a.

Reference:

Hazards Summary Report, Section 7.3.3.

C. FACILITY AND RADIATION MONITORING SYSTEMS Page 26 of 26 Question: C.017 [1.0 point] (20.00)

The reactor is in the AUTOMATIC mode at a power level of 500 kW. The neutron detector from which the control system receives its input signal fails low (signal suddenly goes to zero). As a result:

a. the control system inserts the regulating rod to reduce power, to try to match the power of the failed detector.

b. the control system drops out of the AUTOMATIC mode into the MANUAL mode.

c. the control system withdraws the regulating rod to try to increase power.

d. the reactor scrams.

Answer: C.017 c.

Reference:

Hazards Summary Report, Section 5.5.2.

ML080810447

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