Text

Initial Examination Report No. 50-243/OL-09-01, Oregon State University
	ML090641146
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Site:	Oregon State University
Issue date:	03/11/2009
From:	Johnny Eads; Research and Test Reactors Branch B
To:	Reese S; Oregon State University
	Young P, NRC/NRR/ADRA/DPR, 415-4094
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March 11, 2009 Dr. Steven E. Reese, Director Oregon State University Radiation Center, A100 Corvallis, OR 97331-5903

SUBJECT:

INITIAL EXAMINATION REPORT NO. 50-243/OL-09-01, OREGON STATE UNIVERSITY

Dear Dr. Reese:

During the week of February 23, 2009, the NRC administered an operator licensing examination at your Oregon State University Reactor. 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 H. Eads Jr., Chief Research and Test Reactors Branch B Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Docket No. 50-243

Enclosures:

1. Initial Examination Report No. 50-243/OL-09-01

2. Facility Comments with NRC Resolution

3. Corrected Written Examination cc without enclosures: See next page

March 11, 2009 Dr. Steven E. Reese, Director Oregon State University Radiation Center, A100 Corvallis, OR 97331-5903

SUBJECT:

INITIAL EXAMINATION REPORT NO. 50-243/OL-09-01, OREGON STATE UNIVERSITY.

Dear Dr. Reese:

During the week of February 23, 2009, the NRC administered an operator licensing examination at your Oregon State University Reactor. 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 H. Eads Jr., Chief Research and Test Reactors Branch B Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Docket No. 50-243

Enclosures:

1. Initial Examination Report No. 50-243/OL-09-01

2. Facility Comments with NRC Resolution

3. Corrected Written Examination cc without enclosures: See next page DISTRIBUTION w/ encls.:

PUBLIC PRTB r/f RidsNRRDPRPRTA RidsNRRDPRPRTB Facility File (CHart) O-13 D-07 ADAMS ACCESSION #:ML090641146 TEMPLATE #:NRR-074 OFFICE PRTB:CE IOLB:LA E

PRTB:SC NAME PYoung pty CRevelle car JEads jhe DATE 3/6/09 3/10/09 3/11/09 OFFICIAL RECORD COPY

Oregon State University Docket No. 50-243 cc:

Mayor of the City of Corvallis Corvallis, OR 97331 David Stewart-Smith Oregon Office of Energy 625 Marion Street, N.E.

Salem, OR 97310 Dr. John Cassady, Vice President for Research Oregon State University Administrative Services Bldg., Room A-312 Corvallis, OR 97331-5904 Mr. Todd Keller Reactor Administrator Oregon State University Radiation Center, A-100 Corvallis, OR 97331-5903 Dr. Todd Palmer, Chairman Reactor Operations Committee Oregon State University Radiation Center, A-100 Corvallis, OR 97331-5904 Test, Research, and Training Reactor Newsletter University of Florida 202 Nuclear Sciences Center Gainesville, FL 32611 Eric C Woolstenhulme Idaho National Laboratory P.O. Box 1625, Idaho Falls, ID 83415-3740

U. S. NUCLEAR REGULATORY COMMISSION OPERATOR LICENSING INITIAL EXAMINATION REPORT REPORT NO.:

50-243/OL-09-01 FACILITY DOCKET NO.:

50-243 FACILITY LICENSE NO.:

R-106 FACILITY:

Oregon State University EXAMINATION DATES:

February 24 and 25, 2009 SUBMITTED BY:

___/RA/_______

03/03/2009 Chief Examiner Date

SUMMARY

During the week of February 23, 2009, the NRC administered operator licensing examinations to one Reactor Operator candidate and one Senior Operator Upgrade. Both candidates 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 1/0 0/0 1/0 Operating Tests 1/0 1/0 2/0 Overall 1/0 1/0 2/0

3.

Exit Meeting:

Phillip T. Young, NRC, Examiner Gary M. Wachs, Reactor Supervisor, Oregon State University Todd Keller, Reactor Administrator, Oregon State University The NRC examiner thanked the facility staff for their cooperation during the examination. The examiner reported no generic weaknesses:

FACILITY COMMENTS March 3, 2009 From:

OSTR Staff To:

NRC Examiner Phillip Young

Subject:

RO written examination for Wade Marcum administered 2/24/2009 Mr. Young, In accordance with your request, we are providing this feedback regarding the written RO exam administered to Wade Marcum on Feb 24, 2009. We would like to make the following observations and suggestions:

Question A9 refers to Kexcess. This terminology is not used at the OSTR. The more common term for us is Core Excess. The question is also somewhat confusing and could perhaps be phrased better. We suggest Which ONE of the following is the Technical Specification shutdown margin?

Question A13. The symbols that appear on the typed copy of the exam show up only as boxes.

It is assumed that the missing symbols should be arrows.

Question A18. The question refers to points A and B on the given figure, but the figure has no labeled points on it. The correct answer can be inferred, but points A and B should be present for clarity.

Question B3. We feel that this level of knowledge is not appropriate for an RO examination.

The RO candidate should be aware that record retention requirements exist and where they can be found, but requiring the candidate to memorize the retention requirements does not make him a safer operator.

Question B4. We feel that this level of knowledge is not appropriate for an RO examination.

The RO candidate should be aware that reporting requirements exist and where they can be found, but requiring the candidate to memorize the reporting requirements does not make him a safer operator.

Question B8. The question is somewhat confusing and could perhaps be phrased better. It is unclear exactly what is being asked.

Question B16. We feel that this level of knowledge is not appropriate for an RO examination.

The RO candidate should be aware that notification requirements for all classes of emergencies exist and where these requirements can be found. Requiring the candidate to know which organizations are notified does not make him a safer operator.

Question C13. The question refers to a referencelector. This probably should be reflector.

ENCLOSURE 2

In addition to the above comments, we would like to request the removal of the three questions that we feel are inappropriate for an RO level examination (questions B3, B4 and B16). You have stated that at certain facilities where the Reactor Supervisor is permitted to be away from the facility for a limited amount of time, you expect that the RO candidates will have a deeper level of knowledge in certain areas, and should be able to perform the Reactor Supervisor duties in the event that he is unable to respond. While this is understandable, we do not concur with this opinion for the following reasons:

If this reasoning is followed to its natural conclusion, we should not be training any candidates for RO positions. All operators at the OSTR should be SROs.

The logic assumes that an RO should be able to perform the primary duties of the Reactor Supervisor in the event of an emergency. In the OSTR Emergency Plan, there are no specified primary duties for the Reactor Supervisor during an emergency situation, except to be a member of the radiological assessment team. This is already an expected duty for all RO level operators.

We do not feel that is it desirable that an RO be able to perform the secondary emergency duties of the Reactor Supervisor (i.e. emergency coordinator and recovery coordinator). Our emergency plan has clearly identified lines of succession for each position. If the Reactor Supervisor is unable to respond, someone else will automatically be designated to assume his responsibilities. Under no circumstances will the RO be expected to perform the duties of any other position formally identified in the Emergency Response Plan (except those of the RO).

If you would like additional information, please feel free to contact us at any time. We look forward to hearing your reply, and your official notification of the result of the RO and SRO examinations.

NRC RESOLUTION OF FACILLITY COMMENTS Question A.009 Comment accepted. Official copy of this examination and future use of this question will reflect the facility comment. No change in grading required.

Question A.013 Upon review, it appears the symbol did not make the transition from WordPerfect to MS Word.

Comment accepted. Official copy of this examination and future use of this question will reflect the facility comment. No change in grading required.

Question A.018 Upon review, it appears the diagram did not make the transition from WordPerfect to MS Word.

Comment accepted. Official copy of this examination and future use of this question will reflect the facility comment. No change in grading required.

Question B.003 Comment accepted. Official copy of this examination will reflect the facility comment and show the question as deleted. Grading of the examination will reflect the deleted question.

Question B.004 Comment accepted. Official copy of this examination will reflect the facility comment and show the question as deleted. Grading of the examination will reflect the deleted question.

Question B.008 Comment accepted. Official copy of this examination and future use of this question will reflect the facility comment. No change in grading required.

Question B.016 Comment accepted. Detailed questions on implementation of the Emergency Plan are best handled during the Walkthrough portions of the Operating Test. Official copy of this examination will reflect the facility comment and show the question as deleted. Grading of the examination will reflect the deleted question.

Question C.013 Comment accepted. Corrected the typographical error. Official copy of this examination and future use of this question will reflect the facility comment. No change in grading required.

To clarify my remarks concerning those facilities where the Reactor Supervisor is permitted to be away from the facility for a limited amount of time, it is expected that the Reactor Operator be able to perform the initial requirements for E-Plan implementation.

Section A Reactor Theory, Thermo & Fac. Operating Characteristics Page 8 of 21 Oregon State University Operator License Examination Written Exam with Answer Key February 24, 2009

Section A Reactor Theory, Thermo & Fac. Operating Characteristics Page 1 of 21 Question A.001 (1.0 point)

{1.0}

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

a. Thermal Utilization Factor (f)

b. Fast Non-Leakage Factor (Lf)

c. Reproduction Factor ()

d. Fast Fission Factor ()

Answer:

A.001 a.

Reference:

OSU Training Manual Volume 3, Control Rods, p. 16, 1st ¶.

Question A.002

[1.0 point]

{2.0}

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:

A.002 d.

Reference:

OSU Training Manual Volume 3, Question and Answer on p. 52.

Question A.003

[1.0 point]

{3.0}

The source was removed from an operating reactor. Later, the source was reinstalled and the Reactor Operator noted reactor power increasing LINEARLY. What was the condition of the reactor when the source was inserted? (Assume source has no reactivity worth, and no other changes in reactor parameters.)

The reactor was

a. very subcritical

b. slightly subcritical

c. exactly critical

d. slightly supercritical Answer:

A.003 c.

Reference:

OSU Training Manual Volume 3, Criticality in the Presence of a Neutron Source p. 50.

Section A Reactor Theory, Thermo & Fac. Operating Characteristics Page 2 of 21 Question A.004

[1.0 point]

{4.0}

Reactor power increases from 30 watts to 60 watts in one minute. Reactor period is

a. 30 seconds

b. 42 seconds

c. 60 seconds

d. 87 seconds Answer:

A.004 d.

Reference:

( )

ln P

P t

t t 60 sec ln 2 86.56 0

=

Question A.005

[1.0 point]

{5.0}

A characteristic peculiar to TRIGA fuel is that it has a relatively large (and quickly acting)

a. void coefficient.

b. pressure coefficient.

c. bath temperature coefficient.

d. fuel temperature coefficient.

Answer:

A.005 d.

Reference:

OSU Training Manual Volume 3, Prompt Negative Temperature Coefficient of Reactivity, pp. 66 & 67.

Question A.006

[1.0 point]

{6.0}

An experimenter makes an error loading a rabbit sample. Injection of the sample results in a 100 millisecond period. If the scram setpoint is 1.25 MW and the scram delay time is 0.1 seconds, WHICH ONE of the following is the peak power of the reactor at shutdown. (Assume Rabbit system is operational for this question.)

a. 1.25 MW

b. 2.5 MW

c. 3.4 MW

d. 12.5 MW Answer:

A.006 c.

Reference:

P = P0 et/, P = 1.25 Mwatt x e0.1/0.1 = 1.25 x e = 3.3979.

Section A Reactor Theory, Thermo & Fac. Operating Characteristics Page 3 of 21 Question A.007

[1.0 point]

{7.0}

Which ONE of the following is the correct reason that delayed neutrons enhance control of the reactor?

a. There are more delayed neutrons than prompt neutrons.

b. Delayed neutrons take longer to reach thermal equilibrium.

c. Delayed neutrons increase the average neutron generation time.

d. Delayed neutrons are born at higher energies than prompt neutrons and therefore have a greater effect.

Answer:

A.007 c.

Reference:

Oregon State University Training Manual Volume 3, Delayed Neutrons, p. 30 1st ¶.

Question A.008

[1.0 point]

{8.0}

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

a. The rod speed.

b. Reactor power.

c. The flux shape.

d. The amount of fuel in the core.

Answer:

A.008 c.

Reference:

OSU Training Manual Vol. 3, p. 17 Question A.009

[1.0 point]

{9.0} Question changed per facility comment.

A reactor (not OSTR) has the following reactivity characteristics:

Core Excess Kexcess = $2.50 Standard Rod 1 = $2.25 Reg Rod = $1.10 Standard Rod 2 = $2.30 Which ONE of the following is the Technical Specification shutdown margin?

Which ONE of the following is the shutdown margin allowable by Technical Specifications.

(NOTE: All rods are able to scram, same condition as OSTR Tech Spec.)

a. $5.65

b. $4.00

c. $3.10

d. $0.85 Answer:

A.009 d.

SDM = Worth of rods less Kexcess less reactivity of most worth rod. SDM = $5.65 - $2.50 - 2.30 = 5.65 - $4.80 = 0.85

Reference:

OSU Training Manual Vol. 3, p. 29

Section A Reactor Theory, Thermo & Fac. Operating Characteristics Page 4 of 21 Question A.010

[1.0 point]

{10.0}

A thin foil target of 10% copper and 90% aluminum is in a thermal neutron beam.

Given:

a Cu = 3.79 barns a Al = 0.23 barns s Cu = 7.90 barns s Al =1.49 barns Which ONE of the following reactions has the highest probability of occurring? A neutron

a. scattering reaction with aluminum

b. scattering reaction with copper

c. absorption in aluminum

d. absorption in copper Answer:

A.010 a.

Reference:

Glasstone, S. and Sesonske, A, Nuclear Reactor Engineering, Kreiger Publishing, Malabar, Florida, 1991, § Question A.011

[1.0 point]

{11.0}

You are increasing reactor power on a steady +26 second period. How long will it take to increase power by a factor of 1000?

a. 60 seconds (1 minute)

b. 180 seconds (3 minutes)

c. 300 seconds (5 minutes)

d. 480 seconds (8 minutes)

Answer:

A.011 b.

Reference:

ln (P/P0) x period = time, ln(1000) x 26 = 6.908 x 26 = 179.6 180 seconds Question A.012

[1.0 point]

{12.0}

Which ONE of the following statements is the definition of REACTIVITY?

a. A measure of the core's fuel depletion.

b. A measure of the core's deviation from criticality.

c. Equal to 1.00 K/K when the reactor is critical.

d. Equal to 1.00 K/K when the reactor is prompt critical.

Answer:

A.012 b.

Reference:

OSU Training Manual Vol. 3, p. 10

Section A Reactor Theory, Thermo & Fac. Operating Characteristics Page 5 of 21 Question A.013

[1.0 point]

{13.0} Question changed per facility comment.

Which ONE of the following correctly describes the generation of neutrons from the Am-Be source?

a.

95Am241 - 93Np237 + 24; 24 + 4Be9 - [6C13]* - 6C12 + 0n1

b. 95Am241 - 96Np241 + -10 + ;

00 + 4Be9 - [4Be9]* - 4Be8 + 0n1

c.

95Am241 - 96Np241 + -10 + ;

-10 + 4Be9 -

[3Li9]* - 3Li8 + 0n1

d. 95Am241

-

[S.F.] - 2 fission products + 0n1 Answer:

A.013 a.

Reference:

OSU Training Manual Vol. 3, pp. 41-43.

Question A.014

[1.0 point]

{14.0}

Which ONE of the following describes the difference between a moderator and reflector?

a. A reflector increases the fast non-leakage factor and a moderator increases the thermal utilization factor.

b. A reflector decreases the neutron production factor and a moderator decreases the fast non-leakage factor.

c. A reflector increases the neutron production factor and a moderator increases the fast fission factor.

d. A reflector decreases the thermal utilization factor and a moderator increases the fast fission factor.

Answer:

A.014 a.

Reference:

OSU Training Manual Vol. 3, pp. 12 and 15.

Question A.015

[1.0 point]

{15.0}

After a week of full power operation, Xenon will reach its peak following a shutdown in approximately:

a. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

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

c. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

d. 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months Answer:

A.015 b.

Reference:

OSU Training Manual Vol. 3, p. 24.

Section A Reactor Theory, Thermo & Fac. Operating Characteristics Page 6 of 21 Question A.016

[1.0 point]

{16.0}

Regulating rod worth for a reactor is 0.001 K/K/inch. Moderator temperature increases by 9ºF, and the regulating rod moves 41/2 inches inward to compensate. The moderator temperature coefficient Tmod is

a. +5 x 10-4

b. -5 x 10-4

c. +2 x 10-5

d. -2 x 10-5 Answer:

A.016 a.

Reference:

OSU Training Manual Vol. 3, p. 19 Also, (4.5 x 0.001) ÷ 9 = 0.0045 ÷ 9 = 0.0005 = 5 x 10-4 Question A.017

[1.0 point]

{17.0}

Which ONE of the following is the difference between prompt and delayed neutrons? Prompt neutrons

a. account for less than 1% of the neutron population, while delayed neutrons account for the rest.

b. are released during fast-fission events, while delayed neutrons are released during the decay process.

c. are released during the fission process (fast & thermal), while delayed neutrons are release during the decay process.

d. are the dominating factor in determining reactor period, while delayed neutrons have little effect on reactor period.

Answer:

A.017 c.

Reference:

OSU Training Manual Vol. 3, p. 30.

Section A Reactor Theory, Thermo & Fac. Operating Characteristics Page 7 of 21 Question A.018

[1.0 point]

{18.0} Question changed per facility comment.

Shown below is a trace of reactor period as a function of time. Between points A and B reactor power is:

a. continually increasing.

b. continually decreasing.

c. increasing, then decreasing.

d. constant.

Delete diagram below.

Replace diagram from above with the one shown below.

Answer:

A.018 a.

Reference:

Standard NRC QUESTION Question A.019

[1.0 point]

{19.0}

Which ONE of the following is the correct definition of effective for a TRIGA reactor? The relative amount of delayed neutrons

a. per generation corrected for leakage.

b. per generation corrected for resonance absorption.

c. per generation corrected for time after the fission event.

d. per generation corrected for both leakage and resonance absorption.

Answer:

A.019 d

Reference:

OSU Training Manual Vol. 3, p. 30.

Section A Reactor Theory, Thermo & Fac. Operating Characteristics Page 8 of 21 Question A.020

[1.0 point]

{20.0}

During a reactor startup, criticality occurred at a lower rod height than the last startup. Which ONE of the following reasons could be the cause?

a. Xe135 peaked.

b. Moderator temperature increased.

c. Adding an experiment with positive reactivity.

d. Maintenance on the control rods resulted in a slightly faster rod speed.

Answer:

A.020 c.

Reference:

OSU Training Manual Vol. 3, p. 30.

Section B Normal/Emergency Procedures and Radiological Controls Page 9 of 21 Question B.001

[2 points, 0.5 each]

{2.0}

Identify each of the following conditions as either a safety limit (SL), or a limiting safety system setting (LSSS), or limiting conditions for operation (LCO).

a) The maximum available excess reactivity based on the reference core condition shall not be exceeded $7.55.

b) There is a limiting reactivity for pulse operation such that a maximum fuel element temperature shall not exceed 830 degree C.

c) The temperature in a TRIGA fuel element shall not exceed 1,150 degree C under any mode of operation.

d) The temperature in an instrumented fuel element shall be equal to or less than 510 degree C.

Answer:

B.001 a. = LCO, b. = LCO, c. = SL,

d. = LSSS

Reference:

TSs 2.1, 2.2, and 3.1 Question B.002

[1 point]

{3.0}

A radiation survey is performed on a radioactive material and indicates a dose rate of 35 Rem/hr at contact.

Five hours later, a survey shows 1.5 Rem/hr. How long does it take for a material to decay from 35 Rem/hr to 100 mRem/hr?

a) 7.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months b) 8.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months c) 9.4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months d) 10.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months Answer:

B.002 c.

Reference:

DR = DR*e -t 1.5 rem/hr =35 rem/hr* e -(5hr)

Ln(1.5/35) = -*5 --> =0.623; solve for t: Ln(.1/35)=-0.623 (t) t=9.4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months

Section B Normal/Emergency Procedures and Radiological Controls Page 10 of 21 Question deleted from examination per facility comment.

Question B.003

[2 points, 0.5 each]

{5.0}

Match the requirement for retaining the reactor records in column A with the correct time period in column B (Note: some time period in column B may be used more than once or not at all).

Column A Column B

a. Offsite environmental monitoring surveys

1. Lifetime of the reactor facility

b. Principal maintenance activities

2. 1 year

c. Medical examination for reactor operator

3. 2 years

d. Fuel inventories

4. At least 5 years Answer:

B.003 a. = (1);

b. = (4);

c. = (3);

d. = (4)

Reference:

TS 6.8 Question deleted from examination per facility comment.

Question B.004

[2 points, 0.5 each]

{7.0}

Match the events of a reportable occurrence in columns A with the correct time period required to notify to the NRC in column B (Note: some time period in column B may be used more than once or not at all)

Column A Column B

a. Safety Limit is exceeded

1. 1 working day

b. Power level is exceeded to 1.2 MW(t)

2. 30 days

c. An unanticipated $1.1 change

3. 60 days

d. Level 1 change in the organization

4. 1 year Answer:

B.004 a. = (1);

b. = (1);

c. = (1);

d. = (2)

Reference:

TS 6.7 Question B.005

[1 point]

{8.0}

A radioactive material is decayed at a rate of 30% per day. Determine its half-life?

a) 0.36 day b) 0.58 day c) 1.94 days d) 2.30 days Answer:

B.005 c

Reference:

DR = DR*e -t 30% is decayed, so 70% is still there 70% =100%* e -(1day)

Ln(70/100) = -*1 -->=0.356 t1/2=ln(2)/ -->.693/.356 t=1.94 days

Section B Normal/Emergency Procedures and Radiological Controls Page 11 of 21 Question B.006

[2 points, 0.5 each]

{10.0}

Match type of radiation in column A with its penetrating power in column B (Note: use each item only once)

Colum A Column B a) Alpha

1. Stopped by thin sheet of paper b) Beta

2. Best shielded by water c) Gamma

3. Best shielded by lead (Pb) d) Neutron

4. Stopped by thin sheet of metal Answer:

B.006 a. = (1);

b. = (4);

c. = (3);

d. = (2)

Reference:

Standard NRC Health Physics question Question B.007

[1 point]

{11.0}

A 0.5 curies radioactive source, emitting 1.33 Mev gamma, is to be stored in the reactor building. Assume that no shielding is used; which one of the following doses is indicated at 28.2 ft. away?

a) 2 mrem/hr b) 5 mrem/hr c) 10 mrem/hr d) 20 mrem/hr Answer:

B.007 b.

Reference:

6(C)EN=6*0.5*1.33 = 3.99 Rem/hr at 1 ft.

3.99*square(1) = x*square(28.2)

X=.005 R/hr or = 5mrem/hr Question B.008

[1 point]

{12.0} Question changed per facility comment.

Which ONE of the following is NOT a major product released due to thermal fission of a U235 atom?

Which one of the followings is NOT a distribution from fission energy?

a) Fission fragments b) Prompt neutron c) Delayed neutron d) Prompt Gamma rays e) Excited X-rays Answer:

B.008 e

Reference:

Section B Normal/Emergency Procedures and Radiological Controls Page 12 of 21 Question B.009

[1 point]

{13.0}

Which one of the following is a dose limit for individual members of the public from the licensed operation in a year?

a) 2 mRem b) 50 mRem c) 100 mRem d) 5 Rem Answer:

B.009 c

Reference:

10 CFR 20.1301 Question B.010

[1 point]

{14.0}

Fill in blank The licensee shall ensure that the dose equivalent to the embryo/fetus during the entire pregnancy, due to the occupational exposure of a declared pregnancy woman, does not exceed _________.

a) 100 mRem b) 500 mRem c) 700 mRem d) 5,000 mRem Answer:

B.010 b

Reference:

10 CFR 20.1207 Question B.011

[1 point]

{15.0}

An uranium source emitted 150 mrem/hr is stored in hot cell. Select proper posting for that source a) caution-radiation area b) danger-radiation area c) caution-high radiation area d) grave danger-very high radiation area Answer:

B.011 c.

Reference:

10 CFR 20.1902

Section B Normal/Emergency Procedures and Radiological Controls Page 13 of 21 Question B.012

[1 point]

{16.0}

TS 6.1, Organization, provides information regarding the various management levels and their responsibilities. Which one of the following organization levels is responsible for guidance, oversight, and technical support of reactor operations?

a) Vice-President for Research (Level 1) b) Radiation Center Director (Level 2) c) Reactor Administrator (Level 3) d) Reactor Supervisor (Level 3)

Answer:

B.012 c

Reference:

TS 6.1 Question B.013

[1 point]

{17.0}

Which one of the following is the definition of Total Effective Dose Equivalent (TEDE) specified in 10 CFR 20?

a) The sum of the deep dose equivalent and the committed effective dose equivalent b) The dose that your whole body is received from the source, but excluded from the deep dose.

c) The sum of the external deep dose and the organ dose d) The sum of thyroid dose and external dose Answer:

B.013 a

Reference:

10 CFR 1003 Question B.014

[1 points]

{18.0}

Which one of the following is the class of emergency if actual or projected whole body radiation levels at the site boundary are measured of 20 mRem/hr?

a) Class 0 b) Class 1 c) Class 2 d) Class 3 Answer:

B.014 c

Reference:

EP, Table 5.1

Section B Normal/Emergency Procedures and Radiological Controls Page 14 of 21 Question B.015

[1 point]

{19.0}

Which one of the following equipments is NOT required to be operating during fuel element handling?

a) The area radiation monitor near the fuel movement b) The continuous air monitor c) The ventilation system and stack monitor d) Primary water activity monitor Answer:

B.015 d

Reference:

Procedure OSTROP 11, Fuel Element Handling Procedures.

Question deleted from examination per facility comment.

Question B.016

[1 point]

{20.0}

During reactor class II emergency, the emergency coordinator is required to notify the offsite support agencies. Which one of the following organizers is NOT required to be informed?

a) NRC b) Oregon Department of Energy c) American Nuclear Insurers d) Department of Homeland Security Answer:

B.016 d

Reference:

EP 7.3.1

Section C Facility and Radiation Monitoring Systems Page 15 of 21 Question C.001

[1.0 point]

{1.0}

Which ONE of the following is NOT an interlock associated with pulsing operations.

a. Period greater than 50 seconds.

b. Transient Rod fully inserted.

c. Power less than 1 Kwatt.

d. Switch in Pulse Mode.

Answer: C.001 a.

Reference:

Volume 2, pages 23-28, OSU Triga Manual Question C.002

[1.0 point]

{2.0}

What design feature minimizes flux peaking in the central thimble.

a. An aluminum plug.

b. A Zirconium plug.

c. A cadmium plug.

d. Filling with N2.

Answer: C.002 a.

Reference:

OSU Training Manual Vol I, p, 89 second paragraph.

Question C.003

[1.0 point]

{3.0}

Which ONE of the following is the reason for the holes located at the bottom of the Central Thimble Assembly?

a. To allow for evacuation of the water in the thimble.

b. To allow cooling flow through the thimble, at power.

c. To fit over pins in the safety support plate for proper alignment.

d. To allow for fasteners to bolt the thimble to the bottom support plate.

Answer: C.003 a.

Reference:

OSU Training Manual, p. 90

Section C Facility and Radiation Monitoring Systems Page 16 of 21 Question C.004

[1.0 point]

{4.0}

Which ONE of the following methods is the normal procedure for preventing basin water in the cooling tower from freezing?

a. Running fans in reverse.

b. Use of Hand-held heat guns.

c. Heaters built into water sump.

d. Steam connection from University facilities.

Answer:

C.004 c

Reference:

OSU Training Manual, p. 126, last paragraph.

Question C.005

[1.0 point]

{5.0}

Which ONE of the following components, in the purification system, is PRIMARILY responsible for maintaining the primary coolant system conductivity low.

a. The post-demineralizer filter

b. The pre-demineralizer filter

c. The surface skimmer

d. The demineralizer Answer:

C.005 d.

REFERENCE:

OSTR Training Manual Vol. I, p. 116.

Question C.006

[1.0 point]

{6.0}

What is the purpose of the Cadmium Lined In-Core Irradiation Tube (CLICIT)

a. To allow irradiation of samples by gammas within the core.

b. To allow irradiation of samples by neutrons with an energy level of less than 0.5 ev.

c. To allow irradiation of samples by neutrons with an energy level greater than 0.5 ev.

d. To allow irradiation of samples by alphas produced by the neutron interaction with the cadmium.

Answer:

C.006 c.

Reference:

OSTR Training Manual, Vol. I, p. 81

Section C Facility and Radiation Monitoring Systems Page 17 of 21 Question C.007

[1.0 point]

{7.0}

The ventilation system is designed to maintain reactor bay pressure slightly negative pressure with respect to the atmospheric pressure. If the outside atmospheric pressure increases, which ONE of the following actions will automatically occur to compensate the reactor bay pressure? A pressure regulator will generate a signal to...

a. Increase the Reactor Bay Supply fan speed to increase bay pressure.

b. Decrease the Reactor Bay Exhaust fan speed to increase bay pressure.

c. Go more open on a damper in the ventilation supply ducting increasing bay pressure.

d. Go more closed on a damper in the ventilation exhaust ducting increasing bay pressure.

Answer:

C.007 d.

Reference:

OSTR Training Manual, Vol. I p. 148.

QUESTION C.008

[1.0 point]

{8.0}

While operating in AUTOMATIC mode, the reactor operator depresses the UP button for a control rod. At the same time, the AUTOMATIC circuit energizes to drive the regulating rod up.

Which ONE of the following will actually take place.

a. The ONE ROD WITHDRAWAL interlock applies, neither rod will move.

b. The ONE ROD WITHDRAWAL interlock does not apply, both rods will move.

c. The ONE ROD WITHDRAWAL interlock applies, only the CONTROL ROD will move.

d. The ONE ROD WITHDRAWAL interlock applies, only the REGULATING ROD will move.

Answer:

C.008 b.

Reference:

OSTR Training Manual Vol II, p. 9.

Section C Facility and Radiation Monitoring Systems Page 18 of 21 Question C.009

[1.0 point] {9.0}

Which ONE of the following is the actual method used to generate the rod position indication seen on the control panel?

a. A potentiometer linked to the rod drive motor.

b. Voltage changes generated by the movement of a lead screw between two coils of a transformer.

c. A series of reed switches which as the rod moves up close to generate a current proportional to rod position.

d. A servo motor connected to the UP and DN buttons which when either button is depressed generates a signal proportional to rod speed.

Answer:

C.009 a.

Reference:

OSTR Training Manual, Vol I, fig. 1.26 TRIGA Control Rod Drive Mechanism, p.48.

Question C.010

[1.0 point]

{10.0}

WHICH ONE of the following detectors is used primarily to measure N16 released to the environment?

a. Stack Gas Monitor

b. Air Particulate Monitor

c. Area Radiation Monitor Channel # 5

d. NONE, N16 has too short a half-life to require environmental monitoring.

Answer:

C.010 d.

Reference:

Standard NRC QUESTION

Section C Facility and Radiation Monitoring Systems Page 19 of 21 Question C.011

[1.0 point]

{11.0}

You (the console operator) receive a report of thick black smoke coming from the demin pump.

Where would you send someone to de-energize the breaker that supplies the pump?

a. Room 106 to Sub Distribution Panel A

b. First Floor Hallway Panel "F"

c. Room 106 to Panel "G"

d. Reactor Bay Panel "A" Answer:

C.011 c.

Reference:

OSTROP 22.0 Emergency Power System, Fig. 22.1 One-Line Schematic Power Distribution Question C.012

[1.0 point]

{12.0}

To detect Neutrons, the Uncompensated Ion Chambers are lined with

a.

5B10

b.

6C12

c.

92U235

d.

94Pu239 Answer:

C.012 a.

Reference:

ORST Trn Man Volume 2, § III.C.

Question C.013

[1.0 point]

{13.0} Question changed per facility comment.

Match each purpose in column A with its associated fuel element component listed in column B.

Column A Column B

a. moderator

1. Graphite

b. Referencelector reflector

2. Zirconium-Hydride

c. resonance absorber

3. Erbium

d. burnable poison Answer:

C.013

a. = 2;

b. = 1;

c. = 3;

d. = 3

Reference:

ORST Trn Man Volume 1.

Section C Facility and Radiation Monitoring Systems Page 20 of 21 Question C.014

[1.0 point]

{14.0}

Which one of the following ranges is used for the area radiation monitors (ARMs) at OSTR?

a) 0.1 mR/hr - 10 R/hr b) 1 mR/hr - 10 R/hr c) 10 mR/hr - 20 R/hr d) 20 mR/hr - 10 R/hr Answer:

C.014 a.

Reference:

EP 7.1.2 Question C.015

[2.0 points, 0.333 each]

{16.0}

Identify whether the equipment listed remains energized (ALWAYS ON), reenergizes after emergency generator starts [20 seconds] (EMERGENCY Powered) or remains deenergized (NO POWER) following a loss of normal AC power to the facility.

a. Argon Fan

b. Public Address System

c. Fire Alarm System

d. Stack Monitor Pump

e. Cypher Locks

f. Rabbit Fan Answer:

C.015 a. = NO POWER;

b. = ALWAYS ON;

c. = EMERGENCY;

d. = EMERGENCY; e. = EMERGENCY; f.= NO POWER

Reference:

OSTROP 22.0 Emergency Power System, Figures 22.1, and 22.2

Section C Facility and Radiation Monitoring Systems Page 21 of 21 Question C.016

[1.0 point]

{17.0}

The rabbit system uses air to send samples into and out of the reactor core. Which ONE of the following is thelargest radiological problem associated with using air?

a.

6C14

b.

7N16

c.

8018

d.

18Ar41 Answer:

C.016 d.

Reference:

Standard NRC question.

Question C.017

[1.0 point]

{18.0}

Which one of the following correctly describes the operation of a Thermocouple?

a. A bi-metallic strip which winds/unwinds due to different thermal expansion constants for the two metals, one end is fixed and the other moves a lever proportional to the temperature change.

b. a precision wound resistor, placed in a Wheatstone bridge, the resistance of the resistor varies proportionally to temperature changes.

c. a liquid filled container which expands and contracts proportional to temperature changes, one part of which is connected to a lever.

d. a junction of two dissimilar metals, generating a potential (voltage) proportional to temperature changes.

Answer:

C.017 d.

Reference:

Standard NRC question.

Section C Facility and Radiation Monitoring Systems Page 22 of 21 Question C.018

[2.0 points, 0.333 each]

{20.0}

Identify which core row (A through G) each of the following core components is in.

a. Source

b. Central Thimble

c. Instrumented Fuel Element

d. Transient Rod

e. Safety Rod

f. Rabbit Terminus Answer:

C.018 a. = G;

b. = A;

c. = B;

d. = C;

e. = D;

f. = G

Reference:

OSU Training Material Operator Reference Data, Last Figure.

ML090641146

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