Text

Examination Report 50-602/OL-12-01,University of Texas at Austin
	ML12081A109
Person / Time
Site:	University of Texas at Austin
Issue date:	04/11/2012
From:	Johnny Eads; Division of Policy and Rulemaking
To:	Whaley P; University of Texas at Austin
References
	50-602/OL-12-01
	Download: ML12081A109 (43)
	v • d • e

April 11, 2012

Paul Whaley, Associate Director Nuclear Engineering Teaching Lab University of Texas at Austin NETL-PRC Bldg 159 10100 Burnet Rd Austin, TX 78758

SUBJECT:

EXAMINATION REPORT NO. 50-602/OL-12-01, UNIVERSITY OF TEXAS AT AUSTIN

Dear Mr. Whaley:

During the week of March 12, 2012, the Nuclear Regulatory Commission (NRC) administered an operator licensing examination at your TRIGA 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. John T. Nguyen, at (301) 415-4007 or via internet e-mail John.Nguyen@nrc.gov.

Sincerely,

/RA/ Johnny H. Eads, Jr., Chief Research and Test Reactors Oversight Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation

Docket No. 50-602

Enclosures:

1. Examination Report No. 50-602/OL-12-01

2. Written examination with facility comments incorporated

cc: Michael Krause, Reactor Supervisor cc w/o enclosures: See next page Paul Whaley, Associate Director Nuclear Engineering Teaching Lab University of Texas at Austin NETL-PRC Bldg 159 10100 Burnet Rd Austin, TX 78758

SUBJECT:

EXAMINATION REPORT NO. 50-602/OL-12-01, UNIVERSITY OF TEXAS AT AUSTIN

Dear Mr. Whaley:

During the week of March 12, 2012, the Nuclear Regulatory Commission (NRC) administered an operator licensing examination at your TRIGA 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. John T. Nguyen, at (301) 415-4007 or via internet e-mail John.nguyen@nrc.gov.

Sincerely,

/RA/ Johnny H. Eads, Jr., Chief Research and Test Reactors Oversight Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation

Docket No. 50-602

Enclosures:

1. Examination Report No. 50-602/OL-12-01

2. Written examination with facility comments incorporated

cc: Michael Krause, Reactor Supervisor cc w/o enclosures: See next page DISTRIBUTION: PUBLIC RidsNrrDprPrta RidsNrrDprPrtb Facility File (CRevelle) O-07 F-08

ADAMS ACCESSION #: ML12081A109 TEMPLATE #: NRR-079 Office PROB/CE IOLB/OLA PROB/BC Name JNguyen CRevelle JEads Date 3/21/12 3/21/12 3/21/12 OFFICIAL RECORD COPY

University of Texas Docket No. 50-602 cc: Governor=s Budget and Planning Office P.O. Box 13561 Austin, TX 78711 Bureau of Radiation Control State of Texas 1100 West 49 th Street Austin, TX 78756 Mr. William Powers, Jr., President University of Texas at Austin Nuclear Engineering teaching Laboratory Austin, TX 78758 Mr. Roger Mulder Office of the Governor P.O. Box 12428 Austin, TX 78711 Test, Research, and Training Reactor Newsletter University of Florida 202 Nuclear Sciences Center Gainesville, FL 32611

ENCLOSURE 1 U. S. NUCLEAR REGULATORY COMMISSION OPERATOR LICENSING INITIAL EXAMINATION REPORT

REPORT NO.: 50-602/OL-12-01

FACILITY DOCKET NO.: 50-602

FACILITY LICENSE NO.: R-129

FACILITY: UNIVERSITY of TEXAS at AUSTIN

EXAMINATION DATES: March 12, 2012

SUBMITTED BY: ____/RA/______________________ 03/21/2012 John Nguyen, Chief Examiner Date

SUMMARY

During the week of March 12, 2012, the NRC administered the operator licensing examinations to one Reactor Operator candidate and one Senior Reactor Operator Upgraded candidate. All candidates passed all portions of the examination.

REPORT DETAILS

1. Examiners: John Nguyen, Examiner, NRC

2. Results:

RO PASS/FAILSRO PASS/FAILTOTAL PASS/FAIL Written 1/0 0/01/0 Operating Tests1/0 1

/01/0 Overall 1/0 1

/02/0 3. Exit Meeting:

John Nguyen, NRC Michael Krause, Reactor Manager Hunter Fuentes, Senior Reactor Operator The examiner thanked the facility for their support during the examination and their comments on questions B.1, C.1, and C.6. The examiner indicated no generic weaknesses were noted during the operating examination.

U. S. NUCLEAR REGULATORY COMMISSION RESEARCH AND TEST REACTOR OPERATOR LICENSING EXAMINATION

FACILITY: UNIVERSITY of TEXAS at AUSTIN REACTOR TYPE: TRIGA DATE ADMINISTERED: 03/12/2012 CANDIDATE:

INSTRUCTIONS TO CANDIDATE:

Answers are to be written on the answer sheets provided. Points for each question are indicated in brackets for each question. You must score 70% in each section to pass. Examinations will be picked up three (3) hours after the examination starts.

% of Category % of Candidates Category Value Total Score Value Category 20.00 33.33 A. Reactor Theory, Thermodynamics and Facility Operating Characteristics 20.00 33.33 B. Normal and Emergency Operating Procedures and Radiological Controls

20.00 33.33 C. Plant and Radiation Monitoring Systems FINAL GRADE % TOTALS All work done on this examination is my own. I have neither given nor received aid. ______________________________________ Candidate's Signature

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

1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.

2. After the examination has been completed, you must sign the statement on the cover sheet indicating that the work is your own and you have neither received nor given assistance in completing the examination. This must be done after you complete the examination.

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

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

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

6. The point value for each question is indicated in [brackets] after the question.

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

8. To pass the examination you must achieve a grade of 70 percent or greater in each category.

9. There is a time limit of three (3) hours for completion of the examination.

10. When you have completed and turned in you examination, leave the examination area Page 7 EQUATION SHEET

1 Curie = 3.7 x 10 10 dis/sec 1 kg = 2.21 lbm 1 Horsepower = 2.54 x 10 3 BTU/hr 1 Mw = 3.41 x 10 6 BTU/hr T UA = H m = T c m = Q p (k)2)-( = P 2max seconds 10 x 1 = -4* sec-1 eff 0.1 = )K-(1 CR = )K-(1 CR eff 2 eff 1 2 1 K-1 S = SCR eff -26.06 = SUR eff K-1 K-1 = M eff eff 1 0 CR CR = K-1 1 = M 2 1 eff 10 P = P SUR(t)0 e P = P t 0 P -)-(1 = P 0 - =

eff*- + = K)K-(1 = SDM eff eff K x k K - K = eff eff eff eff 2 1 1 2 0.693 = T K 1)-K ( = eff eff e DR= DR t-0 R 6CiE(n) = DR 2 d DR = d DR 2 2 2 1 2 1 Peak)-( = Peak)-(1 1 2 2 2 2 Page 8 1 BTU = 778 ft-lbf F = 9/5 C + 32 1 gal (H 2O) 8 lbm C = 5/9 (F - 32)

Section A L Theory, Thermo, and Facility Characteristics Page 9 A N S W E R S H E E T Multiple Choice (Circle or X your choice) If you change your answer, write your selection in the blank.

A001 a b c d __

A002 a b c d __

A003 a b c d __

A004 a b c d __

A005 a b c d __

A006 a b c d __

A007 a b c d __

A008 a b c d __

A009 a b c d __

A010 a b c d __

A011 a b c d __

A012 a b c d __

A013 a b c d __

A014 a b c d __

A015 a b c d __

A016 a b c d __

A017 a b c d __

A018 a b c d __

A019 a b c d __

A020 a b c d __

Section B Normal/Emerg. Procedures & Rad Con Page 10 A N S W E R S H E E T Multiple Choice (Circle or X your choice) If you change your answer, write your selection in the blank.

B001 a b c d __

B002 a __ b __ c __ d ___

B003 a b c d __

B004 a b c d __

B005 a b c d __

B006 a b c d __

B007 a b c d __

B008 a b c d __

B009 a b c d __

B010 a b c d __

B011 a b c d __

B012 a b c d __

B013 a b c d __

B014 a b c d __

B015 a b c d __

B016 a __ b __ c __ d ___

B017 a b c d __

B018 a b c d __

B019 a b c d __

B020 a b c d __

Section C Facility and Radiation Monitoring Systems Page 11 A N S W E R S H E E T Multiple Choice (Circle or X your choice) If you change your answer, write your selection in the blank.

C001 a b c d __

C002 a b c d __

C003 a b c d __

C004 a b c d __

C005 a __ b __ c __ d __

C006 a b c d __

C007 a b c d __

C008 a b c d __

C009 a b c d __

C010 a b c d __

C011 a __ b __ c __ d __

C012 a b c d __

C013 a b c d __

C014 a b c d __

C015 a b c d __

C016 a b c d __

C017 a b c d __

C018 a b c d __

C019 a b c d __

C020 a b c d __

- - - - END OF EXAMINATION *****

Section A L Theory, Thermo, and Facility Characteristics Page 12 Question A.1 [1.0 point]

In a subcritical reactor, K eff is increased from 0.931 to 0.946. Which ONE of the following is the amount of reactivity that was added to the core?

a. 1.50 %k/k b. 1.70 %k/k c. 3.40 %k/k d. 5.40 %k/k Question A.2 [1.0 point]

Which ONE of the following isotopes has the highest thermal neutron cross section?

a. Xe-135

b. Sm-149

c. U-235

d. U-238 Question A.3 [1.0 point]

Which ONE of the following best describes the effects of moderator temperature increase on neutron multiplication?

a. Resonance escape probability ; Thermal non-leakage; Rod worth b. Resonance escape probability ; Thermal non-leakage; Rod worth

c. Resonance escape probability ; Thermal non-leakage; Rod worth d. Resonance escape probability ; Thermal non-leakage; Rod worth

Section A L Theory, Thermo, and Facility Characteristics Page 13 Question A.4 [1.0 point]

Two critical reactors at low power are identical, except that Reactor 1 has a beta fraction of 0.0065 and Reactor 2 has a beta fraction of 0.0072. Which ONE of the following best describes the response if an equal amount of positive reactivity is inserted into both reactors?

a. Period of the Reactor 1 will be longer than the period of the Reactor 2

b. Period of the Reactor 1 will be shorter than the period of the Reactor 2

c. Power of the Reactor 1 will be higher than the power of the Reactor 2

d. Power of the Reactor 1 will be lower than the power of the Reactor 2 Question A.5 [1.0 point]

Given the following Core Reactivity Data during startup: Control Rod Total Rod Worth

(%k/k) Rod Worth at Critical

(%k/k) SHIM Rod 1 1.90 1.30

REG Rod 1.20 1.00

SHIM Rod 2 1.80 1.00

Transient Rod 2.60 1.20 After reactor shutdown, the operator plans to remove the SHIM Rod 1 out of the reactor core for an inspection. Which ONE of the following would be the Shutdown Margin specified in the NETL Tech Spec if the operator removes the SHIM Rod 1? Removing the SHIM Rod 1 out of the reactor core would:

a. violate the Tech Spec Shutdown Margin because the calculated result is 0.0 %k/k b. NOT violate the Tech Spec Shutdown Margin because the calculated result is 2.6 %k/k c. violate the Tech Spec Shutdown Margin because the calculated result is - 3.0 %k/k d. NOT violate the Tech Spec Shutdown Margin because the calculated result is 4.5 %k/k

Section A L Theory, Thermo, and Facility Characteristics Page 14 Question A.6 [1.0 point]

Which ONE of the following best describes the beta decay (-1) of a nuclide?

The atomic mass number unchanged, and the number of protons increases by 1.

The atomic mass number unchanged, and the number of protons decreases by 1.

The atomic mass number increases by 1, and the number of protons decrease by 1.

The atomic mass number increases by 2, and the number of protons increase by 1.

Question A.7 [1.0 point]

An example of a FISSIONABLE NUCLEI is:

a. Pu-239

b. U-238 c. U-235 d. U-233 Question A.8 [1.0 point]

The time period in which the MAXIMUM amount of Xe 135 will be present in the core is approximately 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after:

a. a startup to 100%power

b. a scram from 100% power

c. a power increase from 0% to 50%

d. a power decrease from 100% to 50%

Section A L Theory, Thermo, and Facility Characteristics Page 15 Question A.9 [1.0 point]

In a just critical reactor, adding $0.50 worth of reactivity in the SQUARE-WAVE MODE will cause:

a. A sudden drop in delayed neutrons

b. The reactor period to be equal to (-)/ c. A number of prompt neutrons equals to a number of delayed neutrons

d. The resultant period to be a function of the prompt neutron lifetime (T=*/) Question A.10 [1.0 point]

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. Higher moderator temperature (assume negative temperature coefficient).

b. Insertion of a negative reactivity worth experiment.

c. Burnout of a burnable poison.

d. Fuel depletion.

Question A.11 [1.0 point]

Which ONE of the following will be the resulting stable reactor period when a $0.35 reactivity insertion is made into an exactly critical reactor core? Neglect any effects from prompt. Given =0.0070 and =0.1

a. 13 seconds b. 19 seconds c. 28 seconds

d. 31 seconds

Section A L Theory, Thermo, and Facility Characteristics Page 16 Question A.12 [1.0 point]

You are the reactor operator performing two pulsing operations. The first pulse has a reactivity worth of $1.20 which results in a peak power of 200 MW. If the second pulse has a reactivity worth of $2.00, the corresponding peak power is:

Given: =0.0070

a. 1000 MW

b. 1750 MW

c. 2500 MW

d. 5000 MW

Question A.13 [1.0 point]

A reactor with K eff = 0.7 contributes 2000 neutrons in the first generation. Changing from the first generation to the SECOND generation, how many total neutrons are there after the second generation?

a. 2250 b. 2600 c. 3400 d. 4000

Section A L Theory, Thermo, and Facility Characteristics Page 17 Question A.14 [1.0 point]

Which ONE of the following physical characteristics of the TRIGA fuel is the main contributor for the prompt negative temperature coefficient?

a. As the fuel heats up the resonance absorption peaks broaden and increases the likelihood of absorption in U-238 and/or Pu-240

b. As the fuel heats up a rapid increase in moderator temperature occurs through conduction and convection heat transfer mechanisms which adds negative reactivity

c. As the fuel heats up fission product poisons (e.g., Xe) increase in concentration within the fuel matrix and add negative reactivity via neutron absorption

d. As the fuel heats up the oscillating hydrogen in the ZrH lattice imparts energy to a thermal neutron, thereby increasing its mean free path and probability of escape

Question A.15 [1.0 point]

Which ONE of the following describes the term PROMPT DROP

?

a. A reactor is subcritical at negative 80-second period.

b. A reactor has attained criticality on prompt neutrons alone.

c. The instantaneous change in power level due to inserting a control rod.

d. The instantaneous change in power level due to withdrawing a control rod.

Question A.16 [1.0 point]

About two minutes following a reactor scram, the reactor period has stabilized and the power level is decreasing at a CONSTANT rate. Given that reactor power at time t 0 is 90 kW power, what will it be five minutes later?

a. 0.2 kW

b. 2.1 kW

c. 3.4 kW

d. 117.5 kW

Section A L Theory, Thermo, and Facility Characteristics Page 18 Question A.17 [1.0 point]

The reactor is subcritical with the count rate of 100 counts per second (cps) and K eff of 0.950. The control rods are withdrawn until the count rate is doubled. What is the new value of K eff?

a. 0.952 b. 0.975 c. 0.998 d. 1.020 Question A.18 [1.0 point] Xenon-135 (Xe 135) is produced in the reactor by two methods. One is directly from fission; the other is indirectly from the decay of :

a. Xe 136 b. Sm 136 c. Cs 135 d. I 135 Question A.19 [1.0 point]

The reactor is on a CONSTANT positive period. Which ONE of the following power changes will take the longest time to complete?

a. 5%, from 95% to 100%

b. 10%, from 80% to 90%

c. 15%, from 15% to 30%

d. 20%, from 60% to 80%

Section A L Theory, Thermo, and Facility Characteristics Page 19 Question A.20 [1.0 point]

Delayed neutrons are produced by:

a. decay of O-16

b. Photoelectric Effect

c. decay of fission fragments

d. directly from the fission process

Section B Normal, Emergency and Radiological Control Procedures Page 20 Question B.1 [1.0 point] change was made during the exit meeting. Change will not affect the candidate's score An area in which radiation levels from radiation source to the body could result in an individual a dose equivalent of 500 mRem/hr at 30 cm is defined as:

a. Radiation area

b. Controlled Area

c. High Radiation Area.

d. Very High Radiation Area

Question B.2 [1.0 point, 0.25 each]

Match the terms in column A with their respective definitions in column B.

Column A Column B

a. Radioactivity 1. To remove a facility or site safely from service and reduce residual radioactivity to a level that permits in 10 CFR 52. b. Contamination 2. An impurity pollutes or adulterates another substance. The transferable radioactive materials are the sources of ionizing radiations.

c. Dose 3. The quantity of radiation absorbed per unit mass by the body or by any portion of the body.

d. Decommission 4. That property of a substance which causes it to emit ionizing radiation. This property is the spontaneous transmutation of the atoms of the substance.

Section B Normal, Emergency and Radiological Control Procedures Page 21 Question B.3 [1.0 point]

Which ONE of the following reactor Safety System Channels does NOT require its operation in the PULSE MODE?

a. Power Level Scram at 1.1 Mw

b. Fuel Temperature

c. Magnet Current

d. Loss of Watchdog timer Question B.4 [1.0 point]

Per FUEL-1 procedure, movement of CONTROL FOLLOWER requires a minimum shutdown margin:

a. greater than 0.2 %k/k with removal of the most reactive control rod.

b. lesser than 0.2 %k/k with removal of the most reactive control rod.

c. greater than 0.2 %k/k with removal of two most reactive control rods.

d. lesser than 0.2 %k/k with removal of two most reactive control rods.

Question B.5 [1.0 point]

In the event of an emergency that do not threaten the building systems, the emergency assembly area is:

a. The reactor control room

b. The health physics room

c. The emergency support center

d. The equipment access driveway

Section B Normal, Emergency and Radiological Control Procedures Page 22 Question B.6 [1.0 point]

Use the following diagram of an instrumented fuel element. Which ONE of the following is the correct match for the position locator (Column A) to the correct component (Column B)?

Column A Column B I A. Zirconium Hydride-Uranium II B. Stainless steel III C. Samarium Burnable Poison IV D. Graphite Reflector E. Zirconium Rod F. Spacer G. Thermocouples

a. I-C, II-F, III-A, IV-C

b. I-D, II-A, III-E, IV-C

c. I-D, II-E, III-A, IV-G

d. I-C, II-B, III-E, IV-G

Section B Normal, Emergency and Radiological Control Procedures Page 23 Question B.7 [1.0 point]

According to NETL Technical Specifications, which ONE of the following would most likely be considered a reportable occurrence?

a. You receive a bomb threat directed toward the facility.

b. You load an unknown sample to the core, which causes an unexplained change in a $1.20 worth of reactivity.

c. You observe an abnormal loss of core coolant at a rate that exceeds the normal makeup capacity.

d. You miss calculation of reactivity worth, cause reactor scram after performing a pulse.

Question B.8 [1.0 point]

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

a. The sum of thyroid dose and external dose.

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

c. The sum of the deep dose equivalent and the committed effective dose equivalent.

d. The dose that your whole body is received from the source, but excluded from the deep dose. Question B.9 [1.0 point]

Which ONE of the following would be an initiating condition for Non-reactor Specific Events?

a. Fuel cladding damage

b. Earthquake with damage to reactor systems

c. Discovery of forced entry directed to the reactor facility

d. Radioactive contamination of a reactor operator during removal of experiment

Section B Normal, Emergency and Radiological Control Procedures Page 24 Question B.10 [1.0 point]

What is the HALF LIFE of the isotope contained in a sample which produces the following count rates? Time (Minutes)

Counts per Minute (cpm)

Initial count 900 30 740 60 615 90 512 180 294 a. 551 minutes

b. 312 minutes

c. 111 minutes

d. 88 minutes Question B.11 [1.0 point]

During reactor operations involving a removal of pneumatic samples, the Reactor Operator (RO) becomes ill and is taken to the hospital. Only the Senior Reactor Operator (SRO) and an experienced student remain in the facility. Reactor operations:

a. must be discontinued because both an RO and an SRO must be in the facility to satisfy NETL Administrative Policy

b. must be discontinued because both an RO and an SRO must be in the facility to satisfy Technical Specifications

c. may continue until a replacement RO can arrive at the facility within 30 minutes

d. may continue since the SRO can monitor the console while the student can retrieve samples.

Section B Normal, Emergency and Radiological Control Procedures Page 25 Question B.12 [1.0 point]

According to NETL procedure, Class C experiments:

a. require the presence of the Reactor Supervisor

b. require the presence of the Senior Reactor Operator

c. require the presence of the Reactor Operator

d. do not require the licensed operator

Question B.13 [1.0 point]

The maximum operating power level for the operation of the NETL reactor shall be 1.1 MW in the manual, auto and square wave modes. This is an example of:

a. Safety Limit (SL)

b. Limiting Safety System Setting (LSSS)

c. Limiting Conditions for Operation (LCO)

d. Safety Operational Limit (SOL)

Question B.14 [1.0 point]

What is the MINIMUM level of management who may authorize temporary changes to the procedures that do NOT changes their original intent?

a. Reactor Operator

b. Senior Reactor Operator

c. Reactor Supervisor

d. The Reactor Operations Committee

Section B Normal, Emergency and Radiological Control Procedures Page 26 Question B.15 [1.0 point] Assume that there is no leak from outside of the demineralizer tank. You use a survey instrument with a window probe to measure the dose rate from the demineralizer tank.

Compare to the reading with a window CLOSED, the reading with a window OPEN will :

a. increase, because it can receive an additional alpha radiation from (Al-27) (n,), (Na-24) reaction.

b. remain the same, because the Quality Factors for gamma and beta radiation are the same. c. increase, because the Quality Factor for beta and alpha is greater than for gamma.

d. remain the same, because the survey instrument would not be detecting beta and alpha radiation from the tank.

Question B.16 [1.0 point,0.25 each] Identify each of the following surveillances as a channel check (CHECK), a channel test (TEST), or a channel calibration (CAL).

a. During performance of the Daily Checklist, you press a RED button on the NP-1000 to verify the reactor scram.

b. During 500 kW power, you compare the readings of Fuel Temperature 1 and Fuel Temperature 2.

c. Exposing a 2 mCi check source to the continuous air monitor (CAM) detector to verify that its output is operable.

d. Adjust the wide range linear channel in accordance with recent data collected on the reactor power calibration.

Section B Normal, Emergency and Radiological Control Procedures Page 27 Question B.17 [1.0 point]

A radioactive source reads 35 Rem/hr on contact. Five hours later, the same source reads 1.5 Rem/hr. How long is the time for the source to decay from a reading of 35 Rem/hr to 100 mRem/hr? a. 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

b. 7.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

c. 8.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

d. 9.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

Question B.18 [1.0 point]

Which ONE of the following documents requires the NRC approval for changes?

a. Major changes in the startup checklist

b Minor modification to the Technical Specifications c Rearrange chapters in the Safety Analysis Report d Revise the requalification operator licensing examination

Question B.19 [1.0 point]

How many hours per calendar quarter must you perform the functions of an RO or SRO to maintain an active RO or SRO license?

a. 2

b. 4 c. 6

d. 8

Section B Normal, Emergency and Radiological Control Procedures Page 28 Question B.20 [1.0 point]

The radiation from an unshielded Co-60 source is 500 mrem/hr. What thickness of lead shielding will be needed to lower the radiation level to 5 mrem/hr? The HVL (half-value-layer) for lead is 6.5 mm.

a. 26 mm.

b. 33 mm.

c. 38 mm.

d. 44 mm.

Section C Plant and Radiation Monitoring Systems Page 29 Question C.1 [1.0 point] Accept either "a" or "b" as the correct answers since the interlock signal comes from the NM-1000.

The safety interlock system will prevent air actuation if linear power is above 1kW and the PULSE mode is selected. This safety interlock signal comes from:

a. Wide-Range Log Channel (NM-1000)

b. Multi -Range Linear Channel (NM-1000)

c. High Flux Safety Channel (NPP-1000)

d. Control Rod limited switch Question C.2 [1.0 point] You are a reactor operator performing a Square Wave operation. In Square Wave mode, you press a FIRE button and verify the power ramp-up. If demand power is not reached in 10 seconds, the control system will automatically:

a. scram

b. exit to Auto mode

c. exit to Manual mode

d. stay at Square Wave mode Question C.3 [1.0 point] During a reactor startup, you calculate the maximum excess reactivity for reference core conditions of 5.00% k/k. For this excess reactivity, you will:

a. continue to operate because the excess reactivity is within TS limit.

b. increase power to 1 kW and verify the excess reactivity again.

c. shutdown the reactor; immediately report the result to the supervisor due to excess being above TS limit.

d Continue operation, but immediately report the result to the supervisor since the excess reactivity is exceeding the facility operation limit.

Section C Plant and Radiation Monitoring Systems Page 30 Question C.4 [1.0 point] Which ONE of the following best describes the thermocouples in each of the instrumented fuel elements (IFE)?

a. There are consisted of two Resistance Temperature Detectors (RTD) embedded at the midpoint and one inch above vertical center in the IFE.

b. There are consisted of three chromel-alumel thermocouples embedded at the midpoint, one inch above, and below vertical center in the IFE.

c. There are consisted of three Resistance Temperature Detectors (RTDs) embedded at the midpoint and one inch below vertical center in the IFE.

d. There are consisted of three platinum-rhodium thermocouples embedded at the midpoint, one inch above, and below vertical center in the IFE.

Question C.5 [1.0 point, 0.25 each] Match each of the safety channels in column A with the reactor modes of operations in column B. Items in column B can be used once, more than once or not at all.

Column A Column B

a. Fuel Temperature 1. Steady State ONLY

b. Multi-range Linear Power 2. Pulse ONLY

c. Watchdog Timer 3. Square wave ONLY

d. NPP-1000 for nvt measurement 4. All Modes

Section C Plant and Radiation Monitoring Systems Page 31 Q uestion C.6 [1.0 point]

Question deleted during the exit meeting. No correct answer in the distractors. This question will not factor into the candidates' grades Given the configuration of the LIGHTS associated with the SHIM rod/drive as follows: UP:OFF , DOWN: ON , MAGNET: OFF, and no failure of switch lights. Identify the conditions of the SHIM rod. a. Normal condition, rod insertion permissible

b. Abnormal condition, misadjusted rod down limit switch

c. Abnormal condition, rod has stuck above lower limit switch

d. Normal condition, either rod insertion or withdrawal permissible Question C.7 [1.0 point] Which ONE of the following controls the AMOUNT OF REACTIVITY that is inserted by the transient rod during pulse operations?

a. The preset pulse timer setting that vents the pneumatic piston

b. The steady state power of the reactor prior to firing the pulse

c. The pressure of the air applied to the pneumatic piston

d. The position of the cylinder

Question C.8 [1.0 point] Reactor is in a PULSE ready mode and you want to switch to a STEADY STATE mode. Which ONE of the following can cause the control rod interlock when you switch it?

a. SHIM rod drive DOWN and SHIM control rod DOWN

b. SHIM rod drive UP and SHIM control rod DOWN

c. Pneumatic cylinder DOWN

d. Pneumatic cylinder UP

Section C Plant and Radiation Monitoring Systems Page 32 Question C.9 [1.0 point]

An illuminated YELLOW light alarm on the ARM Monitors indicates:

a. calibration is required.

b. battery power has switched OFF.

c. the ARM may be failing or malfunctioning.

d. alert for radiation level in its immediate area.

Question C.10 [1.0 point]

During operation, you receive a message "SCRAM-DAC DIS64 Timeout". This message means that communication is disrupted between:

a. DAC Network and CSC Network

b. DAC Computer and DAC Digital Scanner board

c. DAC Computer and CSC Digital Scanner board

d. CSC Computer and DAC Digital Scanner board

Question C.11 [1.0 point, 0.25 each] Match the inputs listed in column A with their responses listed in column B. (Items in column B may be used more than once or not at all). Assume the reactor is in operation.

Column A Column B

a. Database Timeout 1. Indicate only

b. Magnet Power Grounded 2. Interlocks

c. Pool water conductivity = 1 micomho/cm 3. Alarm and scram

d. Withdrawal of Shim and Transient rods simultaneously in Steady State mode

Section C Plant and Radiation Monitoring Systems Page 33 Question C.12 [1.0 point] Which ONE of the following equipments does NOT require during the thermal power calibration?

a. Micro-volt meter

b. Thermocouple array

c. Micro-amp source

d. Ice bath reference junction for thermocouple array

Question C.13 [1.0 point] Significant quantities of Nitrogen-16 are produced by the irradiation of :

a. air in the beam ports

b. oxygen-16 in the reactor pool

c. air in irradiation cell

d. reactor building atmosphere

Question C.14 [1.0 point] Which ONE of the following best describes the design of fuel element used at NETL reactor?

a. The fuel is a mixture of U-Zn-H alloy containing 8.5% weight of uranium enriched to 19.7% U 235. b. The fuel is a mixture of U-Zn-H alloy containing 19.7% weight of uranium enriched to 8.5% U 235.

c. The fuel is a mixture of U-Zr-H alloy containing 19.7% weight of uranium enriched to 8.5% U 235.

d. The fuel is a mixture of U-Zr-H alloy containing 8.5% weight of uranium enriched to 19.7% U 235.

Section C Plant and Radiation Monitoring Systems Page 34 Question C.15 [1.0 point] The reactor operator is conducting the Thermal Power Calibration. Which ONE of the following is an initial setup before recording pool temperature?

a. Primary water system: ON, Secondary water system: ON, Power level: 500 kW

b. Primary water system: ON, Secondary water system: OFF, Power level: 900 KW

c. Primary water system: OFF, Secondary water system: ON, Power level: 500 kW

d. Primary water system: OFF, Secondary water system: OFF, Power level: 900 kW

Question C.16 [1.0 point]

Given the following sequence of events during the course of pulsing:

(1) The reactor is in steady state at 50 watts (2) Power is applied to the pulse integrator (3) The mode selector switch is in the pulse mode (4) A preset time sets 3 seconds Reactor operator initiates a pulse by pressing the fire button. Which ONE of the following sequences of events takes place?

a. The transient rod air will be energized and de-energized in 1 second. The transient rod will drop back into the core. The transient cylinder automatically drives down.

b. The transient rod air will be energized and de-energized in 3 seconds. The transient rod will drop back into the core. The transient cylinder automatically drives down.

c. The transient rod air will be energized and immediately de-energized. The transient rod will drop back into the core. The transient cylinder stays at the same location.

d. The transient rod air will be de-energized and then energized in 3 seconds. The transient rod and the cylinder stay at the same locations.

Section C Plant and Radiation Monitoring Systems Page 35 Question C.17 [1.0 point]

You are adding the makeup water to the reactor pool. In order to obtain volume of water added, you will multiply a pool level change (in cm) by:

a. 20.4 liters/cm

b. 32.6 liters/cm

c. 49.4 liters/cm

d. 56.4 liters/cm

Question C.18 [1.0 point]

During a reactor operation, you discover the particulate air monitor pump failure. Other monitors are operating. Which ONE of the following is the best action?

a. Continue to operate because the pump failure does NOT affect the operation of the particulate air monitor.

b. Continue to operate because the particulate air monitor may be out of service for a period of 1 week.

c. Shutdown the reactor; immediately report the result to the supervisor because the pump failure is a Tech Spec violation.

d Shutdown the reactor, immediately report the result to the U.S. NRC because it is a reportable occurrence.

Question C.19 [1.0 point]

In order to prevent radiation streaming through a beam port, each beam port contains:

a. a step (or steps) to provide for divergence of the radiation beam

b. an inner shield plug and outer shield plug

c. a lead-filled shutter and a lead-lined door

d. a removable cover plate

Section C Plant and Radiation Monitoring Systems Page 36 Question C.20 [1.0 point]

You perform a fuel element inspection. In measuring the elongation, you find the length of one fuel element exceeds the original length by 1.5875 mm (1/16 inches). For this measurement, you will:

a. continue the fuel inspection because this elongation is within TS limit.

b. continue the fuel inspection because the NETL Tech Spec requires the transverse bend measurement only.

c. stop the fuel inspection; immediately report the result to the supervisor because it is considered a damaged fuel element.

d stop the fuel inspection, immediately report the result to the supervisor since the length is exceeding the facility operation limit.

- - - - End of Section C ***************************** ******************* End of the Exam ***************************

Section A L Theory, Thermo, and Facility Characteristics Page 37 Answer Key A.1 Answer: b

Reference:

A.03 Answer: b

Reference:

Equation Sheet. = (*/) + [(-)/eff] A.5 Answer: a

Reference:

Shut Down Margin = rod worth at critical - SHIM rod 1 worth = (1.3+1.0+1.0+1.2) - 1.9 =4.5-1.9 = 2.6 %k/k Tech Spec SDM =SDM- most reactivity control rod worth (Transient rod) = 2.6 - 2.6 = 0.0 %k/k Violate the TS SDM : 0.0 %k/k < 0.2 %k/k by TS 3.1.2 A.6 Answer a

Reference:

Chart of the Nuclides

A.7 Answer: b

Reference:

Burn, R., Introduction to Nuclear Reactor Operations, 1988 Section 3.2 page 3-2

A.8 Answer b

Reference:

Standard NRC question.

A.11 Answer: b

=($)*; =$0.35*0.007= 0.00245 k/k; T = (-)/ T = (0.0070 - 0.00245)/0.1 x 0.00245 = 18.57 seconds

A.12 Answer: d

=($)*; 1=$1.20*0.007= 0.0084 k/k 2=$2.0*0.007= 0.014 k/k Peak2 = Peak1*(0.014-.007/.0084-.007) 2 = 5000 MW Or Peak2 = Peak1*($2-$1 / $1.2-$1) 2 = 5000 MW A.13 Answer: c

Reference:

TRIGA Fuel Design

A.15 Answer: c

Reference:

A.16 Answer: b

Reference:

P = P 0 e-T/ = 90 kW e(300sec/-80sec) = 900 kW e-3.75 = 0.0235 90 kW = 2.1 kW A.17 Answer: b

Reference:

Count 1*(1-K eff1) = Count 2*(1-K eff2) 100*(1-0.950) = 200*(1-K eff2) 100*(1-0.95) = 200(1- Keff2); K eff2 = 0.975 Peak)-( = Peak)-(1 1 2 2 2 2 Section A L Theory, Thermo, and Facility Characteristics Page 39 A.18 Answer: d

Reference:

§§ 8.1 -8.4, pp. 8 8-14.

A.19 Answer: c

Reference:

Time is related to ratio of final power to initial power. 2:1 is the largest ratio.

A.20 Answer: c

Reference:

Section B Normal, Emergency and Radiological Control Procedures Page 40 Answer Key B.1 Answer: c

Reference:

10 CFR 20

B.2 Answer: a(4) b(2) c(3) d(1)

Reference:

10 CFR 20 B.3 Answer: a

Reference:

TS 3.2.4

B.4 Answer: c

Reference:

SOP FUEL-1,Section V.b B.5 Answer: b

Reference:

Plan-E, Emergency Response, Section D.1.a B.6 Answer: c

Reference:

SAR, Figure 4-29 B.7 Answer: b

Reference:

TS 6.6.2 and TS 3.4.1

B.8 Answer: c

Reference:

10 CFR 20.

B.9 Answer: d

Reference:

Emergency Response Plan, Section 1.2.2 B.10 Answer: c

Reference:

A = A o e -t 294 = 900e

-180, 180 = -ln 0.327, = 0.00623 min

-1 t 1/2 = 0.693 / , = 0.693 / 0.00623 min

-1, = 111 minutes

B.11 Answer: d

Reference:

TS 6.1.3 Section B Normal, Emergency and Radiological Control Procedures Page 41 B.12 Answer: d

Reference:

ADMN-6 B.13 Answer: b

Reference:

Technical Specifications, Section 2.2.2 B.14 Answer: b

Reference:

Technical Specifications 6.3

B.15 Answer: d

Reference:

BASIC Radiological Concept (Betas and alpha don't make through the demineralizer tank)

B.16 Answer: a = TEST; b = CHECK; c = TEST; d = CAL

Reference:

NETL Technical specification § 1, Definitions B.17 Answer: d

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

B.18 Answer: b

Reference:

10 CFR 50.59 B.19 Answer: b

Reference:

10CFR55.53(e)

B.20 Answer: d

Reference:

DR = DR*e -X HVL ( =6.5 mm) means the original intensity will reduce by half when a lead sheet of 6.5 mm is inserted. Find if the HVL is given as follows: 1 = 2* e

-*6.5 ; = 0.10664 Find X: 5 mrem/hr = 500 mrem/hr* e -0.10664*X ; X= 43.2 mm

Section C Plant and Radiation Monitoring Systems Page 42 Answer Key C.1 Accept either "a" or "b" as the correct answers since the interlock signal comes from the NM-1000. Answer: a or b

Reference:

SAR 6.1.1 and walkthrough information C.2 Answer: c

Reference:

OPER-3, Section C

C.3 Answer: c

Reference:

TS 3.1.1 C.4 Answer: b

Reference:

NRC Standard Question C.5 Answer: a (4 ) b(1) c(4) d(2)

Reference:

SAR 6.1.1 through 6.1.5

C.6 Question deleted during the exit meeting. No correct answer in the distractors. This question will not factor into the candidates' grades Answer: b

Reference:

SAR 6.1.4 C.7 Answer: d

Reference:

NRC Standard Question C.8 Answer: d

Reference:

TS 3.2.2 and MAIN-1 C.9 Answer: d

Reference:

MAIN-4, Section B, Response Checks

C.10 Answer: b

Reference:

MAIN-1, Section 7.2 C.11 Answer: a(3) b(3) c(1) d(2)

Reference:

TS 3.3 and MAIN-1

C.12 Answer: c

Reference:

SURV-2,Section IV Section C Plant and Radiation Monitoring Systems Page 43 C.13 Answer: b

Reference:

NRC Standard Question C.14 Answer: d

Reference:

TS 5.3

C.15 Answer: d

Reference:

SURV-2,Section V C.16 Answer: b

Reference:

information during Walkthrough C.17 Answer: c

Reference:

SURV-4, Section A.1.c, page 3

C.18 Answer: b

Reference:

SAR 9.5.4

C.19 Answer: a

Reference:

C.20 Answer: a

Reference:

TS 3.1.4

- - - - END OF THE EXAM ********************************

ML12081A109

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