ML091420145

Initial Examination Report, No. 50-284/OL-09-01, Idaho State University AGN-201M Reactor

ML091420145
Person / Time
Site:	Idaho State University
Issue date:	06/03/2009
From:	Johnny Eads Research and Test Reactors Branch B
To:	Kunze J Idaho State University
Doyle P, NRC/NRR/DPR/PRT, 415-1058
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ML090370726	List: ML090370731 ML091420145
References
50-284/OL-09-01
Download: ML091420145 (17)

Text

June 3, 2009 Dr. Jay F. Kunze Reactor Administrator Idaho State University College of Engineering Campus Box 8060 Pocatello, ID 83209

SUBJECT:

INITIAL EXAMINATION REPORT NO. 50-284/OL-09-01, IDAHO STATE UNIVERSITY AGN-201M REACTOR

Dear Dr. Kunze:

During the week of April 20, 2009, the NRC administered operator licensing examinations at your Idaho State University AGN-201M Reactor. The examinations were 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. Paul Doyle at (301) 415-1058 or via internet e-mail Paul.Doyle@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. 50-284

Enclosures:

1. Initial Examination Report No. 50-284/OL-09-01

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

Please see next page

Dr. Jay F. Kunze June 3, 2009 Reactor Administrator Idaho State University College of Engineering Campus Box 8060 Pocatello, ID 83209

SUBJECT:

INITIAL EXAMINATION REPORT NO. 50-284/OL-09-01, IDAHO STATE UNIVERSITY AGN-201M REACTOR

Dear Dr. Kunze:

During the week of April 20, 2009, the NRC administered operator licensing examinations at your Idaho State University AGN-201M Reactor. The examinations were 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. Paul Doyle at (301) 415-1058 or via internet e-mail Paul.Doyle@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. 50-284

Enclosures:

1. Initial Examination Report No. 50-284/OL-09-01

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

Please see next page DISTRIBUTION w/ encls.:

PUBLIC PRTB r/f RidsNRRDPRPRTA RidsNRRDPRPRTB Facility File (CRevelle) O7 E13 ADAMS ACCESSION #: ML091420145 TEMPLATE #:NRR-074 OFFICE PRTB:CE IOLB:LA PRTB:SC NAME PDoyle:mxc CRevelle JEads DATE 05/22/2009 05/29/2009 06/03/2009 OFFICIAL RECORD COPY

Idaho State University Docket No. 50-284 cc:

Idaho State University ATTN: Mr. Kenyon Hart Reactor Supervisor Campus Box 8060 Pocatello, ID 83209-8060 Idaho State University ATTN: Dr. Richard T. Jacobsen College of Engineering Dean Campus Box 8060 Pocatello, ID 83209-8060 Idaho State University ATTN: Dr. Richard R. Brey Radiation Safety Officer Physics Department Box 8106 Pocatello, ID 83209-8106 Toni Hardesty, Director Idaho Dept. of Environmental Quality 1410 North Hilton Boise, ID 83606 Test, Research and Training Reactor Newsletter 202 Nuclear Sciences Center University of Florida Gainesville, FL 32611

U. S. NUCLEAR REGULATORY COMMISSION OPERATOR LICENSING INITIAL EXAMINATION REPORT REPORT NO.: 50-284/OL-09-01 FACILITY DOCKET NO.: 50-284 FACILITY LICENSE NO.: R-110 FACILITY: Idaho State University AGN-201M Reactor EXAMINATION DATES: April 20 - 23, 2009 SUBMITTED BY: ____________/RA/______________ 05/22/2009 Paul V. Doyle Jr., Chief Examiner Date

SUMMARY

During the week of April 20, 2009, the NRC administered examinations to 2 Reactor Operator license and 2 Senior Reactor Operator license candidates. All four of the candidates passed all portions of their respective examinations.

REPORT DETAILS

1. Examiners: Paul V. Doyle Jr., Chief Examiner, NRC

2. Results:

RO PASS/FAIL SRO PASS/FAIL TOTAL PASS/FAIL Written 2/0 2/0 4/0 Operating Tests 2/0 1/1 3/1 Overall 2/0 1/1 3/1

3. Exit Meeting:

Paul V. Doyle Jr., NRC, Chief Examiner George Immel, ISU, Nuclear Engineering Department Chair Jay Kunze, ISU, Reactor Administrator Kenyon Hart, ISU, Reactor Supervisor At the exit meeting the NRC examiner thanked the staff for their support in the administration of the examinations. The examiner did not note any generic weaknesses on the part of the license candidates.

ENCLOSURE 1

OPERATOR LICENSING EXAMINATION With Answer Key IDAHO STATE UNIVERSITY Week of April 20, 2009 ENCLOSURE 2

Section A Reactor Theory, Thermodynamics and Facility Operating Characteristics Page 2 QUESTION A.01 [2.0 points, 0.5 each]

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

Column A Column B

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

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

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

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

QUESTION A.02 [1.0 point]

135 Xenon-135 (Xe ) is produced in the reactor by two methods. One is directly from fission; the other is indirectly from the decay of:

136

a. Xe 136

b. Sm 135

c. Cs 135

d. I QUESTION A.03 [1.0 point]

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

a. The moderator-to-fuel ratio.

b. The physical dimensions of the core.

c. The strength of installed neutron sources.

d. The current time in core life.

QUESTION A.04 [1.0 point]

The neutron interaction in the reactor core that is MOST efficient in thermalizing fast neutrons occurs with the:

a. Hydrogen atoms in the polyethylene molecules

b. Carbon atoms in the polyethylene molecules

c. Uranium atoms in the fuel

d. Oxygen atoms in the fuel

Section A Reactor Theory, Thermodynamics and Facility Operating Characteristics Page 3 QUESTION A.05 [1.0 point]

The total amount of reactivity added by inserting or withdrawing a control rod from a reference height to any other rod height is called?

a. differential rod worth

b. shutdown reactivity

c. integral rod worth

d. reference reactivity QUESTION A.06 [1.0 point]

For most materials the neutron microscopic cross-section for absorption Fa generally

a. increases as neutron energy increases

b. decreases as neutron energy increases

c. increases as target nucleus mass increases

d. decreases as target nucleus mass increases QUESTION A.07 [2.0 points, 1/2 each]

Using the drawing of the Core Rod Position provided, identify each of the following reactivity worths.

a. Total Rod Worth 1. B - A

b. Actual Shutdown Margin 2. C - A

c. Technical Specification Shutdown Margin Limit 3. C - B

d. Excess Reactivity 4. D - C

5. E - C

6. E - D

7. E - A

Section A Reactor Theory, Thermodynamics and Facility Operating Characteristics Page 4 QUESTION A.08 [1.0 point]

Reactor power is rising on a 30 second period. Approximately how long will it take for power to double?

a. 35 seconds

b. 50 seconds

c. 70 seconds

d. 100 seconds

Section B Normal and Emergency Operating Procedures and Radiological Controls Page 5 QUESTION B.01 [2.0 points 0.5 each]

Identify each of the following statements as a Safety Limit (SL), a Limiting Safety System Setting (LSSS) or a Limiting Condition for Operation (LCO).

a. The core thermal fuse shall melt when heated to a temperature of about 120°C resulting in core separation and reactivity loss greater than 5% )k/k.

b. The shutdown margin with the most reactive safety or control rod fully inserted and the fine control rod fully inserted shall be at least 1% )k/k.

c. The maximum core temperature shall not exceed 200°C during either steady-state or transient operation.

d. The reactor room shall be considered a restricted area whenever the reactor is not secured.

QUESTION B.02 [2.0 points 0.5 each]

Identify whether each of the experiments listed below is Allowed (AL), required Double Encapsulation (DE), or is not allowed (NA) by technical specifications.

a. An experiment containing 22 grams of explosive material.

b. An experiment containing liquid fissionable material.

c. An experiment, calculated upon failure to release an approximate total dose equivalent of 0.02 mSv (1 mrem) for a period of two hours starting at the time, as a result of any airborne pathway.

d. An experiment containing a material corrosive to reactor components.

QUESTION B.03 [1.0 point] Question DELETED - Reference no longer exists.

Per Emergency Plan No. 6, Irradiation of Sample for Laboratory Analysis, which ONE of the following is the maximum dose that an experiment may read, upon removal from the reactor, and not require storage in the isotope storage area to allow for decay, prior to being released to an experimenter?

a. 0.5 mr

b. 1.0 mr

c. 5.0 mr

d. 10 mr

Section B Normal and Emergency Operating Procedures and Radiological Controls Page 6 QUESTION B.04 [2.0 points 0.5 each]

A licensed reactor operator (RO) and a certified observer (CO) are in the reactor room, with a Senior Reactor Operator (SRO) on call while the reactor is operating. The RO is required to leave due to a family emergency.

Identify whether each of the following scenarios is ALLOWED or NOT ALLOWED per technical specifications?

a. The CO takes over control of the reactor and the SRO remains on call.

b. The SRO comes to the control room and directs the actions of the CO who operates the reactor.

c. The SRO takes over operation of the reactor, and the CO remains in the control room.

d. The SRO takes over operation of the reactor, the CO may leave the control room.

QUESTION B.05 [1.0 point]

Given a 1 cm (0.394 inch) thick lead shield reduces the dose rate from an experiment by a factor of 2. A 10 cm (3.94 inch) thick shield will reduce the dose by a factor of approximately

a. 4

b. 20

c. 100

d. 1000 QUESTION B.06 [1.0 point]

Per the emergency plan the EMERGENCY PLANNING ZONE (EPZ) is

a. rooms 19 and 20.

b. rooms 20 and 23.

c. rooms 15, 16, 18, 19, 20, 22, 23 and 24

d. the entire Lillibridge Engineering Laboratory basement.

QUESTION B.07 [1.0 point]

The dose rate from a mixed beta-gamma point source is 100 mrem/hour at a distance of one (1) foot, and is 0.1 mrem/hour at a distance of twenty (20) feet. What percentage of the source consists of beta radiation?

a. 20%

b. 40%

c. 60%

d. 80%

Section B Normal and Emergency Operating Procedures and Radiological Controls Page 7 QUESTION B.08 [1.0 point]

Per Maintenance procedure 2 Procedure to Open the AGN-201 Core Tank, before starting the procedure, the reactor must have been shut down for a minimum of

a. 1 shift (8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />).

b. 1 day (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

c. 3 days (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />).

d. 1 week (128 hours0.00148 days <br />0.0356 hours <br />2.116402e-4 weeks <br />4.8704e-5 months <br />).

Section C Facility and Radiation Monitoring Systems Page 8 QUESTION C.01 [2.0 points, 0.4 each]

Identify each of the following systems as either ENERGIZED or DE-ENERGIZED after depressing the OFF button on the console.

a. Nuclear Instrumentation Channel #3

b. Fixed Radiation Monitor

c. Rod Position Instrumentation

d. Reactor Laboratory Ventilation

e. Control Rod Drives QUESTION C.02 [1.0 point]

What type of detector is used for the Low temperature switch?

a. A simple bi-metallic thermal switch

b. A precision platinum wound resistance temperature detector (RTD)

c. A chromel-alumel (Type K) thermocouple.

d. A copper-constantan (Type T) thermocouple QUESTION C.03 [2.0 points, 0.4 each]

Match the purpose in column A with the correct material from column B.

Column A Column B

a. fast neutron shield 1. Lead

b. reflector 2. Graphite

c. gamma-ray shield 3. Beryllium

d. moderator in core 4. Aluminum

e. moderator in fuse 5. Polyethylene

6. Polystyrene

7. Water

Section C Facility and Radiation Monitoring Systems Page 9 QUESTION C.04 [1.0 point]

Where would you go to de-energize the ventilation system during an emergency?

a. On the reactor room wall opposite room 15 (Reactor Supervisor Office)

b. On the corridor wall just outside the door to room 23 (Subcritical Assembly Laboratory).

c. On the corridor wall just outside the door to room 19 (Reactor Observation Room).

d. Just inside the door to room 22 (Counting Laboratory).

QUESTION C.05 [1.0 point]

Which ONE of the following is NOT an interlock preventing rod withdrawal insertion?

a. Both safety rods must be fully inserted prior to inserting the coarse control rod.

b. Both safety rods must be fully inserted prior to inserting the fine control rod.

c. The coarse control rod must be fully withdrawn prior to inserting the safety rods.

d. The fine control rod must be greater than or equal to half inserted prior to inserting the safety rods.

QUESTION C.06 [1.0 point]

What design feature insures that the thermal fuse melts before the rest of the reactor core? The thermal fuse

a. fuel has double the density of fuel pellets as the rest of the core.

235 238

b. fuel has double the ratio of U to U atoms as the rest of the core.

c. moderator has twice the density of polyethylene to aid in themalizing neutrons.

d. moderator has half the density of polyethylene to aid in themalizing neutrons.

QUESTION C.07 [1.0 point]

Which one of the following is the method used at Idaho State University to generate control rod position indication?

The signal is generated by

a. the output of a synchro-generator linked to the rod drive DC motor.

b. the change in voltage due to movement of a lead screw linked to the rod itself.

c. the changing current due to the closing of multiple magnetic reed switches located the entire length of the rod.

d. a direct output from the rod drive DC motor.

Section C Facility and Radiation Monitoring Systems Page 10 QUESTION C.08 [1.0 point]

Which ONE of the following is the type of detector for Nuclear Instrumentation Channel #1.

a. Argon filled Geiger-Mueller.

235

b. U lined Fission Chamber.

c. BF3 filled Proportional Counter.

d. BF3 filled Ion Chamber.

Section A Reactor Theory, Thermodynamics and Facility Operating Characteristics Page 11 A.01 a, 2; b, 4; c, 1 d, 3 REF:

A.02 d REF:

A.03 c REF:

A.04 a REF:

A.05 a REF:

A.06 b REF:

A.07 a, 7; b, 5; c, 6; d, 2 REF: Standard NRC Question A.08 c t/T REF: P = P0 e --> ln(2) = time ÷ 100 seconds -> time = ln (2) x 100 sec. 0.693 x 100 0.7 x 100 70 sec.

Section B Normal and Emergency Operating Procedures and Radiological Controls Page 12 B.01 a, LSSS; b, LCO; c, SL; d, LCO REF: Technical Specifications §§ 2.2, 3,1(b), 2.1 and 3.4.

B.02 a, NA; b, DE; c, AL; d, DE REF: Technical Specifications §§ 3.3(b), 3.3.(a), 3.3.c(1) and 3.3(b)

B.03 d QUESTION DELETED - Reference no longer exists REF: Experimental Plans No. 6, and 7; Special Safety Considerations Section.

B.04 a, NOT ALLOWED; b, ALLOWED; c, ALLOWED; d, NOT ALLOWED REF: Technical Specification 6.1.11 B.05 d 10 REF: 2 = 1024 . 1000 B.06 b REF: Emergency Plan § 2.8 B.07 c REF: 10CFR20. At 20 feet, there is no beta radiation. Gamma at 20 feet = 0.1 mrem/hour, gamma at 1 foot = 40 mrem/hour. Therefore beta at 1 foot = 60 mrem/hour = 60%.

B.08 b REF: Maintenance Procedure 2

Section C Facility and Radiation Monitoring Systems Controls Page 13 C.01 a, E; b, E; c, D; d, E e, D REF: Rewrite of EQB question, also Operating Procedure # 1 § VII, Shutdown, paragraph D.1.

C.02 a REF: ISU Safety Analysis Report (SAR) § 4.3.4, Interlock System 5th ¶.

C.03 a, 7; b, 2; c, 1; d, 5; e, 6; REF: ISU, Safety Analysis Report (SAR), § 4.2, Table 4.2-1 C.04 a REF: Emergency Plan, § 2.0 EMERGENCY PROCEDURES, Nuclear Emergency ¶ #3.

C.05 d REF: ISU SAR § 3.1 Control Rods C.06 a REF: ISU SAR § 5.10 Safety Devices, p. 103 C.07 a REF: NRC Examination Question Bank, also ISU SAR § 4.3.1 Figure 4.3-1, p. 54 C.08 c REF: ISU SAR § 4.3.2 Channel 1 description on p. 58.