ML071060015

Initial Examination Report No. 50-020/OL-07-01, Massachusetts Institute of Technology

ML071060015
Person / Time
Site:	MIT Nuclear Research Reactor
Issue date:	04/18/2007
From:	Johnny Eads NRC/NRR/ADRA/DPR/PRTB
To:	Bernard J Massachusetts Institute of Technology (MIT)
Eads J, NRR/ADRA/DPR/PRTB, 415-1471
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ML070220030	List: ML070220049 ML071060015 ML071060020
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50-020/OL-07-01
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Text

April 18, 2007 Dr. John Bernard, Director of Reactor Operations Nuclear Reactor Laboratory Massachusetts Institute of Technology 138 Albany Street Cambridge, MA 02139

SUBJECT:

INITIAL EXAMINATION REPORT NO. 50-20/OL-07-01, MASSACHUSETTS INSTITUTE OF TECHNOLOGY

Dear Dr. Bernard:

During the week of March 12, 2007, the NRC administered an operator licensing examination at the Massachusetts Institute of Technology reactor. The examination was conducted according to NUREG-1478, "Non-Power Reactor Operator Licensing Examiner Standards," Revision 1.

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 10 CFR 2.390 of the Commission's regulations, a copy of this letter and the enclosures will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the NRC Web site at (the Public Electronic Reading Room) 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 Patrick Isaac at 301-415-1019.

Sincerely,

/RA/

Johnny Eads, Chief Research and Test Reactors Branch B Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Docket No. 50-20

Enclosures:

1. Initial Examination Report No. 50-20/OL-07-01

2. Examination and answer key (RO/SRO) cc w/enclosures:

Please see next page

Massachusetts Institute of Technology Docket No. 50-20 cc:

City Manager City Hall Cambridge, MA 02139 Department of Environmental Protection One Winter Street Boston, MA 02108 Director, Radiation Control Program Department of Public Health 90 Washington Street Dorchester, MA 02121 Nuclear Preparedness Manager Massachusetts Emergency Management Agency 40 Worcester Road Framingham, MA 01702-5399 Test, Research, and Training Reactor Newsletter University of Florida 202 Nuclear Sciences Center Gainesville, FL 32611

Dr. John Bernard, Director April 18, 2007 of Reactor Operations Nuclear Reactor Laboratory Massachusetts Institute of Technology 138 Albany Street Cambridge, MA 02139

SUBJECT:

INITIAL EXAMINATION REPORT NO. 50-20/OL-07-01, MASSACHUSETTS INSTITUTE OF TECHNOLOGY

Dear Dr. Bernard:

During the week of March 12, 2007, the NRC administered an operator licensing examination at the Massachusetts Institute of Technology reactor. The examination was conducted according to NUREG-1478, "Non-Power Reactor Operator Licensing Examiner Standards," Revision 1.

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 10 CFR 2.390 of the Commission's regulations, a copy of this letter and the enclosures will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the NRC Web site at (the Public Electronic Reading Room) 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 Patrick Isaac at 301-415-1019.

Sincerely,

/RA/

Johnny Eads, Chief Research and Test Reactors Branch B Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Docket No. 50-20

Enclosures:

1. Initial Examination Report No. 50-20/OL-07-01

2. Examination and answer key (RO/SRO) cc w/enclosures:

Please see next page DISTRIBUTION:

PUBLIC PRTB r/f JEads Facility File EBarnhill (O6-F2)

ADAMS ACCESSION #: ML071060015 TEMPLATE #: NRR-074 PACKAGE ACCESSION #: ML070220030 OFFICE PRTB:CE IOLB:LA PRTB:BC NAME PIsaac:cah EBarnhill JEads DATE 04/17/2007 04/17/2007 04/18/2007 OFFICIAL RECORD COPY

U. S. NUCLEAR REGULATORY COMMISSION OPERATOR LICENSING INITIAL EXAMINATION REPORT REPORT NO.: 50-20/OL-07-01 FACILITY DOCKET NO.: 50-20 FACILITY LICENSE NO.: R-37 FACILITY: Massachusetts Institute of Technology EXAMINATION DATES: 03/12/2007 - 03/14/2007 EXAMINER: Patrick Isaac, Chief Examiner SUBMITTED BY: /RA/ 04/05/2007 Patrick Isaac, Chief Examiner Date

SUMMARY

During the week of March 12, 2007, the NRC administered Operator Licensing Examinations to two Senior Reactor Operator Instant (SROI) and two Reactor Operator (RO) candidates. One SROI candidate failed the operating tests. All the other candidates passed their respective portions of the examinations.

ENCLOSURE 1

REPORT DETAILS

1. Examiner:

Patrick Isaac, Chief Examiner

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:

Personnel attending:

Edward S. Lau, Asst. Superintendent for Reactor Operations, NRL Frank Warmsley, Operations and Training Coordinator, NRL Patrick Isaac, NRC, Chief Examiner The Chief Examiner agreed to make the following changes to the written examination:

Question C.8 - Accept both "a" and "d" as correct.

Question C.13 - Delete the question (No longer true due to recent modifications to the cooling tower).

U. S. NUCLEAR REGULATORY COMMISSION NON-POWER REACTOR INITIAL LICENSE EXAMINATION FACILITY: MIT REACTOR TYPE: MITR DATE ADMINISTERED: 3/12/2007 CANDIDATE:

INSTRUCTIONS TO CANDIDATE:

Answers are to be written on the answer sheet provided. Attach the answer sheets to the examination. Points for each question are indicated in paren-theses for each question. A 70%

overall is required to pass the examination. Examinations will be picked up three (3) hours after the examination starts.

% OF CATEGORY  % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 20.00 33.3 A. REACTOR THEORY, THERMODYNAMICS AND FACILITY OPERATING CHARACTERISTICS 20.00 33.3 B. NORMAL AND EMERGENCY OPERATING PROCEDURES AND RADIOLOGICAL CONTROLS 20.00 33.3 C. PLANT AND RADIATION MONITORING SYSTEMS FINAL GRADE

% TOTALS ALL THE WORK DONE ON THIS EXAMINATION IS MY OWN. I HAVE NEITHER GIVEN NOR RECEIVED AID.

CANDIDATE'S SIGNATURE ENCLOSURE 2

Section A: L Theory, Thermodynamics & Facility Operating Characteristics Page 2 ANSWER SHEET Multiple Choice (Circle or X your choice)

If you change your answer, write your selection in the blank.

MULTIPLE CHOICE 001 a b c d 002 a b c d 003 a b c d 004 a b c d 005 a b c d 006 a b c d 007 a b c d 008 a b c d 009 a b c d 010 a b c d 011 a b c d 012 a b c d 013 a b c d 014 a b c d 015 a b c d 016 a b c d 017 a b c d 018 a b c d 019 a b c d 020 a b c d

(***** END OF CATEGORY A *****)

Section B Normal, Emergency and Radiological Control Procedures Page 3 ANSWER SHEET Multiple Choice (Circle or X your choice)

If you change your answer, write your selection in the blank.

MULTIPLE CHOICE 001 a b c d 002 a b c d 003 a b c d 004 a b c d 005 a b c d 006 a b c d 007 a b c d 008 a b c d 009 a b c d 010 a b c d 011 a b c d 012 a b c d 013 a b c d 014 a b c d 015 a b c d 016 a b c d 017 a b c d 018 a b c d 019 a b c d

(***** END OF CATEGORY B *****)

Section C Facility and Radiation Monitoring Systems Page 4 ANSWER SHEET Multiple Choice (Circle or X your choice)

If you change your answer, write your selection in the blank.

MULTIPLE CHOICE 001 a b c d 002 a b c d 003 a b c d 004 a b c d 005 a b c d 006 a b c d 007 a b c d 008 a b c d 009 a b c d 010 a b c d 011 a b c d 012 a b c d 013 a b c d 014 a b c d 015 a b c d 016 a b c d 017 a b c d 018 a b c d 019 a b c d 020 a b c d

(********** END OF EXAMINATION **********)

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 not received or 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.

6. Fill in the date on the cover sheet of the examination (if necessary).

7. Print your name in the upper right-hand corner of the first page of each section of your answer sheets.

8. The point value for each question is indicated in parentheses after the question.

9. Partial credit will NOT be given.

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

11. When you are done and have turned in your examination, leave the examination area as defined by the examiner.

EQUATION SHEET C C C C Q = m cp T = Q = m h C

Q = UA T SCR = S/(1-Keff)

CR1 (1-Keff)1 = CR2 (1-Keff)2 26.06 (eff) (1-Keff)0 SUR =))))))))))))) M = ))))))))))

( - ) (1-Keff)1 SUR = 26.06/ M = 1/(1-Keff) = CR1/CR0 P = P0 10SUR(t) SDM = (1-Keff)/Keff C

P = P0 e(t/) Pwr = W f m (1-)

P = )))))))) Po R* = 1 x 10-5 seconds

= (R*/) + [(-)/eff] = R*/(-)

= (Keff-1)/Keff eff = 0.1 seconds-1

= Keff/Keff 0.693 T1/2 = ))))))

DR1D12 = DR2D22 DR = DRoe-t 6CiE(n)

DR = ))))))))

R2 1 Curie = 3.7x1010 dps 1 kg = 2.21 lbm 1 hp = 2.54x103 BTU/hr 1 Mw = 3.41x106 BTU/hr 1 BTU = 778 ft-lbf EF = 9/5EC + 32 1 gal H2O . 8 lbm EC = 5/9 (EF - 32)

Section A: L Theory, Thermodynamics & Facility Operating Characteristics Page 7 QUESTION A.1 [1.0 point]

Core excess reactivity (ex) changes with

a. fuel element burnup

b. control rod height

c. neutron energy level

d. reactor power level QUESTION A.2 [1.0 point]

Which ONE of the following is the definition of the term Cross-Section?

a. The probability that a neutron will be captured by a nucleus.

b. The most likely energy at which a charge particle will be captured.

c. The length a charged particle travels past the nucleus before being captured.

d. The area of the nucleus including the electron cloud.

QUESTION A.3 [1.0 point]

The method for determining Calculated Thermal Power is described by:

a. Primary power plus reflector power plus shield power.

b. Primary power plus reflector power minus shield power.

c. Primary power plus shield power minus reflector power.

d. Primary power minus reflector power minus shield power.

Section A: L Theory, Thermodynamics & Facility Operating Characteristics Page 8 QUESTION A.4 [1.0 point]

As primary coolant temperature increases, control rod worth:

a. decreases due to lower reflector efficiency.

b. decreases due to higher neutron absorption in the moderator.

c. increases due to the increase in thermal diffusion length.

d. remains the same due to constant poison cross-section of the control rods..

QUESTION A.5 [1.0 point]

The term reactivity may be described as

a. a measure of the cores fuel depletion.

b. negative when Keff is greater than 1.0.

c. a measure of the cores deviation from criticality.

d. equal to $.50 when the reactor is prompt critical.

QUESTION A.6 [1.0 point]

The table provided lists data taken during a core loading. Estimate the number of fuel elements needed to go critical.

a. 24 Count Rate Number for Fuel Elements

b. 27 842 2

c. 30 886 7 1052 12

d. 38 1296 17 4210 22

Section A: L Theory, Thermodynamics & Facility Operating Characteristics Page 9 QUESTION A.7 [1.0 point]

The peak differential reactivity worth of the shim blades occur at the midpoints of their calibration curves, but the peak differential reactivity worth of the regulating rod occurs at the bottom of its calibration curve because:

a. the regulating rod is made of cadmium and the shim blades of boron-impregnated stainless steel.

b. the regulating rod is shadowed by the shim bank when it is above the bank height.

c. the regulating rod is a cylinder, while the blades are paddle-shaped.

d. the full-in' position of the regulating rod is six inches above that of the shim blades.

QUESTION A.8 [1.0 point]

The Doppler effect is described by the:

a. blue glow seen around the fuel elements.

b. broadening of resonance peaks as fuel temperature increases.

c. slower response time in comparison to moderator temperature feedback.

d. spectral shift of Uranium-235 as moderator temperature increases.

QUESTION A.9 [1.0 point]

The Fast Fission Factor () is defined as The ratio of the number of neutrons produced by

a. fast fission to the number produced by thermal fission.

b. thermal fission to the number produced by fast fission.

c. fast and thermal fission to the number produced by thermal fission.

d. fast fission to the number produced by fast and thermal fission.

Section A: L Theory, Thermodynamics & Facility Operating Characteristics Page 10 QUESTION A.10 [1.0 point]

Given the data in the table to the right, which ONE of the following is the closest to the half-life of the material?

TIME ACTIVITY

a. 11 minutes 0 minutes 2400 cps

b. 22 minutes 10 minutes 1757 cps 20 minutes 1286 cps

c. 44 minutes 30 minutes 941 cps

d. 51 minutes 60 minutes 369 cps QUESTION A.11 [1.0 point]

For a doubling time of 25 seconds the corresponding reactor period is:

a. 25 seconds

b. 36 seconds

c. 50 seconds

d. 81 seconds QUESTION A.12 [1.0 point]

A reactor operator understands that:

a. The more neutrons multiply during startup the lower the shim blades are at critical.

b. There is no fixed relationship between neutron level and criticality.

c. Neutron multiplication during startup is just neutrons getting lost at a slower rate.

d. Without the Sb-Be source the reactor would not go critical.

Section A: L Theory, Thermodynamics & Facility Operating Characteristics Page 11 QUESTION A.13 [1.0 point]

In a subcritical reactor, K eff is increased from 0.861 to 0.946. Which ONE of the following is the amount of reactivity that was added to the reactor core?

a. 0.085 delta k/k

b. 0.104 delta k/k

c. 0.161 delta k/k

d. 0.218 delta k/k QUESTION A.14 [1.0 point]

A fissile material is one which will fission upon the absorption of a THERMAL neutron. Which ONE of following listed isotopes is not a fissile material?

a. Th232

b. U233

c. U235

d. Pu239 QUESTION A.15 [1.0 point]

As criticality is approached during a startup, blade withdrawal is made in smaller and smaller increments and with longer and longer intervals between withdrawals because:

a. as the reactor approaches criticality, the number of neutron generations required to attain equilibrium increases.

b. gamma rays begin to swamp the neutron signal in the nuclear detectors.

c. the fraction of delayed neutrons appraches zero as criticality is approached.

d. subcritical multiplication becomes less important.

Section A: L Theory, Thermodynamics & Facility Operating Characteristics Page 12 QUESTION A.16 [1.0 point]

Delayed neutrons are essential to reactor safety because:

a. more delayed neutrons are produced than prompt neutrons.

b. delayed neutrons take longer to thermalize than do prompt neutrons, resulting in longer reactor periods during power changes.

c. delayed neutrons are born at different energies than prompt ones, thereby ensuring a broad neutron energy distribution.

d. delayed neutrons increase the average neutron lifetime, resulting in a longer reactor period during power changes.

QUESTION A.17 [1.0 point]

Coolant flows through a reactor core at a rate of 50 GPM, resulting in a coolant temperature increase of 6 degrees F. The power of the reactor is:

a. 5.3 kW.

b. 14.7 kW.

c. 44.0 kW.

d. 329.1 kW.

QUESTION A.18 [1.0 point]

Which ONE of the following describes the behavior of xenon?

a. Xenon is produced from both fission and the decay of iodine.

b. Xenon peaks upon shutdown and remains at the peak until power operation resumes.

c. Xenon reactivity worth varies linearly with reactor power.

d. Equilibrium xenon is worth 1 Beta.

Section A: L Theory, Thermodynamics & Facility Operating Characteristics Page 13 QUESTION A.19 [1.0 point]

Identify the PRINCIPAL source of heat in the reactor after shutdown?

a. Stored energy from the reactor and core materials

b. Spontaneous fission within the core

c. Decay of fission products

d. Cosmic radiation causing fission QUESTION A.20 [1.0 point]

Which ONE of the following is the reason why it requires 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of constant power operation before thermal equilibrium is attained in the MITR-II reactor?

a. The time required for equilibrium Xenon and Samarium conditions to be established.

b. The time required for the large volume of the Deuterium tank to heat up.

c. The shield coolant system has a small flowrate to accomplish adequate mixing before temperature is uniformly stabilized.

d. The graphite reflector has a large heat capacity and is slow to reach equilibrium temperature distribution.

Section B Normal, Emergency and Radiological Control Procedures Page 14 QUESTION B.1 [1.0 point]

An accessible area within the facility has general radiation levels of 325 mrem/hour. What would be the EXPECTED posting for this area?

a. "Caution, Very High Radiation Area"

b. "Danger, Airborne Radioactivity Area"

c. "Danger, High Radiation Area"

d. "Caution, Radiation Area" QUESTION B.2 [1.0 point]

While working on an experiment, you receive the following radiation doses: 100 mrem (),

25 mrem (), and 5 mrem (thermal neutrons). Which ONE of the following is your total dose?

a. 175 mrem

b. 155 mrem

c. 145 mrem

d. 130 mrem QUESTION B.3 [2.0 points, 1/2 each]

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

a. Gamma 1. Stopped by thin sheet of paper

b. Beta 2. Stopped by thin sheet of metal

c. Alpha 3. Best shielded by light material

d. Neutron 4. Best shielded by dense material

Section B Normal, Emergency and Radiological Control Procedures Page 15 QUESTION B.4 [1.0 point]

10CFR50.54(x) states: A licensee may take reasonable action that departs from a license condition or a technical specification (contained in a license issued under this part) in an emergency when this action is immediately needed to protect the public health and safety and no action consistent with license conditions and technical specifications that can provide adequate or equivalent protection is immediately apparent. Per 10CFR50.54(y), which one of the following is the minimum level of authorization for this action?

a. Reactor Operator licensed at the facility.

b. Senior Reactor Operator licensed at the facility.

c. Facility Manager (or equivalent at facility).

d. The U.S. Nuclear Regulatory Commission Project Manager QUESTION B.5 [1.0 point]

Which ONE of the following statements specifies a condition which satisfies Technical Specification Shutdown Margin requirements?

a. The reflector dump time must be at least twice the initial measured value.

b. No less than five shim blades are operable and the inoperable blade is at the operating position or higher.

c. With the most reactive blade and regulating rod fully withdrawn the reactor can be made at least 1% deltaK/K subcritical from the cold Xenon equilibrium critical condition.

d. Variable reactivity effects (samples) shall be in their most negative reactive state.

QUESTION B.6 [1.0 point]

As permitted by 10 CFR 50.59, the MITR may:

a. Modify systems and change the Technical Specifications (TS) if the NRC is notified afterwards.

b. Perform new and little understood experiments when they are for research.

c. Determine the affects of modifications and their impact on TS.

d. Redefine the boundaries of accidents previously analyzed in the Safety Analysis Report (SAR).

Section B Normal, Emergency and Radiological Control Procedures Page 16 QUESTION B.7 [1.0 point]

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

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

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

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

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

QUESTION B.8 [1.0 point]

Containment integrity is required whenever:

a. the H2 concentration in the air space above the core exceeds 1.0 volume percent.

b. maintenance is being performed on the rod control system.

c. the reactor is not secured.

d. the emergency cooling system is not operable.

QUESTION B.9 [1.0 point]

A small radioactive source is to be stored in an accessible area of the reactor building. The source reads 2 R/hr at 1 foot. Assuming no shielding is to be used, a Radiation Area barrier would have to be erected from the source at least a distance of approximately:

a. 400 feet

b. 40 feet

c. 20 feet

d. 10 feet

Section B Normal, Emergency and Radiological Control Procedures Page 17 QUESTION B.10 [1.0 point]

In the event of an On-Site Evacuation, personnel should be directed to proceed to the:

a. Campus Police Headquarters (Bldg. W31)

b. Operations Office (NW12-116).

c. NW13 Receiving Room.

d. NW13 Machine Shop.

QUESTION B.11 [1.0 point]

Safety Limits are

a. limits on variables associated with core thermal and hydraulic performance which are established to protect the integrity of the fuel clad.

b. settings for automatic protective devices related to those variables having significant safety functions.

c. settings for ANSI 15.8 suggested reactor scrams and/or alarms which form the protective system for the reactor or provide information which requires manual protective action to be initiated.

d. the lowest functional capability or performance levels of equipment required for safe operation of the reactor.

QUESTION B.12 [1.0 point]

What type of radiation hazard is associated with D2O?

a. Alpha

b. Beta

c. Gamma

d. Neutron

Section B Normal, Emergency and Radiological Control Procedures Page 18 QUESTION B.13 [1.0 point]

What is the maximum power level allowed if the reactor top shield is NOT in place?

a. 100 W

b. 500 W

c. 100 kW

d. 250 kW QUESTION B.14 [1.0 point]

The maximum reactivity worth of a single non-secured experiment allowed by Technical Specifications is:

a. 0.2 % delta K/K

b. 0.5 % delta K/K

c. 1.0 % delta K/K

d. 1.8 % delta K/K QUESTION B.15 [1.0 point]

You have not performed the functions of an RO or SRO in the past 6 months. Per the Regulations, prior to resuming activities authorized by your license, how many hours must you complete in that function under the direction of an RO or SRO as appropriate?

a. 4

b. 6

c. 12

d. 40

Section B Normal, Emergency and Radiological Control Procedures Page 19 QUESTION B.16 [1.0 point]

A reactor startup is declared "non-routine" if:

a. a thermal power calibration has not been performed.

b. a significant shift has occurred in the radial power distribution.

c. the refueling sequence involved fuel flipping.

d. the startup is to be performed when xenon is peaking.

QUESTION B.17 [1.0 point]

While responding to a safety system scram alarm, you are required to perform a MAJOR scram of the reactor if:

a. any building area monitor is above the alarm point.

b. a low level in the primary core tank is verifed.

c. any plenum monitor is above the alarm point.

d. a safety limit was exceeded.

QUESTION B.18 [1.0 point]

During a normal reactor startup the reactor attains criticality before the shim bank height reaches the 0.5" below ECP position. The required action is to:

a. limit power to 100 kW until the cause has been determined.

b. establish/maintain an infinite period and recalculate the ECP.

c. lower the shim bank at least 1.0" and determine the cause.

d. drive all rods in and recalculate the ECP.

Section B Normal, Emergency and Radiological Control Procedures Page 20 QUESTION B.19 [1.0 point]

Shim blade withdrawal motion is limited to four inches by the "subcritical position" interlock circuit. Which ONE of the following is NOT a reason for incorporating the subcritical interlock into the shim blade circuit?

a. To maintain the shim blade bank at a uniform height during final approach to criticality.

b. To establish a level, below the critical position, to which the shim blades may be individually withdrawn in one step.

c. To provide a convenient reference point at which the operator can pause to make a complete instrument check before bringing the reactor to criticality.

d. To maintain the shim blade bank at a uniform height sufficient to maintain subcritical multiplication on the startup channels.

Section C Facility and Radiation Monitoring Systems Page 21 QUESTION C.1 [1.0 point]

At what building pressure should the containment pressure relief system be placed on line?

a. 3 psi

b. 2 psi

c. 2.5 psi

d. 1 psi QUESTION C.2 [1.0 point]

Which ONE of the following alarm conditions will result in an automatic scram?

a. High Pressure Reactor Inlet.

b. High Level Emergency Power Channel.

c. Low Level Dump Tank.

d. Low Flow Shield Coolant.

QUESTION C.3 [1.0 point]

A major concern when responding to any casualty that affects the reflector is:

a. possible tritium exposure.

b. toxicity of the heavy water.

c. corrosiveness of irradiated heavy water.

d. possible N-16 exposure.

Section C Facility and Radiation Monitoring Systems Page 22 QUESTION C.4 [1.0 point]

Which ONE of the following describes decay heat removal capability while on Emergency Power?

a. Primary coolant system auxiliary pump MM2 can be restarted after resetting the low-voltage protection.

b. Primary coolant system pump MM1 can be restarted after resetting the low-voltage protection.

c. Standby Transfer Pump DM-2 will automatically start on high temperature.

d. Natural circulation provides cooling since pumping power is not available.

QUESTION C.5 [1.0 point]

Which ONE of the following automatic actions will occur, if the on-line waste tank reaches a high level alarm condition?

a. City water inlet solenoid valve closes.

b. Waste tank vent valve closes.

c. On-line sewer pump trips.

d. Sump pumps trip.

QUESTION C.6 [1.0 point]

What action should always be taken to maximize effectiveness of the emergency plan for airborne releases?

a. Shut down the reactor; Isolate containment.

b. Shut down the reactor; Leave ventilation on.

c. Lower shim blades to subcritical; Isolate containment.

d. Minor scram..

Section C Facility and Radiation Monitoring Systems Page 23 QUESTION C.7 [1.0 point]

The reactor is operating in automatic control at 80% power when the "High Pressure Reactor Inlet" alarm annunciates. Which one of the following changes, if occurring simultaneously, would NOT require the reactor to be scrammed?

a. Reactor period is slightly negative. Regulating rod moving out.

b. Core T higher that normal.

c. MPS-3A (Heat exchanger outlet pressure) reads high.

d. Core purge flow reads 5 cfm.

QUESTION C.8 [1.0 point]

Which ONE of the following statements describes the operation of the ventilation system?

a. If the main intake damper fails to close within ten seconds of a trip signal, the auxiliary damper will close.

b. If temperature of the air leaving the preheat coils drops below 35EF, the main intake damper will close but the intake fan continues to run.

c. If the main intake damper fails to close, it can be operated manually by using a lanyard inside containment near the damper.

d. In the "weekend-open" position, if activity is detected the plenum monitors will trip the inlet dampers and intake fan.

QUESTION C.9 [1.0 point]

If the reactor hold-down grid latch is opened, then the interlock associated with the grid latch will:

a. trip the primary pumps.

b. dump the reflector.

c. trip the D2O pumps.

d. initiate emergency core cooling.

Section C Facility and Radiation Monitoring Systems Page 24 QUESTION C.10 [1.0 point]

Which ONE of the following indications is not indicative of a fission product release from a fuel element?

a. Increasing readings on core purge monitor.

b. Increasing readings on plenum air monitor.

c. Increasing readings on N-16 monitor.

d. Increasing readings on NW12 gamma monitor.

QUESTION C.11 [1.0 point]

The operator in the control room would be required to notify personnel on the reactor top, if during refueling:

a. a positive steady-state period is observed when no fuel or dummies are being moved.

b. subcritical multiplication levels decrease by a factor of 2 or more.

c. a negative steady-state period is observed while fuel or dummies are being moved.

d. radiation levels decrease by a factor of 2 or more.

QUESTION C.12 [1.0 point]

Which of the following indications will automatically actuate outside of the control room when the 'trouble NW-12 gamma monitor' scam alarms.

a. A red light and a warning horn in operations office.

b. A blue light and a bell at the reception area.

c. A siren and backlighted signs in the containment building.

d. A horn in building NW12 and backlighted signs at entrances.

Section C Facility and Radiation Monitoring Systems Page 25 QUESTION C.13 [1.0 point] DELETED A rapid shift of cooling tower flow from the basins to spray may cause a reactor scram by causing a:

a. a temporary reduction in secondary flow.

b. pressure pulse that is transmitted via the heat exchanger to the primary, where it appears as a loss of flow on MPS-6/6A.

c. rapid cooldown of the primary and secondary systems.

d. temporary lack of heat removal until spray becomes effective.

QUESTION C.14 [1.0 point]

Which ONE of the following conditions will result in a Spent Fuel Storage Pool (SFSP) Alarm?

a. Leak

b. SFSP Pump trip on thermal overload

c. Securing pump from the SFSP control panel

d. High temperature ion column inlet flow QUESTION C.15 [1.0 point]

Which ONE of the following describes the design feature that provides for containment overpressure protection?

a. A pressure relief blower automatically starts.

b. A containment relief valve automatically opens.

c. A containment relief valve can be manually opened.

d. The stack exhaust control damper will automatically open on differential pressure.

Section C Facility and Radiation Monitoring Systems Page 26 QUESTION C.16 [1.0 point]

The reason for limiting cleanup system temperature to less than 50 degrees-C is to:

a. minimize gaseous release when sampling the primary.

b. prevent damage to the mixed-bed ion exchanger resin.

c. prevent coolant flashing when passing through filters.

d. limit the inaccuracy of the conductance probes, which are not temperature compensated.

QUESTION C.17 [1.0 point]

The automatic action associated with the Sewer radiation monitor high alarm during NORMAL system operation is:

a. the Sump pump trips.

b. the Sewer pump trips.

c. the inlet City Water solenoid valve closes.

d. the isolation valve closes to secure flow to the sewer.

QUESTION C.18 [1.0 point]

Which ONE of the following actions should the console operator perform immediately, if during a 1PH1 ] NW13 rabbit irradiation, the rabbit station radiation monitor alarms?

a. Eject the sample into the hot cell using the "Abort Auto Transfer" pushbutton.

b. Shutdown the reactor and when radiation levels are less than the permissible limit, use the 1PH1 "Eject" pushbutton to remove the sample.

c. Commence a normal reactor shutdown and dump the reflector.

d. Depress and hold the radiation alarm reset pushbutton to allow for the automatic transfer of the sample.

Section C Facility and Radiation Monitoring Systems Page 27 QUESTION C.19 [1.0 point]

Which ONE of the following is the reason for shutting down the reactor if the compressed air system pressure is less than 40 psig?

a. Personnel airlock cannot be operated.

b. Eventual loss of containment integrity.

c. Loss of ability to dump the reflector.

d. Prevent trip on low secondary flow indication.

QUESTION C.20 [1.0 point]

The primary concern associated with the pressure relief system charcoal filters becoming submersed during a large leak of primary coolant is:

a. loss of efficiency in removing particulates.

b. possible spontaneous combustion during dryout.

c. reduction in relief flow capability to relieve pressure.

d. possible spread of contamination from leaks in the filter housing.

Answer Key A.1 a REF: Reactor Training Manual - Core Excess and Shutdown Margin.

A.2 a REF: Reactor Training Manual - Cross Section.

A.3 a REF: PM 2.4, pg. 5 A.4. c REF: Reactor Physics Notes, Reactivity Feedback, Sect. 5 A.5 c REF: Reactor Training Manual - Reactivity A.6 a REF: Standard NRC question A.7 d REF: RSM 1.6.2 A.8 b REF: Reactor Physics Notes, Reactivity Feedback A.9 c REF: Reactor Training Manual - Neutron Life Cycle A.10 b REF: Reactor Training Manual - Reactivity A.11 b REF: Reactor Physics Notes, Reactivity Kinetics, pg. 18 Doubling Time = 0.693 x Period = 25 secs, Period = 36 secs.

A.12 b REF: Glasstone, 1958, CHAP 14 A.13 b REF: Reactor Physics Notes, Reactivity Kinetics, pg. 2 1 = (0.861 - 1)/0.861 = -0.161k/k; 2 = (0.946 - 1)/0.946 = -0.057k/k

= 2 - 1 = -0.057 - (-0.161) = +0.104 delta k/k A.14 a REF: Glasstone and Sesonske, Third Ed. § 1.45 A.15 a REF: MIT Reactor Physics Notes on Subcritical Multiplication A.16 d REF: Reactor Physics Notes, Reactivity Kinetics, pg. 8-10

Answer Key A.17 c REF: PM 2.4.2 Power = (Mass flow rate)(Specific heat)(temperature increase)

Power = (50 GPM)(8.34 lbs/gallon)(1 Btu/lb-deg F)(6 deg F)(60 min/hour)

Power = (150,120 Btu/hour)(1 kW/3413 Btu/hour) = 44.0 kW A.18. a REF: Reactor Physics Notes on Reactivity Feedback, pg. 21 A.19 c REF: LaMarsh, pgs 318 - 320 A.20. d REF: Glasstone and Sesonske, Nuclear Reactor Engineering, Chapter 5, Section 5.114

Answer Key B.1 c REF: Reactor Training Manual - 10CFR20 B.2 d REF: Reactor Training Manual - Ionizing Radiation B.3 a, 4; b, 2; c, 1; d, 3 REF: Reactor Training Manual - Health Physics B.4 b REF: 10CFR50.54(y).

B.5 b REF: TS 3.9 B.6 c REF: SOP I & 10 CFR 50.59 B.7 a REF: 10 CFR 20.1003 Definitions B.8 c REF: TS 3.5.1 B.9 c REF:

DR1 DR2 DR1 2 2000 2

= X 22 = X X2 = x 1 = 400 ft 2 X = 20 ft X 2 2 X 12 DR2 5 B.10 d REF: PM 5.3.1 - 5.3.4 B.11 a REF: TS 2.1 B.12 b REF: Nuclear Reactor Engineering, Glasstone & Sesonske, Section 2.85, p 60; MIT Exam Bank, Catagory C, #4.

B.13 c REF: TS 3.11.2.d B.14 b REF: TS 6.1 B.15 b REF: 10CFR55.53(f)(2))

Answer Key B.16 b REF: PM 2.3.2 B.17 c REF: PM 5.1.3 B.18 c REF: PM 2.3, Step 11, p 2.

B.19 d REF: RSM 4.2.

Answer Key C.1 b REF: RSM 8.4, AOP PM 5.5.7 C.2 d REF: RSM, Pages 9.3 to 9.5 C.3 a REF: RSM, Page 8.31 C.4 a REF: RSM, Page 8.31 C.5 a REF: RSM, Page 8.19 C.6 a REF: Emergency Plan Section 4.3.1.2.1.

C.7 d REF: PM 5.2.11 C.8 a, d REF: RSM, Page 8.12 C.9 a REF: PM 2.7, p 3.

C.10 d REF: AOP PM 5.8.2 C.11 a REF: PM 3.3.1, p 2.

C.12 b REF: AOP PM 5.6.4, RSM 7.6 C.13 b DELETED REF: Facility Comment to Written Exam administered in Sept. 2000 C.14 a REF: PM 5.7.12 C.15 c REF: RSM, Page 8.18 C.16 b REF: RSM, Page 3.3

Answer Key C.17 a REF: RSM Pages 7.7, 8.19; SAR Rev 36, Sec 12.2; SR#-0-88-11, p 2.

C.18 a REF: PM 1.10, Step 4.14B, p 14.

C.19 b REF: PM 5.5.4 C.20 b REF: PM 5.2.14