Text

Provides NRC W/Encl Exam Comments for Exam Given on 941024 at Plant
	ML20077B415
Person / Time
Site:	Brunswick
Issue date:	11/04/1994
From:	Richard Anderson; CAROLINA POWER & LIGHT CO.
To:	Ebneter S; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
Shared Package
ML20077B352	List: ML20077B415;
References
	NUDOCS 9411290229
	Download: ML20077B415 (8)
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. e ENCLOSURE 3 CP&L Carolina Power & Light Company Roy A. Anderson PO Box 10429 Vice President Southport NC 28461 Brunswick Nuclear Plant 910 457-2496 Novenber 4,1994 Serial: BSEP 94-0445 Mr. Stewart D. Ebncter Regional Administrator l ATTENTION: Mr. T. A. Peeb!es l United States Nuclear Regulatory Commission 101 Marietta Street, N. W., Suite 2900 Atlanta, GA 30323 BRUNSWlCK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 DOCKET NO. 50-325 & 50-324/ LICENSE NO. DPR-71 & DPR-62 COMMENTS ON NRC LICENSE EXAMINATION

Dear Mr. Ebneter:

l This letter provides the NRC Staff with the enclosed examination comments for the examination given on October 24,1994 at the Brunswick Nuclear Plant.

If you have any questions regarding this submittal please contact Mr. M. S. Williams at (910) 457-2056.

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Sincerely,

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A. Anderson t

SHC/shc (examcom.ltr)

Attachments cc: NRC Document Control Desk Mr. P. D. Milano, NRC/NRR Senior Project Manager - Brunswick Mr. D. C. Payne, NRC Region ll, Operations Branch Mr. C. A. Patterson, NRC Senior Resident inspector - Brunswick The Honorable H. Wells, Chairman - North Carolina Utilities Commission 9411290229 941122 PDR ADOCK 05000324 V PDR

i ENCLOSURE 1 i

BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 NRC DOCKET NOS. 50-325 & 50-324 OPERATING LICENSE NOS. DPR-71 & DPR-62 COMMENTS ON NRC LICENSE EXAMINATION 6

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BRUNSWICK NUCLEAR PLANT ,

COMMENTS ON NRC LICENSE EXAMINATION ,

Reactor Operator (RO) question No. 79 / Senior Reactor Operator (SRO) question No. 75:

NRC

Reference:

SM-305-A, Fuel Handling Student Study Material, Rev 0, pg 32, LO 16.b, KA 295023K204 ,

There are two correct answers. The intent of the question was to test knowledge of the "All Rods In" interlock circuitry for Refueling. Choice "C" is a correct answer. The plant conditions given in the question have SRMs A and C below the minimum counts of 3 cps for refueling. This leaves SRMs B and D to satisfy Technical Specification 3.9.2. These two instruments are in physically opposite quadrants of the core. As such, these SRMs can not satisfy the requirement for two operable SRMs, one in the quadrant where fuel is being moved, and one in an adjacent quadrant. Choice "B" is also a correct answer.

Recommend accepting answers "B" and "C" 1

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B. CONDITIONS MONITORED The refueling interlock circuitry senses various refueling bridge conditions and control rod positions. Depending on the sensed condition, interlocks may be actuated which prevent fuel movement (Bridge Reverse Motion Stop, Main Hoist Block, Auxiliary Hoist Block) and/or prevent control rod withdrawal (Rod Block). The circuitry senses:

ALL RODS IN BRIDGE NEAR/OVER REACTOR VESSEL ANY HOIST LOADED MAIN HOIST NOT FULL-UP ALL RODS IN - To minimize loading fuel into a cell containing a withdrawn control rod, it is required that all control rods are fully inserted when fuel is being loaded into or removed from the reactor core.

The position for each control rod is established by the closure of full-in magnetically-operated reed switches within the control rod drive mechanism. The rod full-in reed switch must be closed for each control rod before the ALL RODS i IN signal is generated.

BRIDGE NEAR/OVER REACTOR VESSEL - The assumption is that the refueling bridge is in a position where fuel can be loaded into the reactor vessel.

The refueling bridge is provided with two limit switches attached to the drive housing (LSI & 132). The switches actuate before the refueling bridge or any of its hoists are physically located over the reactor core providing indication that the bridge is near or over the reactor vessel.

LS1 is active when the reactor mode switch is in REFUEL.

LS2 is active when the reactor mode switch is in STARTUP.

ANY HOIST LOADED - The assumption is that a fuel assembly is being moved using the refueling bridge.

Each hoist (Main, Monorail, and Frame-mounted) provide a hoist loaded signal to 1 the refueling interlock circuitry. Load sensing is accomplished by Dillon Force Switches for each hoist; the switches are actuated at 720# for the Main Hoist and !

at 325# for the Monorail and Frame-mounted Hoists providing positive indication whenever fuel is loaded on any hoist.

MAIN HOIST NOT FULL-UP - Indicates that the Main Hoist is being lowered for fuel movement even it is not loaded. ,

LOI-CLS-SM-305-A Page 32 of 99 Rev.O

RNACTOR OPERATOR Page 47 e

( U STION: 079 (1.00)

Given the following plant conditions:

-- Reactor has been shutdown for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

-- Control rod 13-12 indicates notch "00" but does not indicate a full in, green light

-- SRMs A and C are reading 2 cps

-- SRMs B and D are reading 5 and 8 counts respectively

-- Fuel pool level is 23 feet 2 inches above the fuel assemblies seated in the spent fuel storage rack With the above conditions which one of the following would prevent removal of spent fu21 from the reactor core, either procedurally or because of a refueling interlock?

a. The level of water above the fuel assemblies when seated in the spent fuel storage rack is not high enough.

b. The required number of SRM channels are not available per Technical Specifications. ,

c. The failure of rod 13-12 to generate a rod full in signal to RMCS.

d. The required time before spent fuel can be removed from the reactor core has not elapsed.

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'SENI.OR REACTOR OPERATOR Paga 44 ;

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QUESTION: 075 (1.00)

Given the following plant conditions: l 0 - Reactor has been shutdown for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

- Control rod 13-12 indicates notch "00" but does not indicate a full in, green light i'

-- SRMs A and C are reading 2 cps

-- SRMs B and D are reading 5 and 8 counts respectively

-- Fuel pool level is 23 feet 2 inches above the fuel assemblies -

seated in the spent fuel storage rack !

With the above conditions which one of the following would prevent i removal of spent fuel from the reactor core, either procedurally or because of a refueling interlock?

a. The level of water above the fuel assemblies when seated in the ,

spent fuel storage rack is not high enough.

b. The required number of SRM channels are not available per :

Technical Specifications.

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c. The failure of rod 13-12 to generate a rod full in signal to !

RMCS.

d. The required time before spent fuel can be removed from the ;

reactor core has not elapsed.

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! REFUELING OPERATIONS

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3/4.9.2 INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 During CORE ALTERATIONS, the requirements for the source range monitors (SRMs) shall be: .

a. Two SRMs* shall be OPERABLE, one in the core quadrant where fuel is being moved and one in an adjacent quadrant. For an SRM to be considered OPERABLE, it shall be inserted to the normal operating level and shall have a minimum of 3 cps except as specified in d and e below.

b. The SRMs shall give a continuous visual indication in the Control Room.

c. The " shorting links" shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn ** and shutdown margin demonstrations. _

d. During a core SPIRAL UNLOAD the count rate may drop below 3 cps.

e. Prior to a core SPIRAL RELOAD, up to four fuel assemblies shall be loaded into different control cells containing control blades around r

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\x' each SRM to obtain 3 cps. Until these assemblies have been loaded, the 3 cps count rate is no't required.

APPLICABILITY: OPERATIONAL CONDITION 5 t ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and fully insert all insertable control rods. The provisions of Specification 3.0.3 are not applicable.

l The use of special movable detectors during CORE ALTERATIONS in place of the normal SRM nuclear detectors is permissible as long as these special detectors are connected to the normal SRM circuits.

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- Not required for control rods removed per Specifications 3.9.10.1 or 3.9.10.2.

i BRUNSWICK - U.IIT 2 3/4 9-3 Amendment No. 114 i

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Figure 2 - IN-CORE INSTRUMENTATION IDCATION O .

Pag- 23 of 39 Rev.0 OPS-CLS-SM-009-A

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SITE SPECIFIC EXAMINATION REACTOR OPERATOR LICENSE REGION 2 CANDIDATE'S NAME:

FACILITY: Brunswick 1 & 2 REACTOR TYPE: BWR-GE4 DATE ADMINISTERED: 94/10/24 INSTRUCTIONS TO CANDIDATE:

Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires a final grade of at least 80%. Examination papers will be picked up four (4) hours after the examination starts.

CANDIDATE'S TEST VALUE SCORE %

100.00 % TOTALS FINAL GRADE All work done on this examination is my own. I have neither given nor received aid.

Candidate's Signature I

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O-l Page 2 l EEACTOR OPERATOR i

! ANSWER SHEET Multiple Choice (Circle or X your choice)

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

MULTIPLE CHOICE 023 a b c d 001 a b c d 024 a b c d 002 a b c d 025 a b c d l

003 a b c d 026 a b c d 004 a b c d 027 a b c d l l'

005 a b c d 028 a b c d l 006 a b c d 029 a b c d I

I 007 a b c d 030 a b c d l

l 008 a b c d 031 a b c d 009 a b c d 032 a b c d l

010 a b c d 033 a b c d 011 a b c d 034 a b c d 012 a b c d 035 a b c d 013 a b c d 036 a b c d i

1 014 a b c d 037 a b c d l l

l 015 a b c d 038 a b c d 016 a b c d 039 a b c d )

I j 017 a b c d 040 a b c d I 018 a b c d 041 a b c d 019 a b c d 042 a b c d j 020 a b c d 043 a b c d 021 a b c d 044 a b c d 022 a b c d 045 a b c d i

t _ _ _ - _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ . _ _ _ _ ._________________________ _ _ ___ ____

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4w A-REACTOR OPERATOR Page 3 ANSWER SHEET Multiple Choice (circle or X your choice)

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

046 a b c d 068 a b c d MULTIPLE CHOICE 069 a b c d 047 a b c d 070 a b c d 048 a b c d 071 a b c d 049 a b c d 072 a b c d 050 a b c d 073 a b c d 051 a b c d 074 a b c d 052 a b c d 075 a b c d 053 a b c d 076 a b c d 054 a b c d 077 a b c d 055 a b c d 078 a b c d 056 a b c d 079 a b c d 057 a b c d 080 a b c d 058 a b c d 081 a b c d l 1

059 a b c d 082 a b c d j 060 a b c d 083 a b c d I i

1 061 a b c d 084 a b c d ]

062 a b c d 085 a b c d I

063 a b c d 086 a b c d l 064 a b c d 087 a b c d 065 a b c d 088 a b c d 066 a b c d 089 a b c d 067 a b c d 090 a b c d 1

e a REACTQR OPERATOR Page 4 ANSWER SHEET Multiple Choice (Circle or X your choice)

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

091 a b c d MULTIPLE CHOICE 092 a b c d 093 a b c d 094 MATCHING a

b c

d MULTIPLE CHOICE 095 a b c d 096 a b c d 097 a b c d 098 a b c d 099 a b c d

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

e-a Page 5 NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination, the following rules apply:

1. Cheating on the examination will result in a denial of your application and could result in more severe penalties.

2. After you complete the examination, 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.

3. To pass the examination, you must achieve a grade of 80 percent or greater.

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

5. There is a time limit of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for completing the examination.

6. Use only black ink or dark pencil to ensure legible copies.

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7. Print your name in the blank provided on the examination cover sheet and the answer sheet.

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8. Mark your answers on the answer sheet provided and do not leave any question blank.

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

10. Restroom trips are permitted, but only one applicant at a time will be ,

allowed to leave. Avoid all contact with anyone outside the examination room to eliminate even the appearance or possibility of cheating. i

11. When you complete the examination, assemble a package including the examination questions, examination aids, and answer sheets and give it I to the examiner or proctor. Remember to sign the statement on the examination cover sheet.

12. After you have turned in your examination, leave the examination area as i defined by the examiner.

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THIS PAGE IS INTENTIONALLY BLANK

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REACTQR OPERATOR Page 7 QUESTION: 001 (1.00)

Which one of the following describes an " intent" temporary procedure change?

Procedure intent is changed if a temporary procedure change:

a. places the plant at a lower power level.

b. adds a new method to start a pump.

c. widens the nuclear boiler water level operating range.

d. modifies the text alongside a Q/C holdpoint.

QUESTION: 002 (1.00)

Which one of the following is the preferred method of installing a Temporary Caution Tag on a RTGB panel?

a. Caution tags are placed on the vertical portion of the RTGB as close as possible to the affected component, with the caution tag sheet placed in an orange colored book kept in an annunciator procedure holder.

b. A numbered orange caution tag sticker is placed on or near the affected component and the caution tags are placed on the horizontal portion of the RTGB.

c. A numbered orange caution tag sticker is placed on or as near as possible to the affected component, with the caution tag sheet placed in an orange colored book kept in an annunciator procedure holder.

d. A numbered orange caution tag sticker is placed on the component along with the caution tags, which are folded as necessary to prevent obstructing of controls and indications.

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REACTOR OPERATOR Page 8 QUESTION: 003 (1.00)

What action is required when independent verification is required for hanging a clearance?

a. Each verification must always be separated by time and distance, to ensure independence.

b. Only the second operator is required to initial the clearance tag sheet.

c. Independent verification must be used when restoring equipment with a completed valve and electrical lineup to service.

d. Management oversight verification may be used in place of independent verification, for Control Room components.

QUESTION: 004 (1.00)

Which one of the following is the maximum time period allowed for Temporary Plant Modifications (TPMs) from date of installation, without requiring an extension?

a. Three months

b. Six months

c. Twelve months

d. 18 months I

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REACTQR OPERATOR Page 9 i l

QUESTION: 005 (1.00)

A High Pressure Coolant Injection (HPCI) system motor operated valve, being stroke timed in accordance with the Inservice Test (IST) Program, is found to exceed the full stroke time limit on the first attempt.

Which one of the following actions is required?

a. Do not stroke the valve a second time, declare the valve inoperable and generate a WR/JO.

b. Stroke the valve a second time, if satisfactory the valve is operable.

c. Reverify the system test alignment is correct, then attempt a second stroke test of the valve.

d. Manually operate the valve through a complete cycle to check for binding, then stroke the valve a second time.

QUESTION: 006 (1.00)

Which one of the following indications on an annunciator window, indicates the annunciator card has been pulled?

a. Black Dot

b. Yellow Dot

c. Red Dot

d. Blue Dot

r O* O REACTQR OPERATCR Page 10 QUESTION: 007 (1.00)

You are performing a valve lineup on a valve with an orange handwheel, in a system requiring independent verification.

Which one of the following describes the requirements that apply to this valve?

a. The orange handwheel designates this valve as a throttle valve, the independent verifier should watch you position it.

b. Independent verification requirements are waived for this valve, due to excessive radiation levels.

c. This valve is normally full open and should be stroked in the clockwise direction one-half turn and then reopened.

d. The valve handwheel should be turned in the counterclockwise direction to verify position.

QUESTION: 008 (1.00)

Which one of the following process computer indications indicates a failed or bypassed LPRM on the "LPRM Readings Edit" function?

a. Magenta "V"

b. Blue "nnS"

c. Yellow "nn"

d. Red "++"

b - - - _ _ - _ - _- - - - _ - - - - - - - - - _

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O O REACTOR OPERATOR Page 11 QUESTION: 009 (1.00)

During operation at rated conditions the Control Operator reports an increase in conductivity on P603 Conductivity Recorder, G31-CRS-R601 and a decrease in reactor power of 4%.

Which one of the following is the cause of these indications?

This condition indicates that a:

a. reactor water cleanup resin bed is exhausted, releasing accumulated silica.

b. main condenser tube has failed.

c. resin injection has occurred.

d. condensate deep bed demineralizer resin bed is exhausted, releasing accumulated boron.

QUESTION: 010 (1.00)

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

It is the sum of the:

a. Shallow Dose Equivalent, Whole Body (SDE, WB) and the Committed l Effective Dose Equivalent (CEDE). ]

b. Deep Dose Equivalent (DDE) and the Committed Dose Equivalent (CDE).

c. Deep Dose Equivalent (DDE) and the Committed Effective Dose 1 Equivalent (CEDE).

d. Shallow Dose Equivalent, Max Extremity (SDE, ME) and the Deep Dose Equivalent (DDE).

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REACTOR OPERATOR Page 12 QUESTION: 011 (1.00)

Given the following Conditions:

-- 21 year old radiation worker

-- Current NRC Forms 4 and 5 are on file for the worker

-- The worker has received 47.0 Rem Committed Dose Equivalent (CDE) for the year

-- The workers current annual Deep Dose Equivalent (DDE) dose is 0.0 Rem Which one of the following is the maximum amount of DDE that the worker can receive for the remainder of the year, in accordance with 10 CFR 20?

a. 2.5 Rem

b. 3.0 Rem

c. 8.0 Rem

d. 13.0 Rem QUESTION: 012 (1.00)

Which one of the following is the minimum time period requiring abstinence from alcohol, preceding any scheduled shift in accordance with 10 CFR 26, " Fitness for Duty Program?"

a. 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months

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

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

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

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REACTOR OPERATOR Page 13 I l

QUESTION: 013 (1.00)

Which one of the following designates a Main Control Room annunciator that indicates an Emergency Operating Procedure entry condition?

a. Red backlighting with blue bar.

b. Red backlighting with amber bar.

c. White backlighting with amber bar.

d. White backlighting with blue bar.

QUESTION: 014 (1.00)

Which one of the following describes the dispersement of the 50 gpm control rod drive cooling water flow, assuming no control rod movement?

a. All 50 gpm passes through the Pressure Control Valve (F003).

b. 44 gpm though the Pressure Control Valve (F003) and 6 gpm to the recirculation pump seal purge.

c. 44 gpm through the Pressure Control Valve (F003), 2 gpm through the insert stabilizing valves and 4 gpm through the withdraw stabilizing valves.

d. 44 gpm through the Pressure Control Valve (F003), 4 gpm through the insert stabilizing valves and 2 gpm through the withdraw stabilizing valves.

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dh a REACTOR OPERATOR Page 14 QUESTION: 015 (1.00)

With the plant at normal operating pressure and temperature the operator partially closes the control rod drive Pressure Control Valve (F003).

Which one of the following describes the effect this will have on control rod speeds?

Partially closing the pressure control valve:

a. decreases control rod insertion speed.

b. increases control rod withdrawal speed.

c. decreases control rod scram speed.

d. increases control rod scram speed.

QUESTION: 016 (1.00)

Which one of the following conditions will cause the reactor manual control system to initiate a SELECT BLOCK?

a. Momentarily positioning (< 2 seconds) the RMCS timer test switch to TEST.

b. The operator attempts to select a rod that has a withdraw error.

c. The operator attempts to select a rod that has an insert error,

d. Withdraw signal lasting for more than 2 seconds with the rod out notch override switch in OFF.

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O w REACTOR OPERATOR Page 15 QUESTION: 017 (1.00)

Given the following plant conditions:

-- Unit 2 is operating at 100% power

-- Annunciator, " SPEED CONTROL A SIGNAL FAIL" is received for the "A" reactor recirculation pump ;

I Which one of the following is the expected effect on operation of the I "A" recirculation pump?

The "A" recirculation pump speed:

a. will run back to minimum due to the low output signal from the controller. J

b. can only be changed by the individual pump controller at P603.

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c. will remain at its existing value until the scoop tube lock. can be reset.

d. will not change due to loss of power to the scoop tube l positioner. l i

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l QUESTION: 018 (1.00)

During operation at 65% reactor power, a spurious electrical fault j causes the "A" reactor recirculation pump discharge valve to close to l the 88% open position. Assume a normal valve lineup.

Which one of the following is the automatic response of the reactor recirculation system?

l a. The "A" pump trips.

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b. Scoop tube lockup occurs on both pumps.

c. Scoop tube lockup on the "A" pump. 4 l

d. The "A" pump speed runs back to 28% of maximum. l 1

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O O REACTOR OPERATOR Page 16 QUESTION: 019 (1.00)

Which one of the following RHR pump combinations will lose its power supply if a fault occurs on 4160 Vital Bus E3?

a. lA/2A

b. 1B/2B

c. 1C/2C ,

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d. 1D/2D I l

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QUESTION: 020 (1.00)

The LPCI valves listed below have time interlocks, which do not allow them to be closed or throttled following a LPCI initiation signal. j

-- LPCI Outboard injection Valve, F017A(B)

-- Inboard Injection Valve, FO15A(b, I

-- RHR Heat Exchanger Bypass Valve, FO48A(B)

Which one of the following describes the time interlocks associated with these valves? I

a. FO17A(B) 3 minutes ,

F015A(B) duration of LPCI initiation signal l FO48A(B) 5 minutes !

b. F017A(B) 5 minutes FO15A(B) 5 minutes FO48A(B) 3 minutes

c. F017A(B) 3 minutes F015A(B) 10 minutes F048A(B) 5 minutes i l

d. FO17A(B) 5 minutes l FO15A(B) duration of LPCI initiation signal FO48A(B) 3 minutes l

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r-A A REACTOR OPERATOR Page 17 QUESTION: 021 (1.00)

The following conditions exist for Unit 1:

-- Reactor water level +110 inches

-- Drywell pressure +2.0 psig

-- Reactor pressure 100 psig

-- Alarm "HPCI VAC BRKR VLV F075/F079 NOT FULL OPEN"

-- Alarm "HPCI TURB TRIP"

-- Alarm "HPCI TRIP SOL ENER" Which one of the following is the response of HPCI, if reactor pressure increases to greater than 120 psig and HPCI injection is required?

HPCI low pressure isolation:

a. must be manually reset before HPCI will automatically inject,

b. will automatically reset and HPCI will inject.

c. must be manually reset and the injection valve must be manually opened.

d. will automatically reset but Group 4 steam supply valves must be manually opened before HPCI will automatically inject.

QUESTION: 022 (1.00)

Which one of the following describes how injection from HPCI is terminated and prevented if HPCI is not operating during an ATWS event?

a. Depress the HPCI manual isolation pushbutton. j

b. Close the HPCI Injection Valve, E41-F006. I I

c. Depress and hold the HPCI turbine trip pushbutton.

d. Place the HPCI auxiliary oil pump in " Pull-To-Lock."

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REACTQR OPERATOR Page 18 QUESTION: 023 (1. 00)

When operating HPCI for level or pressure control in accordance with the EOPs, the operator is directed to defeat the HPCI high suppression pool

_(SP) level suction transfer in accordance with EOP-01-SEP-10, " Circuit Alteration Procedure."

Which one of the following describes the affect this has on the HPCI suction valve logic?

The HPCI suction valves:

a. can not be transferred manually or automatically on a CST low level signal or SP high level signal.

b. can only be transferred manually if a CST low level signal and a SP high level signal exist.

c. will automatically transfer if a CST low level signal exists,

d. will automatically transfer if a SP high level signal exists.

QUESTION: 024 (1.00)

Which one of the following is the purpose of the " white" indicating light, E21-DS20B/A, associated with each core spray pump?

The white light indicates that the:

a. pumps associated 4160 VAC breaker is closed with control power available.

b. pump was manually stopped while an initiation signal was still present.

c. pumps associated 4160 VAC breaker was locally closed with emergency diesel generator power being supplied. J

d. pump has tripped on a fault with a valid initiation signal present.

r REACTOR OPERATOR Page 19 QUESTION: 025 (1.00)

Which one of the following is the power supply to Core Spray Pump, 1B?

Core Spray Pump, 1B is powered from 4160V Emergency Bus:

a. El.

b. E2.

c. E3.

d. E4.

QUESTION: 026 (1.00)

Given the following conditions on Unit 2:

-- An ATWS has occurred

-- Rods are being driven in using RMCS per LEP-02

-- Stand by liquid control (SLC) Pump 2A is OUT-OF-SERVICE

-- Reactor vessel water level is being maintained at TAF

-- Reactor power is 5%.

-- SLC Pump 2B is injecting

-- HPCI is being used for vessel level control There is a subsequent Loss of Off-Site Power and only DG3 starts and loads onto its respective emergency bus.

What action shcald be taken to continue boron injection?

a. Inject boron with HPCI.

b. Inject boron witn RCIC.

c. Inject boron with CRD Pump 2A.

, d. Inject boron with RWCU Pump 2A.

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b U REACTOR OPERATOR Page 20 QUESTION: 027 (1.00)

The following conditions exist for Unit 2:

-- The reactor has scrammed and the mode switch is in SHUTDOWN ;

-- The problem has been identified and corrected

-- Alarm "SDV HI HI LEVEL RPS TRIP" is actuated The operator places the scram discharge volume high level scram keylock I switch to BYPASS, turns the scram reset switch to both directions and I then places the mode switch to STARTUP. ,

Which one of the following describes the expected reactor protection j system response? ,

a. No system response for the present plant conditions.

b. The reactor scram will reset and remain reset.

c. The reactor scram will reset and again scram.

d. The reactor scram will reset when the scram discharge volume drains.

QUESTION: 028 (1.00)

Which one of the following describes the status of the RPS components, IMMEDIATELY following an inadvertent manual actuation of alternate rod insertion (ARI)?

a. Scram Pilot Solenoid Valves - De-energized Backup Scram Valves - Open ARI Solenoid Valves - De-energized

b. Scram Pilot Solenoid Valves - Energized Backup Scram Valves - Closed ARI Solenoid Valves - De-energized

c. Scram Pilot Solenoid Valves - De-energized Backup Scram Valves - Open ARI Solenoid Valves - Energized

d. Scram Pilot Solenoid Valves - Energized Backup Scram Valves - Closed ARI Solenoid Valves - Energized

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REACTOR OPERATOR Page 21 QUESTION: 029 (1.00)

A reactor startup is in progress with IRMs on Range 2. While withdrawing SRM detectors, the operator inadvertently selects IRM detector E drive unit.

Which one of the following describes the effect this has on the IRM system?

a. IRM E detector will NOT withdraw because companion APRM E is downscale.

b. IRM E detector will withdraw and "IRM A UPSCALE /INOP" annunciator will be received.

c. IRM E will withdraw and a control rod block will be generated when the detector begins to withdraw.

d. No action occurs because the range selector switch for IRM E is on Range 2 and the mode switch is not in "RUN."

QUESTION: 030 (1.00)

Given the following plant conditions:

-- Reactor startup is in progress

-- Range switches for IRM's A, C, E, are on Range 3

-- IRM G is bypassed.

Which one of the following describes the status of the SRM division "A" rod blocks?

a. No SRM division "A" rod blocks are bypassed.

b. Only the SRM downscale is bypassed.

c. Both SRM INOP and downscale are bypassed.

d. Both SRM downscale and detector not full in are bypassed.

r e e Page 22 REACTQR OPERATOR QUESTION: 031 (1.00)

During a reactor startup the SRM system loses monitoring capability and a rod block is received.

Which one of the following power sources has been lost?

a. 24 Volt DC

b. 120 VAC Emergency Power

c. RPS Power System

d. 120 VAC UPS QUESTION: 032 (1.00)

Which one of the following is the maximum number of LPRMs that can have their function switch out of OPERATE for APRM channel "D," without generating a "APRM UPSCL TRIP /INOP SYS B" annunciator?

a. 3

b. 4

c. 6 l l

d. 7 ,

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we REACTOR OPERATOR Page 23 QUESTION: 033 (1.00)

With Unit 1 operating at 100% power, APRM channel "C" has the following LPRMS bypassed:

04-37B 36-05B 12-29C 04-21D 28-29A With these conditionsLPRM 20-05D fails downscale.

Which one of the following is required, assuming no other APRMs are inoperable? [(A listing of LPRM detectors included in APRM channels is available.]

a. No action is required, continue operation with no restrictions.

b. BypassLPRM 20-05D and continue operation with no restrictions.

c. BypassLPRM 20-05D and declare APRM channel "C" inoperable,

d. BypassLPRM 20-05D, declare APRM channel "C" inoperable and insert a trip on RPS channel "A."

QUESTION: 034 (1.00)

Which one of the following describes the function of the reference leg backfill system?

a. Prevent flashing of the reference leg during rapid depressurization below 400 psig by providing cooling water to the reference leg.

b. Ensure the reference leg remains full.

c. Prevent erroneous low water level indication during a major steam leak,

d. Prevent migration of non-condensable gases into the reference leg by increasing reference leg spillover.

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BEACT,OR OPERATOR Page 24 QUESTION: 035 (1.00)

While restoring the RCIC system to standby status following an automatic initiation per OP-16, "RCIC System Operating Procedure," the operator inadvertently depresses the STEAM ISOLATION manual pushbutton instead of the TURBINE TRIP pushbutton.

Which one of the following describes the response of the RCIC Steam Supply Inboard Isolation Valve, E51-F007 and RCIC Steam Supply Outboard Isolation Valve, E51-F008?

a. Only the Steam Supply Inboard Isolation Valve, E51-F007, closes.

b. Only the Steam Supply Outboard Isolation Valve, E51-F008, closes.

c. Both the Steam Supply Inboard Isolation Valve, E51-F007, and the Steam Supply Outboard Isolation Valve, E51-F008, close,

d. NEITHER the Steam Supply Inboard Isolation Valve, E51-F007,or the steam Supply Outboard Isolation Valve, E51-F008, close.

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a-Page 25 SEACTOR OPERATOR QUESTION: 036 (1.00)

Given the following plant conditions on Unit 2:

-- A small break LOCA is in progress

-- HPCI is inoperable

-- Reactor vessel level TAF

-- Reactor pressure 800 psig and lowering

-- All low pressure ECCS pumps running

-- Seven ADS valves opened on automatic initiation A loss of off-site power occurs and all diesel generators tie onto their emergency buses to restore power.

Which one of the following describes the ADS valve response?

The ADS valves will:

a. remain open.

b. close then re-open as low pressure ECCS pumps sequence on.

c. close then re-open 105 seconds after ECCS pumps sequence on.

d. close and will not automatically re-open.

QUESTION: 037 (1.00)

Which one of the following describes the automatic alignment of the ADS pneumatic supply with a LOCA signal present?

The ADS pneumatic supply automatically aligns to the:

a. pneumatic nitrogen system (PNS) backed up by reactor building ;

noninterruptible air (RNA) system. l

b. PNS backed up by backup nitrogen system. l l

c. backup nitrogen system.

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d. RNA system. I 1

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A Ib SEACTOR OPERATOR Page 26 QUESTION: 038 (1.00)

Given the following reactor vessel level inputs to the digital feedwater control system (DFCS):

-- LT-N004A 183 inches

-- LT-N004B 190 inches

-- LT-N004C 188 inches

-- Level select switch on P-603 selected to "A" Which one of the following will be the selected level input of the DFCS?

a. 183 inches

b. 187 inches

c. 188 inches

d. 190 inches QUESTION: 039 (1.00)

Which one of the following will automatically remove the "3 element permissive" from the digital feedwater control system?

a. Reactor vessel level instrument out of tolerance

b. Reactor feed pump control signal failure

c. Steam flow and feed flow decrease to less than 20% of total.

d. Reactor Scram

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e REACTOR OPERATOR Page 27 QUESTION: 040 (1.00)

During operation at 100% power, RPS Motor Generator Set "A" trips.

Which one of the following describes the affect this has va the Standby Gas Treatment (SBGT) system?

a. SBGT train "A" must be declared inoperable.

b. Only the SBGT train selected with the "SBGT A(B) PREF" control switch on RTGB XU-51 will start.

c. Neither SBGT train will automatically start.

d. Both SBGT trains will automatically start.

QUESTION: 041 (1.00)

Diesel Generator (DG) 3 tripped on high lube oil temperature while operating in LOCAL MANUAL. With DG3 in this condition, the master / slave breaker from 2D to E3 trips open.

l Which one of the following describes the expected response of DG3?

DG3 will:

a. start and energize bus E3.

b. start and trip after 45 seconds. l

c. not start due to local manual status.

d. not start until the lockout is manually reset.

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KEACTOR OPERATOR Page 28 QUESTION: 042 (1.00)

Following a scram from 100% power the ball check valve in the insert I port for one control rod drive mechanism malfunctions, and fails to unseat and shift position.

Which one of the following is the effect on that control rod's ability to scram?

The control rod will:

a. not insert until reactor and CRD pressures equalize.

b. hydraulically lock-up and remain stationary.

c. will fully insert at slower than normal speed.

d. will fully insert at higher than normal speed.

QUESTION: 043 (1.00)

Reactor power is 30%. Rod 06-23 from step 19 is selected for insertion to position 00 in accordance with GP-10. The RWM annunciator alarms with the following RWM OD screen:

SELF-TEST: FAULT SEQUENCE: A2 BPWS MODE: OPERATE BLOCKS: INSERT WITHDRAW STEP: 19 POWER: BELOW LPAP SR 06 - 23 : 08 IB WB A2 - 19 HELP DISPLAY OFF ETC Which one of the following describe why insert and withdraw blocks have occurred?

a. The selected rod is a select error.

b. The selected rod has invalid rod position data.

c. The RWM has detected a position error.

d. The self test has detected a critical failure.

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REACTOR OPERATOR Page 29 QUESTION: 044 (1.00)

A plant shutdown is in progress. The last control rod from step 6B has been inserted and another control rod has been selected.

The RWM display indicates:

SELF TEST:OK SEQUENCE: B2X BPWS 3 MODE: OPERATE BLOCKS: INSERT STEP: 6B POWER:BELOW LPSP WITHDRAW SR 30 -- 07 : 00 SE IB WB IE 06 -- 39 : 14 B2X--6B HELP DISPLAY OFF ETC What action should be taken to clear ALL existing RWM blocks?

a. Select any control rod from Rod Sequence Checkoff Sheet, step 6B.

b. Select any control rod from Rod Sequence Checkoff Sheet, step 6A.

c. Select and withdraw control rod 06-39 to position 16.

d. Withdraw control rod 30-07 to position 04.

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O O REACTOR OPERATOR Page 30 QUESTION: 045 (1.00)

When measuring total core flow, the individual recirculation loop flows are summed when both loops are in service.

Which cac of the following describes the method by which the circuitry j determines if both loops are in service.

The flow circuitry:

a. looks at the status of the recirculation pump breakers and the pump discharge isolation valves.

b. monitors for a minimum value of " forward" flow in each of the loops.

c. looks at the status of the recirculation pump breakers and the pump suction isolation valves.

d. monitors for minimum speed (28% limiter) operation of each of the recirculation pumps.

QUESTION: 046 (1.00) .

Which one of the following describes how the recirculation jet pump design provides core reflood capability?

a. All ECCS flow entering the core must flow through the jet pumps.

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b. The height of the jet pump throat inlets are at 2/3 active core j height. I

c. All flow leaving the active core area during a design basis accident must flow through the jet pumps.

d. Leakage at the jet pump mixer / diffuser joint is less than the capacity of two LPCI pumps.

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REACTOR OPERATOR Page 31 QUESTION: 047 (1.00)

Which one of the following describes the direct effect initiation of standby liquid control (SLC) has on the reactor water cleanup (RWCU) system.

Initiation of SLC will:

a. trip the RWCU pumps.

b. initiate a full Group 3 PCIS isolation,

c. bypass the filter demineralizers.

d. close the Outboard RWCU Isolation Valve, F004.

QUESTION: 048 (1.00)

RHR loop "A" is in shutdown cooling maintaining a 45 degrees F/hr cooldown rate.

Conditions are:

-- Recirculation loop temperature 298 degrees F

-- RHR HX outlet temperature 205 degrees F

-- RHR total flow rate 4150 gpm RHR HX service water (SW) outlet temperature 168 degrees F and slowly rising

-- RHR HX service water flow 3800 gpm

-- Both RHR Service Water Booster pumps are operating Which one of the following describes the required action for the above conditions? [20P-17 section 5.4, Shutdown Cooling Mode is available.]

a. Throttle open RHR SW Outlet Valve, E11-PDV-F068A.

b. Throttle closed RHR HX 2A Bypass Valve, E11-F048A.

c. Throttle closed RHR HX 2A Outlet Valve, E11-F003A.

d. Throttle open LPCI Outboard Injection Valve, E11-F017A.

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BEACTpR OPERATOR Page 32 QUESTION: 049 (1.00)

During operation at 100% control rod 26-15 drifts in to the " full in" position due to a leaking scram inlet valve.

Which one of the following will be the control rod position indication?

a. Green " full in" light and no indication on the 4-rod display.

b. Green " full in" light and "00" on the 4-rod display.

c. NO Green " full in" and "00" on the 4-rod display,

d. NO Green " full in" and no indication on the 4-rod display.

QUESTION: 050 (1.00)

Given the following conditions:

-- Unit 2 is at 40% power

-- APRM "D" fails "Downscale"

-- NO operator actions have been taken Which one of the following describes the automatic response of the Rod Block Monitor (RBM) system? (Consider only the effect of the RBM system.)

RBM channel "B":

a. enforces the high trip setpoint (HTSP).

b. automatically shifts to APRM "F".

c. sends a rod withdrawal block to RMCS. j 1

d. rod block logic is bypassed.

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o w w REACTOR OPERATOR Page 33 QUESTION: 051 (1.00)

Control rod 30-31 is selected for movement when all "D" level LPRMs associated with the rod fail downscale.

Which one of the following describes the effect the RBM will have on any subsequent rod motion?

The RBM will:

a. allow inward motion only.

b. allow outward motion only.

c. block both inward and outward motion.

d. allow both inward and outward motion.

QUESTION: 052 (1.00)

A LOCA signal on Unit 1 has resulted in an automatic closure signal to the Containment Spray Valves FOl6A/B and F021A/B.

Which one of the following conditions will allow manual opening of these valves, with the LOCA signal still present?

a. Drywell pressure less than 2.7 psig,

b. Containment spray valve control (THINK) switch in MANUAL with drywell pressure above 2.7 psig.

c. 2/3 core height manual override keylock switch in OVERRIDE with drywell pressure above 2.7 psig,

d. As soon as LPCI refloods the reactor vessel above 2/3 core height regardless of drywell pressure.

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a ah BEACTOR OPERATOR Page 34 QUESTION: 053 (1.00)

Which one of the following are the suppression chamber pressures, at which the initiation and termination of suppression pool sprays is required?

INITIATION TERMINATION

a. After Exceeding 13 psig Before Decreasing Below 2.5 psig

b. After Exceeding 13 psig Before Decreasing Below 2.7 psig

c. Before Exceeding 11 psig After Decreasing Below 2.7 psig

d. Before Exceeding 13 psig After Decreasing Below 2.5 psig QUESTION: 054 (1.00)

Both units are at rated power when the following Unit 1 Pneumatic Nitrogen System (PNS) alarms are received:

-- PNS SYS DIV I HDR PRESS LO

-- PNS SYS DIV I HDR PRESS LO-LO

-- PNS SYS DIV II HDR PRESS LO

-- PNS SYS DIV II HDR PRESS LO-LO

-- PNS SYS N2 STOR TANK PRESS LO Which one of the following describes the effect on the inboard main steam isolation valves (MSIVs)?

The MSIVs:

a. remain open due to automatic alignment of noninterruptible air to the drywell on low PNS pressure,

b. fail closed as pneumatic pressure depletes sufficiently to reposition MSIV pilot control valves.

c. immediately close due to de-energization of MSIV solenoids on low pneumatic supply pressure.

d. remain open due to automatic alignment of backup nitrogen supply to drywell on low PNS pressure.

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QUESTION: 055 (1.00)

Which one of the following are the power supplies to the two solenoids associated with each main steam isolation valve (MSIV)?

INBOARD MSIVs OUTBOARD MSIVs

a. 125 VDC "A" 125 VDC "B" RPS "A" RPS "B"

b. 125 VDC "B" 125 VDC "A" RPS "B" RPS "A"

c. 125 VDC "A" RPS "A" 125 VDC "B" RPS "B"

d. RPS "A" 125 VDC "A" RPS "B" 125 VDC "B" QUESTION: 056 (1.00)

During operation at 100% power a sudden complete loss of EHC occurs.

Which one of the following is the expected response of the turbine and associated support systems?

a. Turbine stop and control valves fail closed on low EHC pressure, turbine control valve fast closure actuates RPS trip, bypass valves function for approximately 60 seconds then fail closed.

b. Turbine trip on overspeed, turbine stop valves < 90% open actuates RPS trip, bypass valves fail closed.

c. Turbine trip on low EHC pressure, turbine stop valves < 90% open actuates RPS trip, bypass valves function for approximately 60 seconds then fail closed.

d. Turbine trip on low EHC pressure, turbine stop valves < 90% open actuates RPS trip, bypass valves fail closed.

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Page 36 REACTOR OPERATOR QUESTION: 057 (1.00) i Which one of the following is the normal suction source for the control rod drive pumps during operation at 100% power?

a. Main condenser hotwell.

b. Discharge of condensate pumps prior to the steam packing exhauster.

c. Direct from condensate storage Lank (CST).

d. Hotwell reject to CST line.

QUESTION: 058 (1.00)

A Unit 2 4160V AC breaker has been racked in following maintenance on its associated pump. The operator racking in the breaker installed the control power fuses but failed to position the breaker toggle switch to the "ON" position.

Which one of the following is the affect on breaker operability?

The breaker will have:

a. normal control room indications but no remote operation will be possible.

b. normal control room indications but can be only be closed by local pushbutton.

c. no control room indications and cannot be operated by the remote controls.

d. no control room indications but may be closed with the local pushbutton.

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Ow REACTpR OPERATOR Page 37 QUESTION: 059 (1.00)

During operation at 100% power, Augmented Off-Gas (AOG) system annunciator, APP UA-39 2-2, DISCHARGE H2 CONC HIGH is received.

Which one of the following is the first automatic action that results from this condition.

a. Isolation and bypass of the AOG system after a 30 second time delay.

b. Immediate actuation of the AOG process timer.

c. AOG Bypass Valve, HCV-102 and Filter House Loop Seal Reservoir Drain Valve, SV-4907 close after a 15 minute time delay.

d. Recombiner strip heaters immediately energize.

QUESTION: 060 (1.00)

During operation at 100% power, a valid main steam line radiation signal greater than three times full power background is received on channels "A", "B", and "C."

Which one of the following statements describes ALL of the protective actions which result?

a. Reactor scrams from main steam line radiation signal, full Group 1 isolation and isolation of the condenser mechanical vacuum pump.

b. Reactor scrams from MSIV position signal, Group 1 isolation, isolation of the main steam line drains and isolation of the condenser mechanical vacuum pump.

c. Half scram RPS channel "A" and PCIS channel "A" Group 1 isolation signal.

d. Reactor scram from main steam line radiation signal, Group 1 isolation and isolation of recirculation sample valves.

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I EEACTOR OPERATOR Page 38 QUESTION: 061 (1.00)

Which one of the following systems injects inside the reactor vessel shroud?

a. Reactor Water Cleanup Return

b. Standby Liquid Control

c. Reactor Feedwater

d. Low Pressure Coolant Injection QUESTION: 062 (1.00)

The Control Room Operator records the following integrator data, four hours after the previous data was taken:

-- Drywell floor drain sump; gallons pumped since last reading

= 384 gallons

-- Drywell equipment drain sump; gallons pumped since last reading

= 1176 gallons Which one of the following is the reactor coolant system unidentified leakage rate?

a. 1.6 gpm.

b. 3.3 gpm.

c. 4.9 gpm.

d. 6.5 gpm.

O O w w BEACT,OR OPERATOR Page 39 QUESTION: 063 (1.00)

Given the following conditions on Unit 2:

-- Reactor mode switch is in " REFUEL"

-- Control rod 22-31 is withdrawn for testing

-- Rod select power switch is in the "ON" position Which one of the follcwing will generate a rod block with these conditions?

A rod out block will result if the:

a. operator turns the rod select power switch to "OFF" then "ON" and selects rod 30-31.

b. refueling bridge is over the core.

c. rod select pushbutton for control rod 38-31 is depressed.

d. refueling bridge console is deenergized.

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QUESTION: 064 (1.00)

Unit 1 is in the process of eff loading the core when the Unit 1 fuel pool cooling pump fails.

Which one of the following systems can be aligned to cool the Unit 1 fuel pool?

a. Supplemental spent fuel pool cooling system

b. RHR "A" loop

c. RHR "B" loop

d. Unit 2 fuel pool cooling system

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w J}EACT,OR OPERATOR Page 40 QUESTION: 065 (1.00)

Which one of the following describes why EOP-01, " Level Power Control" directs the operator to inhibit ADS?

Inhibiting ADS:

a. prevents discharging boron to the suppression pool via the SRVs.

b. ensures RPV pressure remains high enough to scram control rods.

c. ensures adequate natural circulation is retained for boron mixing.

d. prevents the power excursion from an uncontrolled coolant injection.

QUESTION: 066 (1.00)

During a Unit 2 startup, a loss of extraction steam resulted in the following conditions:

-- Reactor power 60%

-- Core flow 30 mlbs/hr

-- Core plate d/p 6.0 psid

-- Generator output 425 MWe Which one of the following is the required action? [ Figure 1, Thermal Power Limitations is available.]

a. Initiate Select Rod Insert

b. Manually scram the reactor

c. Raise core flow above 35 mlbs/hr

d. Commence a normal reactor shutdown

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BEACTpR OPERATOR Page 41 QUESTION: 067 (1.00)

A Loss Of Coolant Accident (LOCA) concurrent with a Loss Of Off Site Power and failure of DG 4 to start occurred on Unit 2 at 3:00 pm. At 3:08 pm the Shift supervisor directed the operator to place the Bus E2 to Bus E4 crosstie breaker control switch in "MAINT."

Which one of the following describes the operation of the Bus E2 to Bus E4 crosstie breakers with these conditions?

The Bus E2 to Bus E4 crosstie breakers may:

a. be closed ten minutes after the LOCA signal (3:10 pm).

b. be closed ten minutes after the control switch is placed in "MAINT" (3:18 pm).

c. not be closed unless the control switch is placed in either the "SBO" or " Fire" positions.

d. not be closed in accordance with current plant procedures.

QUESTION: 068 (1.00)

Which one of the following would be an indication of a loss of 4160V l Bus, E2? ,

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a. Partial Group 2, 6, and 8 isolations.

l b. Partial Group 1, 2, 3, 6, 8, and 10 isolations.

l l c. Two diesel generators start.

l d. Automatic start of DC lube oil pumps.

A A REACTpR OPERATOR Page 42 QUESTION: 069 (1.00)

Following a reactor scram on Unit 2 the operator can NOT determine that reactor power is less than 3%.

Which one of the following immediate operator actions for a reactor scram is ONLY performed on Unit 2?

a. VERIFY or PLACE the Feedwater/ Level Controller setpoint to +170 inches.

b. WHEN reactor power is below the APRM downscale setpoint, TRIP the Main Turbine.

c. When steam flow is less than 3 millon lb/hr, PLACE the Reactor MODE Switch to SHUTDOWN.

d. WHEN Reactor Vessel level is above + 170 inches AND increasing, IF two Reactor Feed Pumps are running, THEN trip one.

QUESTION: 070 (1.00)

During a transient, RPV pressure increased to 1110 psig and started to fall due to the opening of SRVs. RPV level has remained in the normal operating range and no operator action has been taken.

Which one of the following is the number of SRVs which have automatically opened?

a. Two

b. Three

c. Four

d. Five

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REACTpR OPERATOR Page 43 QUESTION: 071 (1.00)

Which one of the following is a direct result of excessive steam l carryunder?

Excessive steam carryunder will:

l l a. decrease the amount of feedwater preheating.

b. increase moisture carryover.

c. decrease the margin to the MCFR core thermal limit.

d. decrease the turbine efficiency.

QUESTION: 072 (1.00)

Which one of the following is the purpose of the 45% demand, recirculation flow control speed limiter #2? :

Speed limiter #2:

a. prevents running the recirculation pump at high speed with the discharge valve only partially open.

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b. provides NPSH protection for the recirculation jet pumps at i low feedwater flow.

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c. limits reactor power so that the feedwater system will be able to maintain or recover reactor water level on loss of a reactor feed pump. 3 i

d. limits recirculation flow such that possible operation in the )

I areas of instability is reduced.

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BEACTpR OPERATOR Page 44 QUESTION: 073 (1.00)

During operation at 100% power a gross failure of both seals on recirculation pump "B" occurs.

Which one of the following is the amount and type of drywell leakage?

a. 55-65 gpm of Unidentified leakage

b. 55-65 gpm of Identified leakage

c. 110-130 gpm of Unidentified leakage

d. 110-130 gpm of Identified leakage QUESTION: 074 (1.00)

When controlling primary containment pressure below 1.8 psig, EOP-02

" Primary Containment Control Procedure," directs the operator to " VENT THE DRYWELL USING SBGT (OP-10) AS REQUIRED."

Which one of the following describes the vent flowpath specified in i OP-10?

a. Suppression chamber via the 2 inch line utilizing the 1/2 inch l post LOCA inlets to SBGT.

b. Suppression chamber via the 20 inch line utilizing the normal inlet (F) damper to SBGT.

c. Drywell via the 18 inch line utilizing the normal inlet (F) damper to SBGT.

d. Drywell through the 2 inch bypass line and suppression chamber through the 2 inch line, utilizing the normal inlet (F) damper to SBGT.

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w Page 45 SEACTOR OPERATOR QUESTION: 075 (1.00)

Unit 1 is operating at 75% power with the following conditions:

-- Suppression pool water temperature is 85 degrees F with one loop of suppression pool cooling in service

-- Safety relief valve (SRV) operability surveillance is in progress

-- Following operation of SRV "L," there is indication that it is stuck open Which one of the following criteria requires the operator to manually scram the reactor.

a. Suppression pool temperature continues to increase with all available cooling in service.

b. As soon as it is recognized the SRV will not close.

c. The safety relief valve has been open fcr 2 minutes.

d. Suppression pool average water temperature increases to 105 degrees F.

QUESTION: 076 (1.00)

Which one of the following initial plant conditions would result in the worst case, rod drop accident?

a. Between 10% and 25% reactor power.

b. Between 29% and 64% reactor power.

c. Between 50% rod density and 10% reactor power. )

d. Less than 50% rod density.

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O O w w BEACTpR OPERATOR Page 46 i

QUESTION: 077 (1.00) ,

Following an incomplete reactor scram the Shift Supervisor has entered EOP-01-LEP-02, " Alternate Control Rod Insertion" to insert rods. 1 Which one of the following alternate rod insertion methods require a reset of the reactor protection system?

Control rod insertion by:

a. de-energizing the scram pilot valve solenoids.

b. venting the scram air header.

c. using the reactor manual control system.

d. increasing cooling water header pressure.

QUESTION: 078 (1.00)

AOP-32.0, " Plant Shutdown From Outside Control Room," cautions the operator to minimize throttling actions on DC motor-operated valves.

Which one of the following describes the reason for this caution?

Throttling actions on DC motor-operated valves should be minimized to:

a. prevent exceeding the duty cycle on the motors,

b. prevent excessive drain on the vital batteries.

c. limit valve wear from throttling actions.

d. limit electrical arcing in an explosive gas atmosphere, which may be present.

O O REACTpR OPERATOR Page 47 QUESTION: 079 (1.00) ;

Given the following plant conditions:

-- Reactor has been shutdown for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

-- Control rod 13-12 indicates notch "00" but does not indicate a full in, green light

-- SRMs A and C are reading 2 cps

-- SRMs B and D are reading 5 and 8 counts respectively

-- Fuel pool level is 23 feet 2 inches above the fuel assemblies seated in the spent fuel storage rack With the above conditions which one of the following would prevent l removal of spent fuel from the reactor core, either procedurally or because of a refueling interlock?

a. The level of water above the fuel assemblies when seated in the spent fuel storage rack is not high enough.

b. The required number of SRM channels are not available per Technical Specifications.

c. The failure of rod 13-12 to generate a rod full in signal to RMCS.

d. The required time before spent fuel can be removed from the reactor core has not elapsed.

QUESTION: 080 (1.00)

Given the following plant conditions on Unit 2:

1

-- Refueling operations are in progress l

-- A spent fuel bundle is dropped and damaged (

-- AREA RAD REFUEL FLOOR HIGH annunciator is alarming I l

Which one of the following is a required immediate operator action?

a. Start the standby gas treatment system.

b. Order evacuation of the Reactor Building.

c. Isolate Reactor Building ventilation.

d. Order all fuel movement stopped.

I I

a, REACTOR OPERATOR

e. Page 48 !

QUESTION: 081 (1.00)

EOP-01-SEP-01, " Primary Containment Venting," directs venting of primary containment via the suppression chamber, if drywell pressure cannot be maintained below 70 psig, or hydrogen concentration reaches 1%.

Which one of the following is the reason that this path is the preferred method to vent the primary containment under these conditions?

Venting from the suppression chamber:

a. condenses any steam from the primary containment.

b. prevents excessive drywell-to-torus differential pressure.

c. utilizes the scrubbing action of the to) 'o remove iodine.

d. dilutes the explosive hydrogen concentru . n.

QUESTION: 082 (1.00)

At the end of a Unit 2 operating cycle with the nlant in coastdown and all rods out, a transient results in a peak re ur pressure of 1130 psig for 5 seconds.

Which one of the following describes the affect the ATWS-RPT will have on the reactor recirculation pumps?

The ATWS-RPT will:

a. not actuate unless RPV LL No. 2 is present at the same time,

b. actuate and trip the recirculation MG set drive motor breaker.

c. actuate and trip the recirculation MG set field breaker

d. actuate and trip the two breakers installed in series between the generator and the recirculation pump motor.

a . .

A e

REACTOR OPERATOR Page 49 QUESTION: 083 (1.00)

During operation at 80% power on Unit 2, the "A" Reactor Recirculation Pump trips.

The following plant conditions exist following the pump trip: 1 1

-- The reactor is in region B on the thermal power limitations map j

-- Multiple LPRM upscale /downscale alarms are periodically annunciating and clearing ;

-- APRM power oscillations are 6% peak to peak on two channels and ;

7% on four channels !

Which one of the following is an immediate operator action with these conditions?

a. Increase core flow to greater than 35 mlbs/hr using the "B" :

reactor recirculation pump. I

b. Decrease core flow to less than 35 mlbs/hr using the "B" reactor l

recirculation pump.

c. Reduce reactor power to a level compatible with core flow by 1 inserting control rods. !

d. Insert a manual reactor scram.

QUESTION: 084 (1.00)

Which one of the following is an indication of a jet pump failure during operation at 100% power?

a. Decrease in indicated total core flow.

b. Core plate differential pressure decrease.

c. Recirculation loop flow decrease in the loop with the failed jet pump.

d. Recirculation pump discharge pressure increase on the loop with the failed jet pump.

e , ,

O O REACT,OR OPERATOR Page 50 QUESTION: 085 (1.00)

Following a loss of main condenser vacuum the bypass valves are being used to lower reactor pressure.

Which one of the following vacuum readings corresponds to the lowest value at which the bypass valves will still be effective in reducing reactor pressure? (Consider ONLY actual plant setpoints per AOP-37.0 for your answer)

a. 19 inches Hg

b. 10 inches Hg

c. 7 inches Hg

d. O inches Hg QUESTION: 086 (1.00)

Which one of the following describes the function of the low condenser vacuum bypass switches?

a. Bypass the low vacuum turbine trip while reducing generator ,

output following loss of a circulating water pump.

1

b. Bypass the bypass valve low vacuum closure, during ATWS l conditions. I

c. Allow operation of the mechanical vacuum pump during a loss of condenser vacuum transient. ,

1

d. Allow operation of the steam seals when establishing initial I condenser vacuum using the mechanical vacuum pump. l

REACTOR OPERATOR Page 51 QUESTION: 087 (1.00) i Which one of the following describes the operational impact a loss of 125V DC control power has on a 4160V AC breaker?

A 4160V AC breaker that has lost control power will:

a. not trip on a fault and can not be tripped from the control room.

b. not trip on a fault but can be tripped from the control room.

c. trip on a fault but can not be tripped from the control room.

d. trip on a fault and can be tripped from the control room.

QUESTION: 088 (1.00)

Given the following plant conditions:

-- Reactor has scrammed

-- RCIC is operating maintaining vessel level

-- HPCI was secured by the operator Which one of the following describes the affect a loss of Division I, 125V DC will have on RCIC?

Following a loss of Division I 125V DC, the RCIC:

a. Turbine Steam Supply Valve, (E51-F045) will not close on a high vessel level condition.

b. turbine will trip on overspeed.

c. turbine speed will decrease to approximately 2000 RPM.

d. system inboard isolation logic will actuate.

j

e , ,

e a Page 52 EEACTpR OPERATOR QUESTION: 089 (1.00)

Fhich one of the following describes why EOP-01-RSP, " Reactor Scram Frocedure," directs the operator to " TRIP TURBINE" instead of letting it trip automatically on reverse power?

The automatic reverse power trip:

a. will result in excessive arcing in the main generator output breakers.

b. may cause a pressure spike sufficient to rupture the LP turbine relief diaphragas.

c. will result in an automatic cold start of the diesel generators.

d. may place an unnecessary load on the main turbine thrust bearing.

QUESTION: 090 (1.00)

Which one of the following limits the rate at which reactor vessel level can be lowered using reactor water cleanup reject at rated temperature and pressure.

a. Regenerative heat exchanger cooldown rate.

b. Nonregenerative heat exchanger outlet temperature.

c. Temperature change limit in the reactor vessel drain line.

d. Pressure downstream of reject Flow Control Valve, G33-F033.

e w

a w

i i

EEACTOR OPERATOR Page 53 l I

I i

QUESTION: 091 (1.00) 1 Which one of the following is one of the basis for the DRYWELL SPRAY l INITIATION LIMIT?

Initiation of drywell sprays while in the " SAFE" area of the DRYWELL SPRAY INITIATION LIMIT prevents:

a. excessive cycling of the primary containment vacuum breakers.

b. de-inertion of the containment.

c. chugging in the SRV tailpipes.

d. drawing suppression pool water up the SRV tail pipes.

QUESTION: 092 (1.00)

Which one of the following cooling water system conditions requires a manual reactor scram if pressure cannot be im:nediately restored?

a. Conventional Service Water pressure is steady at 36 psig with one TBCCW Service Water Supply Valve (SW-V4) in its throttled position.

b. Nuclear Service Water pressure is 25 psig and decreasing and Diesel Generator Building service water header was left isolated.

c. All Turbine Building Closed Cooling Water system pumps are running and pressure is 45 psig.

d. All Reactor Building Closed Cooling Water system pumps are running and pressure is 65 psig and decreasing.

o .

A e REACTpR OPERATOR Page 54 QUESTION: 093 (1.00)

Given the following plant conditions:

-- A loss of off-site power has occurred

-- RBCCW PUMP DISCH HDR PRESS LOW annunciator is alarming

-- Discharge header pressure indicates 30 psig Which one of the following describes the automatic restart feature (s) of the RBCCW pumps as the diesel generators energize the E buses.

The RBCCW pumps will:

a. start only if their control switch is in AUTO.

b. start only if their control switch is in ON.

c. start if their control switch is in AUTO or ON.

d. not restart unless their control switch is placed to OFF then to AUTO or ON.

QUESTION: 094 (2.00)

Match the pneumatic (air / nitrogen) system failure automatic actions from Column A with the actuation pressure listed in Column B.

The items from column B may be used once, more than once, or not at all and only a single answer may occupy one answer space.

COLUMN A COLUMN B AUTOMATIC ACTIONS ACTUATION PRESSURE

a. Air Compressors A, B and C start 1. 85 psig

b. Service Air Isolation Valves, 2. 95 psig PV-706-1 and PV-706-2 close

3. 100 psig

c. Standby Reactor Building Air Compressors start 4. 105 psig

d. Backup Nitrogen Rack Isolation Valves, 5. 107 psig RNA-SV-5481 and RNA-SV-5482 open

6. 109 psig

7. 110.5 psig l l

l

O O REACTOR OPERATOR Page 55 QUESTION: 095 (1.00)

Which one of the following pairs of Group 1 isolation signals will cause the Recirculation Sample Valves, B32-F019 and B32-F020 to close?

a. Reactor Low Level #3, Main Steam Line High Flow

b. Condenser Vacuum, Main Steam Line Low Pressure

c. Turbine Building Area Temperature High, Main Steam Line Tunnel Temperature High

d. Reactor Low Level #3, Main Steam Line High Radiation QUESTION: 096 (1.00)

During a plant startup and heatup a loss of both CRD pumps occurs.

Which one of the following is the MINIMUM reactor pressure allowed before a manual reactor scram is required?

a. 700 psig

b. 750 psig

c. 800 psig

d. 850 psig

o .

O w

REACTOR OPERATOR Page 56 QUESTION: 097 (1.00)

Unit 2 was operating at 100% power when a LOCA occurred.

The following plant conditions exist:

-- Drywell reference leg area temperature is 375 degrees F

-- Drywell pressure is 5 psig

-- Reactor pressure is 100 psig Which one of the following reactor water level instruments would indicate a valid level? [ Caution 1, Reactor Water Level is available]

a. Fuel zone (NO36/37) with a reading of greater than +150 inches

b. Narrow range (N004A/B/C) with a reading of +152 inches

c. Wide range (N026A/B) with a reading of +152 inches

d. Shutdown range (N027A/B) with a reading of +152 inches QUESTION: 098 (1.00)

The temperature control leg of EOP-03, " Secondary Containment Control Procedure" asks: "IS A PRIMARY SYSTEM DISCHARGING INTO THE REACTOR BUILDING?"

Which one of the following is a primary system as referenced in this step?

A Primary System is:

a. any plant safety-related system required to be operable in Modes 1, 2 and/or 3.

b. a system whose leak rate will decrease as reactor pressure decreases.

c. a system required to shutdown the reactor or provide long-term core cooling.

d. any plant system whose pressure will change with drywell and/or torus pressure.

e',s b b EEACTOR OPERATOR Page 57 QUESTION: 099 (1.00)

Regarding EOP-03-SCCP, " Secondary Containment Control Procedure."

Which one of the following is the significance of exceeding a Maximum Normal Operating Value?

a. The "EQ" Envelope for operating equipment in the area will be exceeded.

b. Personnel access necessary for safe shutdown of the plant will be prevented.

c. Operating systems in the affected area will exceed the expected normal operating parameters. l

d. Equipment necessary for safe shutdown of the plant will fail.

l 1

I

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

a CD REACTpR OPERATOR Page 58

'vi431tB COFY ANSWER: 001 (1.00) c.

REFERENCE:

OAP-010 Rev. O page 10 LOI-CLS-LP-201-C LO 06 Facility Exam Bank, Record Sheet 4534

[4.2/4.2]

294001A102 ..(KA's)

ANSWER: 002 (1.00) c.

REFERENCE:

AI-110 Rev. 5 page 4

[3.9/4.5]

294001K102 ..(KA's)

ANSWER: 003 (1.00) c.

REFERENCE:

OAI 58 page 17 Lesson Plan 1: LOI-CLS-LP-201-C Objective: 01F

[3.9/4.5]

294001K102 ..(KA's)

ANSWER: 004 (1.00) a.

I u,* 4 h by l\Vilij!!fi u Jt a

't 8 )

EEACTpR OPERATOR Page 59

REFERENCE:

Plant Operating Manual Volume XXII OPLP-22 Rev. 2 page 6 (Book 33)

[3.9/4.5]

294001K102 ..(KA's) l l

ANSWER: 005 (1.00) l

a. '

REFERENCE:

001-04 Rev. 53 Page 8

[3.3/4.3]

l 294001A111 ..(KA's)

ANSWER: 006 (1.00) c.

REFERENCE:

00I-05 Rev. 27 page 8

[4.5/4.3]

294001A113 ..(KA's)

ANSWER: 007 (1.00) d.

REFERENCE:

001-13 Rev. 36 page 1

[3.7/3.7]

I 294001K101 ..(KA's)

O )

REACTOR OPERATOR Page 60 ANSWER: 008 (1.00) a.

REFERENCE:

OPS-CLS-SM-OSSA Rev. O page 8

[3.2/3.4]

294001A115 ..(KA's)

ANSWER: 009 (1.00) d.

REFERENCE:

AOP-26.0 Rev.7 page 5

[2.9/3.4]

294001A114 ..(KA's)

ANSWER: 010 (1.00) c.

REFERENCE:

10 CFR 20.1003 page 374

[2.8/3.4]

294001K103 ..(KA's)

ANSWER: 011 (1.00) b.

l 1

1

g -

O REACTOR OPERATOR Page 61

REFERENCE:

10 CFR 20.1201 page 377

[3.3/3.8]

294001K103 ..(KA's)

ANSWER: 012 (1.00) b.

REFERENCE:

10 CFR 26.20 (a) (1) page 482

[2.7/3.7]

294001A103 ..(KA's)

ANSWER: 013 (1.00) a.

REFERENCE:

001-05 page 11 1

[4.5/4.3] l 294001A113 ..(KA's) l l

l ANSWER: 014 (1.00) d.

REFERENCE:

SM-008-B Rev. O page 3 LO 6.a

[2.8/2.8]

201001A308 ..(KA's)

$ )

REACTOR OPERATOR Page 62 ANSWER: 015 (1.00) b.

REFERENCE:

SM-008-B Rev. O page 7 LO 3.d

[2.8/2.8]

201001K110 ..(KA's)

ANSWER: 016 (1.00) d.

REFERENCE:

SM-007- A pages 14 and 21 LO 10 d. OP-7 page 28

[3.2/3.1]

201002A301 ..(KA's)

ANSWER: 017 (1.00) c.

REFERENCE:

APP A-06 5-1 page 60 SM-002-A LO 30 g.

[3.2/3.3]

202002K305 ..(KA's)

ANSWER: 018 (1.00) d.

A a,- ,

REACTOR OPERATOR Page 63

REFERENCE:

SM-002-A page 28 LO 17 d.

[3.0/3.0]

202002K402 ..(KA's)

ANSWER: 019 (1.00) a.

REFERENCE:

SM-017-A page 4 LO 17 a.

[3.5/3.5]

203000K201 ..(KA's)

ANSWER: 020 (1.00) d.

REFERENCE:

SM-017-A page 8 LO 10

[4.2/4.2]

203000K401 ..(KA's)

ANSWER: 021 (1.00) d.

REFERENCE:

SM-019A page 32 LO 3 n. APP-011-1 pages 5, 41, and 62

[Recent Plant Modification]

[4.3/4.3]

206000K407 ..(KA's)

O h REACTOR OPERATOR Page 64 ANSWER: 022 (1.00) d.

REFERENCE:

VI/0EOP-01-UG Rev. 20 page 23

[4.2/4.0]

206000G013 ..(KA's)

ANSWER: 023 (1.00) c.

REFERENCE:

SM-019-A pages 13 and 14 LOs 3.o. and 3.p.

[3.7/3.8]

206000K419 ..(KA's)

ANSWER: 024 (1.00) b.

REFERENCE:

SM-018-A page 7 LO 4.b.

[3.8/3.6]

209001A401 ..(KA's) l ANSWER: 025 (1.00) I b.

l l

l 1

O, !

REACTOR OPERATOR Page 65

REFERENCE:

SM-018-A page 8 LO 13.a

[3.0/3.1]

209001K201 ..(KA's)

ANSWER: 026 (1.00) c.

REFERENCE:

LEP-03 Rev. 12 page 3 SM-005-A LO 8.f Facility Exam Bank Question Number, LLE0049 Modified

[4.3/4.4) 211000K301 ..(KA's)

ANSWER: 027 (1.00)

C.

REFERENCE:

SM-003-A Rev. 1, Page 27 LO 8

[4.0/4.1) 212000A216 ..(KA's)

ANSWER: 028 (1.00) d.

REFERENCE:

SM-003-A Rev.1 page 12 LOs 4.b, 4.e and 4.f

[3.4./3.4]

212000A108 ..(KA's)

i f-~s u.

REACTOR OPERATOR Page 66 l l

l ANSWER: 029 (1.00) c.

REFERENCE:

SM-009-B Rev. O page 9 LO 3.a

[3.0/3.1]

215003K503 ..(KA's)

ANSWER: 030 (1.00) d.

REFERENCE:

SM-009-A Rev.O page 12 LO 3.a

[3.2/3.2]

215004K406 ..(KA's)

ANSWER: 031 (1.00) a.

REFERENCE:

SM-009-A Rev. O page 15

[3.1/3.3) 215004K602 ..(KA's)

ANSWER: 032 (1.00) a.

O O

mv BEACTDR OPERATGR Page 67

REFERENCE:

SM-009 Rev. O page 2 LO 4.g

[4.0/4.0]

215005K301 ..(KA's)

ANSWER: 033 (1.00) c.

REFERENCE:

BSEP/Vol. II/SD-09 page 68

[2.9/3.2]

215005K504 ..(KA's)

ANSWER: 034 (1 00) d.

REFERENCE:

SM-118-A Rev. O page 9 LO 5.1

[3.1/3.2]

216000K501 ..(KA's)

ANSWER: 035 (1.00) d.

REFERENCE:

SM-016-A page 38 LO 4.b

[3.4/3.3]

217000A203 ..(KA's)

l O O l REACTOR OPERATOR Page 68 l

l AESWER: 036 (1.00) l I

b. j

REFERENCE:

1 SM-020-A Rev.0 page 15 LO 16.c Facility Exam Bank, Question Number LLA0113 Modified

[3.9/4.1]

218000K601 ..(KA's)

ANSWER: 037 (1.00)

C.

REFERENCE:

SM-020-A Rev. O page 6 LO 8

[3.4/3.6]

218000A203 ..(KA's)

ANSWER: 038 (1.00) d.

REFERENCE:

SM-032-C Rev. 2 page 21 LO 6.a

[3.8/3.8]

259002A101 ..(KA's)

ANSWER: 039 (1.00) d.

O bb mv REACTOR OPERhTOR Page 69

REFERENCE:

SM-032-C Rev. 2 page 47 LO 5

[3.G/3.O]

259002A306 ..(KA's)

ANSWER: 040 (1.00) d.

REFERENCE:

SM-010-A Rev. 2 page 24 LO 9.g

[3.1/3.2]

261000K605 ..(KA's)

ANSWER: 041 (1.00) a.

REFERENCE:

SM-039-A page 72 LOs 3 and 4

[3.8/3.7]

264000K408 ..(KA's)

ANSWER: 042 (1.00) c.

REFERENCE:

SM-008-A page 18 LO 6.d

[3.6/3.7]

201003K404 ..(KA's) l l

I i

A A REACTOR OPERATOR Page 70 ANSWER: 043 (1.00) d.

REFERENCE:

SSM 27-2B RWM page 23 LO 7.a Facility Exam Bank, Question Number LLN0171

[3.2/3.1]

201006A301 ..(KA's)

ANSWER: 044 (1.00) c.

REFERENCE:

SSM 27-2B RWM page 23 LO 7.a Facility Exam Bank, Question Number LLN0007

[3.2/3.2]

201006A405 ..(KA's) l ANSWER: 045 (1.00) a.

REFERENCE:

SM-002-A Rev. O page 21 LO 8 l

[3.6/3.7]

202001K101 ..(KA's) l l

ANSWER: 046 (1.00) 1

b. j i

l

Ow O

w REACTOR OPERATOR Page 71

REFERENCE:

SM-002-A Rev. O page 12 LO 5 i

[3.9/3.9]

202001K401 ..(KA's)

ANSWER: 047 (1.00) d.

l

REFERENCE:

SM-014-A Rev. O page 32 LO 9.d

[3.7/3.8) 204000K108 ..(KA's)

ANSWER: 048 (1.00) a.

REFERENCE:

1 1

20P-17 Rev .106 page 32 Section 5.4.2 caution 64/65 Facility Exam Bank, '

l Question Number LLN0096 Modified

[3.2/3.3]

205000G010 ..(KA's)

ANSWER: 049 (1.00) a.

l

REFERENCE:

l SM-007-A Rev. O page 7 LO 3 l

[3.3/3.3) I l

214000K105 ..(KA's) l l

O O REACTOR OPERATOR Page 72 ANSWER: 050 (1.00) d.

REFERENCE:

SM-009-E Rev. 01 page 17 LO 4

[3.1/3.3]

215002A203 ..(KA's)

ANSWER: 051 (1.00) d.

REFERENCE:

SM-009-E Rev. 01 page 4 LO 7.c

[2.8/3.1]

215002K605 ..(KA's)

ANSWER: 052 (1.00) b.

REFERENCE:

SM-017-A Rev. O page 25 LO 26

[3.5/3.4]

226001A403 ..(KA's)

ANSWER: 053 (1.00) d.

r-A w

A w

REACTOR OPERATOR Page 73

REFERENCE:

EOP-02 Primary Containment Control Steps PC/P-4 and 5 SM-017-A LO 26

[3.8/3.6]

230000A402 ..(KA's)

ANSWER: 054 (1.00) b.

REFERENCE:

SM-025-A Rev. O page 11 LO 9.b

[3.2/3.2]

239001K602 ..(KA's)

ANSWER: 055 (1.00) a.

REFERENCE:

SM-025-A Rev. O page 64 LO 6

[3.2/3.3]

239001K201 ..(KA's)

ANSWER: 056 (1.00) c.

REFERENCE:

SM-026-C Rev. 2 page 24 LO 15

[3.8/3.9]

245000A207 ..(KA's)

8 O REACTOR OPERATOR Page 74 ANSWER: 057 (1.00) d.

REFERENCE:

CLS-SM-032-A Rev. 2 page 81

[3.1/3.1]

256000K105 ..(KA's)

ANSWER: 058 (1.00) a.

REFERENCE:

SM-050B Rev. 2 page 24 LO 4

[3.4/3.5]

262001G007 ..(KA's)

ANSWER: 059 (1.00) a.

REFERENCE:

SM-030-A Rev. O pages 38 and 39

[3.3/3.3) 271000A301 ..(KA's)

ANSWER: 060 (1.00) l a.

l l

1

e e REACTOR OPERATOR Page 75

REFERENCE:

SM-003-A Rev. 1 fage 19 SM-030-A Rev. O page 21 SM-011-B Rev. 2 page 2 LO 8.a

[3.9/4.2]

239001A205 ..(KA's)

ANSWER: 061 (1.00) b.

REFERENCE:

SM-001-A Rev. O page 21 LO 7

[3.4/3.5]

290002K112 (KA's)

ANSWER: 062 (1.00) a.

REFERENCE:

SM-006-A Rev. O page 17 LO 3.b

[3.4/3.6]

268000A401 ..(KA's)

ANSWER: 063 (1.00) a.

REFERENCE:

SM-007-A pages 15, 16, 38 and 45 j 1

[3.3/4.1] )

1 234000K402 ..(KA's) l l

~ '

O O REACTOR OPERATOR Page 76 ANSWER: 064 (1.00) c.

REFERENCE:

SM-017-A Rev .0 page 13 LO 4

[2.9/3.0]

233000K102 ..(KA's)

ANSWER: 065 (1.00) d.

REFERENCE:

SM-300-E Rev. 2 page 9 step RC/L-12 LO 23 Facility Exam Bank, Question Number LLE0338 Modified

[4.1/4.3]

295037K101 ..(KA's)

ANSWER: 066 (1.00) b.

REFERENCE:

AOP-03 page 4 step 3.2(3) SM-002-A LO 39 Facility Exam Bank, Question Number LLA0115

[3.8/3.9]

295014K106 ..(KA's)

ANSWER: 067 (1.00) b.

}

A &

REACTOR OPERATOR Page 77

REFERENCE:

SM-050-B page 20 LO 13.c

[3.5/3.7) 295003A204 ..(KA's)

ANSWER: 068 (1.00) b.

REFERENCE:

AOP 36.1 Rev. 11 page 4 section 2.2 Facility Exam Bank, Question Number LLA0204 Modified

[3.7/3.7) 295003K306 ..(KA's)

ANSWER: 069 (1.00)

C.

REFERENCE:

SM-300-C Rev. 1 page 2 LO 2

[4.1/4.2]

295006G010 ..(KA's)

ANSWER: 070 (1.00) c.

REFERENCE:

SM-020-A page 7 LO 9

[3.9/4.1]

295007A104 ..(KA's)

\

1 g g REACTOR OPERATOR Page 78 ANSWER: 071 (1.00) c.

REFERENCE:

SM-001-A Rev. O page 12 LO 9

[2.7/2.9]

295009K101 ..(KA's)

ANSWER: 072 (1.00) c.

REFERENCE:

GM-002-A Rev. O page 25 LO 3.h

[3.2/3.3) 295009K301 ..(KA's)

ANSWER: 073 (1.00) a.

REFERENCE:

SM-00-A Rev. O pages 7 and 8 LO 15.c

[3.4/3.8]

295010A201 ..(KA's)

ANSWER: 074 (1.00) d.

& A REACTOR OPERATOR Page 79

REFERENCE:

2-OP-10 Rev. 49 page 22 SM-010-A page 12 LO 3

[3.1/3.4]

295010A105 ..(KA's)

ANSWER: 075 (1.00) b.

REFERENCE:

AOP 30.0 Rev. 3 page 3 SM-020-A Rev. O LO 22

[3.8/3.6]

295013G010 ..(KA's)

ANSWER: 076 (1.00) c.

REFERENCE:

AOP-02 . Rev. 5 page 4

[3.9/3.9]

295014K207 ..(KA's)

ANSWER: 077 (1.00) d.

REFERENCE:

EOP-01-LEP-02 Rev. 10 page 17 SM-300-J Rev. O LO 5.

[4.0/4.1]

295015K204 ..(KA's) j

e O

w REACTOR OPERATOR Page 80 ANSWER: 078 (1.00) a.

REFERENCE:

AOP-32.0 Rev. 24 page 5

[4.0/4.1]

295016K202 ..(KA's)

ANSWER: 079 (1.00)

c. o r h' db-t / / 9 /9 y

REFERENCE:

SM-305-A page 32 Rev. O LO 16.b

[3.2/3.4]

295023K204 ..(KA's)

ANSWER: 080 (1.00) d.

REFERENCE:

AOP-07.0 Rev .2 page 3 LO 31

[3.9/3.9]

295023G010 ..(KA's)

ANSWER: 081 (1.00)

C.

1 e a !

REACTOR OPERATOR Page 81 l l

l 1

REFERENCE:

SM-300-K Rev. O page 3 LO 3

[3.5/4.0]

295024K307 ..(KA's)

ANSWER: 082 (1.00) b.

REFERENCE:

SM-002-A page 33 Rev. O LO .19

[3.9/4.1]

295025K204 ..(KA's)

ANSWER: 083 (1.00) d.

REFERENCE:

AOP-04 Rev.9 page 4

[3.8/3.7) 295001G010 ..(KA's)

ANSWER: 084 (1.00) b.

REFERENCE:

AOP-4.4 Rev. 4 page 3

[3.4/3.4]

295001K207 ..(KA's)

e . a g g w w REACTOR OPERATOR Page 82 ANSWER: 085 (1.00) b.

REFERENCE:

AOP-37 Rev. 3 page 3 SM-025-A Rev.0 LO 9.f

[3.2/3.3]

295002K204 ..(KA's)

ANSWER: 086 (1.00) d.

REFERENCE:

SM-025-A page 18 LO 5.j

[3.4/3.4]

295002G006 ..(KA's)

ANSWER: 087 (1.00)

a. l

REFERENCE:

SM-50-B Rev. 2 page 23 LO 7 i

[3.3/3.4]

l 295004K105 ..(KA's) 1 ANSWER: 088 (1.00) 1 a.

i 1

1

y y REACTOR OPERATOR Page 83

REFERENCE:

SM-016-A Rev. 1 page 51 LO 8.c

[3.2/3.3) 295004A204 ..(KA's)

ANSWER: 089 (1.00) c.

REFERENCE:

SM-300-C page 5

[3.2/3.3]

295005K208 ..(KA's)

ANSWER: 090 (1.00) b.

REFERENCE:

20P-14 Rev. 88 page 15 CAUTION SM-014-A LO 9.e

[3.1/3.1]

295008K209 ..(KA's)

ANSWER: 091 (1.00) b.

REFERENCE:

SM-300-L page 56 LO 3.g

[3.3/3.5) 295012K101 ..(KA's)

e ~ s REACTOR OPERATOR Page 84 ANSWER: 092 (1.00) a.

REFERENCE:

AOP-19.0 Rev. 6 No Facility LO

[3.4/3.6]

295018K202 ..(KA's)

ANSWER: 093 (1.00) c.

REFERENCE:

i SM-021-A Rev. 1 page 7 LO 9.a Facility Exam Bank, Quection Number LLN0081 Modified

)

[2.9/2.9]

295018A205 ..(KA's)

ANSWER: 094 (2.00)

a. 7

b. 4

c. 2

d. 2

REFERENCE:

AOP-20.0 Rev. 15 SM-046-A LO 7

[3.5/3.4]

295019K302 ..(KA's)

a O \

REACTOR OPERATOR Page 85 )

1 ANSWER: 095 (1.00) d.

REFERENCE:

SM-012-A Rev. O page 9 LO 6

[3.4/3.8]

295020A206 ..(KA's)

ANSWER: 096 (1.00) c.

REFERENCE:

AOP-02.1 Rev. 4 page 3

[3.7/3.9]

295022K301 ..(KA's)

ANSWER: 097 (1.00) c.

REFERENCE:

Unit 2 Caution 1 Chart

[3.7/3.9]

295028A203 ..(KA's)

ANSWER: 098 (1.00) b.

C "a g

REACTOR OPERATOR Page 86

REFERENCE:

SM-300-M page 11 LO 4.a

[3.7/3.9]

295032A105 ..(KA's)

ANSWER: 099 (1.00) c.

REFERENCE:

SM-300-M page 2 LO 4.c Facility Exam Bank, Question Number LLE0291

[3.0/3.2]

295036A201 ..(KA's)

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

r~s

. L. J

. TEST CROSS REFERENCE Page 1 RO Exam BWR Reactor Organized by Quest i on Number QUESTION VALUE REFERENCE 001 1.00 9001011

, j["#h' 002 1.00 9001012 f . ;. 0 i 003 1.00 9001013 i'+s v: t- k 6 UUk 004 1.00 9001014 005 1.00 9001016 006 1.00 9001017 007 1.00 9001018 008 1.00 9001020 009 1.00 90C1021 010 1.00 9001023 011 1.00 9001024 012 1.00 9001025 013 1.00 9001028 014 1.00 9001030 015 1.00 22022 016 1.00 9001033 1 l

017 1.00 22031 018 1.00 15478 019 1.00 9001037 1 020 1.00 9001039 l 021 1.00 20144 022 1.00 9001041 023 1.00 9001042 024 1.00 22038 1 025 1.00 9001045 !

026 1.00 9001046 027 1.00 22041 028 1.00 9001049 029 1.00 9001050 030 1.00 9001051 031 1.00 25102 032 1.00 9001053 033 1.00 9001054 l 034 1.00 9001055 035 1.00 9001056 036 1.00 9001057 037 1.00 9001058 038 1.00 9001059 039 1.00 9001060 040 1.00 9001061 041 1.00 9001062 042 1.00 22023 043 1.00 9001064 044 1.00 9001065 045 1.00 9001066 -

046 1.00 9001067 ,

! ' 'i f

~

047 1.00 9001069 048 1.00 9001070 049 1.00 20132 i

- A S

- TEST CROSS REFERENCE Page 2 l

RO Exam BWR Reactor Organized by Question Number QUESTION VALUE REFERENCE 050 1.00 9001074 051 1.00 9001075 f * -

., 052 1.00 9001076

, 7 053 1.00 9001078 a

054 1.00 9001079 055 1.00 9001080 056 1.00 9001081 057 1.00 9001082 058 1.00 22073 059 1.00 9001084 060 1.00 9001086 061 1.00 9001087 062 1.00 9001088 063 1.00 9001090 064 1.00 20153 065 1.00 9001092 066 1.00 9001093 067 1.00 9001095 068 1.00 9001096 069 1.00 9001098 070 1.00 9001099 071 1.00 9001101 072 1.00 25081 073 1.00 9001103 074 1.00 9001104 075 1.00 9001105 076 1.00 9001107 077 1.00 9001109 078 1.00 9001111 079 1.00 9001112 080 1.00 9001113 081 1.00 20174 082 1.00 9001117 083 1.00 9001121 084 1.00 25147 085 1.00 9001123 086 1.00 9001124 087 1.00 9001125 088 1.00 9001126 089 1.00 25171 090 1.00 9001128 091 1.00 9001129 092 1.00 22093 093 1.00 9001131 094 2.00 9001132 " ~

095 1.00 9001133 096 1.00 9001134 097 1.00 20088 098 1.00 22119

Ow h -

. TEST CROSS REFERENCE Page 3 RO Exam BWR Reactor j l

Organized by Question Number j i

QUESTION VALUE REFERENCE 099 1.00 9001140 100.00

,' a ,

+

~; r y . ------

t),, i ,e , -> :; ,. i) \_ ------

p ,

!. i v .h >

100.00 i 'l ,

' , ('

J

O O

. TEST CROSS REFERENCE Page 4 RO Exam BWR Reactor Organized by KA Group PLANT WIDE GENERICS QUESTION VALUE KA 001 1.00 294001A102 012 1.00 294001A103 005 1.00 294001A111 7' e 1.00 294001A113 J*006 E

'js

! - 57 013 1.00 294001A113

'" ' ' ' ' -iJ ! 009 1.00 294001A114 008 1.00 294001A115 007 1.00 294001K101 003 1.00 294001K102 002 1.00 294001K102 004 1.00 294001K102 010 1.00 294001K103 011 1.00 294001K103 PWG Total 13.00 PLANT SYSTEMS Group I QUESTION VALUE KA 014 1.00 201001A308 015 1.00 201001K110 016 1.00 201002A301 017 1.00 202002K305 018 1.00 202002K402 019 1.00 203000K201 020 1.00 203000K401 022 1.00 206000G013 021 1.00 206000K407 023 1.00 206000K419 024 1.00 209001A401 025 1.00 209001K201 026 1.00 211000K301 028 1.00 212000A108 027 1.00 212000A216 029 1.00 215003K503 030 1.00 215004K406 031 1.00 215004K602 032 1.00 215005K301 033 1.00 215005K504 "

034 1.00 216000K501 035 1.00 217000A203 037 1.00 218000A203 036 1.00 218000K601

a A i l

. TEST CROSS REFERENCE Page 5 RO Exam BWR Reactor Organized by KA Group PLANT SYSTEMS ,, ,

?1h\ i Group I l / {' g ,. ! - g, I g, q, , .(

QUESTION VALUE KA 038 1.00 259002A101 039 1.00 259002A306 040 1.00 261000K605 041 1.00 264000K408 PS-I Total 28.00 Group II QUESTION VALUE _

KA 042 1.00 ;!01003K4 04 043 1.00 01006 A3 01 044 1.00 201006A405 045 1.00 202001K101 046 1.00 202001K401 047 1.00 204000K108 048 1.00 205000G010 049 1.00 214000K105 050 1.00 215002A203 051 1.00 215002K605 052 1.00 226001A403 053 1.00 230000A402 060 1.00 239001A205 055 1.00 239001K201 054 1.00 239001K602 056 1.00 245000A207 057 1.00 256000K105 058 1.00 262001G007 059 1.00 271000A301 PS-II Total 19.00 Group III QUESTION VALUE KA 064 1.00 233000K102 063 1.00 234000K402 i 062 1.00 268000A401 ^" l 061 1.00 290002K112 +

PS-III Total 4.00 ;

l I

- 8 h

. TEST CROSS REFERENCE Page 6 RO Exam BWR Reactor Organized by KA Group PLANT SYSTEMS QUESTION VALUE KA PS Total 51.00 EMERGENCY PLANT EVOLUTIONS Aj, e

": q r ' , ;e .

Group I 2 I'Vi f QUESTION VALUE KA 089 1.00 295005K208 069 1.00 295006G010 070 1.00 295007A104 071 1.00 295009K101 072 1.00 295009K301 074 1.00 295010A105 073 1.00 295010A201 066 1.00 295014K106 076 1.00 295014K207 077 1.00 295015K204 081 1.00 295024K307 082 1.00 295025K204 065 1.00 295037K101 EPE-I Total 13.00 Group II QUESTION VALUE KA 083 1.00 295001G010 084 1.00 295001K207 086 1.00 295002G006 085 1.00 295002K204 067 1.00 295003A204 068 1.00 295003K306 088 1.00 295004A204 087 1.00 295004K105 090 1.00 295008K209 091 1.00 295012K101 075 1.00 295013G010 078 1.00 295016K202 093 1.00 295018A205 092 1.00 295018K202 '

I 094 2.00 295019K302 095 1.00 295020A206 096 1.00 295022K301 097 1.00 295028A203

O h

. . TEST CROSS REFERENCE Page 7 RO Exam BWR Reactor Organized by KA Group EMERGENCY PLANT EVOLUTIONS

&/Acor4 f .

Group II l'*i ; !j % l , ,

. ,t QUESTION VALUE KA EPE-II Total 19.00 Group III QUESTION VALUE KA 080 1.00 295023G010 079 1.00 295023K204 098 1.00 295032A105 099 1.00 295036A201 EPE-III Total 4.00 !

l EPE Total 36.00

______ l Test Total 100.00 s

3 h-Ik' 5.0 startup 5.4 Shutdown Cooline Mode (Continued) .

Loop 5.4.2 Procedural Steps l

NOTE: RHR System piping should be flushed prior to cooling the reactor below a pressure of 60 psig; otherwise, there will NOT be enough driving force to completely warm and flush the piping.

If this procedure was entered from Section 8.3, and RPV l 1.

pressure is below 6f psig, then perform the following:

a. N/A Steps 5.4.2.2 through 5.4.2.53 of this section.

b. Continue in this procedure at Step 5.4.2.54

2. Lower suppression pool level to -31 inches per Section [

I 8.7 of this procedure and return to Step 5.4.2.3 of this section.

Select RRR loop to be placed in shutdown cooling and j 3.

place circuit breaker switches to the ON position for I the selected RHR loop valves:

B RHR Loop A RHR Loop Ell-F006B MCC 2XS compt DL1 E11-F006A MCC 2XA Compt DE9 Ell-F006D MCC 2XB Compt DL2 Ell-F006C MCC 2XA Compt DF0 Ell-F011B MCC 2XB Compt DL6 Ell-F009 MCC 2XA Compt DH3 E11-V33 MCC 2XB Compt DM9 Ell-V32 MCC 2XA Compt DC3 Ell-F009 MCC 2XA Compt DH3

4. Open Loop A(B) LPCI Injection Check Valve Bypass Valve, l E11-V32(V33).

g

5. Open Loop A(B) LPCI Inboard Injection Valve ,

Ell-F015A(B). I Verify open Loop A(B) LPCI Outboard Injection Valve, g 6.

Ell-F017A(B). i l

7. Close RRR HX 2A(2B) Outlet Valve, Ell-F003A(B). I

8. Verify the following valves open: l g

l

a. RHR HX 2A(2B) Bypass Valve, Ell-F048A(B).

b. RHR llX 2A(2B) Inlet Valve, Ell-F047A(B). l Page 23 of 170 l 20P-17 Rev. 106

a w

O

+

5.0 STARTUP 5.4 Shutdown Cooline Mode (Continued) .

5.4.2 Procedural Steps (Continued)

9. Verify RHR RX 2A(2B) RHR to Torus Drain Controller, l Ell-HC-604A(B), is set to zero output.

10. Open RHR HX 2A(2B) Drain To Suppression Pool Valve, g Ell-F0llA(B). I

11. Verify RHR HX 2A(2B) Drain To RCIC Valve, Ell-F026A(B), -

is closed.

12. Place Control Switch, Ell-S41A(B), to the OPEN position to energize the Air Supply to E11-F053A(B) RRR to Torus Drain Valve.

13. Monitor Suppression Pool temperature during the flush to the Suppression Pool.

14 If Suppression Pool temperature exceeds 95 deg F, then l perform the following:

a. Go To E0P-02-PCCP and execute it concurrently with this procedure.

b. Log Suppression Pool temperature at one-hour inte rvals .

CAUTION Water can be drained faster through Ell-CV-F053A(B) than can be made up I

through E11-V32(V33). If RHR pressure drops too far below RPV pressure, i

water hammer may occur. If RHR pressure drops below 40 psig, adequate RHR loop sample flow may NOT be available at the sample station. g

15. With RHR to Torus Drain Controller, Ell-HC-604A(B) in MANUAL, SLOWIN throttle open RHR to Torus Drain Control Valve, E11-CV-F053A(B), to flush RHR HX downstream l piping to the suppression pool.

16. Maintain RPV water level between 182 and 192 inches [

during the flush to the suppression pool as follows: I

a. If-RPV water level CANNOT be maintained above 182 i inches, then close Ell-F011A(B).

- b. When RPV water level is restored to normal, then i 3

reopen Ell-F0llA(B).

Page 24 of 170 20P-17 Rev. 106 l

r~%

h w

b 5.0 STARTUP 5.4 Shutdown Cooline Mode (Continued) 5.4.2 Procedural Steps (Continued)

17. If suppression pool level approaches -27 inches, then perform the following:

a. With RHR to Torus Drain Controller, E11-HC-604A(B) in MANUAL, throttle closed RHR RX 2A(2B) RHR to Torus Drain Control Valve, E11-CV-F053A(B) to reduce flushing rate to the suppression pool,

b. Drain Suppression Pool to Radwaste in accordance with Section 8.7, if necessary.

j

18. While flushing RHR pump downstream piping to the suppression pool, verify RHR HX 2A(2B) outlet '

conductivity as read on E11-CIS-R001A(B) on Instrument '

Rack H21-P018(P021) is less than 10 pmho/cm, OR direct l

E&RC to sample RHR HX outlet for conductivity.

19. If RHR HX 2A(2B) shell side outlet temperature indicates less than 194*F, OR outlet conductivity is above 10 paho/cm, then perform the following:

a. Direct E&RC to sample RHR HX outlet for conductivity.

b. Continue flush until conductivity is satisfactory.

20. When BOTH of the following exist:

I

- RHR HX 2A(2B) outlet temperature as read on either l

E41-TR-R605 on Panel H12-P614, Point 3(4), or ERFIS shutdown cooling display, is within 110*F of RPV g water temperature, AND

- Conductivity is less than 10 mho/cm, Then shut down RHR HX downstream piping flush to l suppression pool as follows:

a. With RER to Torus Drain Controller., E11-HC-604A(B) l i

in MANUAL, close RHR HX 2A(2B) RHR to Torus Drain Control Valve, Ell-CV-F053A(B).

b. Close Loop A(B) LPCI Injection Check Valve Bypass Valve, Ell-V32(V33).

c. Close Loop A(B) LPCI Inboard Injection Valve, l Ell-F015A(B).

Page 25 of 170 l 20P-17 Rev. 106

A,., O l 5.0 STARTUP 5.4 Shutdown Cooline Mode (Continued) .

5.4.2 Procedural Steps (Continued)

21. Reduce RPV pressure to less than 125 psig per GP-05. l t

22. When RPV pressure is below 125 psig, then continue in i I

this procedure. ,

23. Reset Group 8 isolation logic by depressing both isolation reset push buttons on RTGB Panel P601.

R17 CAUTION Failure to complete each valve operation before performing the next sequential step for Steps 5.4.2.24 through 5.4.2.37 may cause inadvertent draining of the reactor to the Suppression Pool.

24. Verify Shutdown Cooling Inboard Suction Isolation Valve, g Ell-F009, is closed.

t

25. Open Shutdown Cooling Outboard Suction Throttle Valve, Ell-F008.

CAUTION Both loops of LPCI should NOT be isolated during the shutdown cooling mode of operation.

26. Initiate LCOs in accordance with 01-04 for both of the following:

- A(B) loop LPCI per Tech Spec 3.5.3.2, and

- A(B) loop suppression pool cooling per Tech Spec 3.6.2.2.

27. Close RHR Pump 2A & 2C(2B & 2D) Suppression Pool Suction Valve, Ell-F020A(B). l 28, close the following valves:

a. RHR Pump 2A(2B) Suppression Pool Suction Valve, Ell-F004A(B).

b. RHR Pump 2C(2D) Suppression Pool Suction Valve, E11-F004C(D).

l 20P-17 Rev. 106 Page 26 of 170 l

M_

g :s 4

5.0 STARTUP 5.4 Shutdown Cooline Mode (Continued) ,

5.4.2 Procedural Steps (Continued)

29. Open the following valves: ,

a. RHR Pump 2A(2B) Shutdown Cooling Suction Valve, E11-F006A(B).

b. RER Pump 2C(2D) Shutdown Cooling Suction Valve, ,

Ell-F006C(D).

30. Open the following valves to fill shutdown cooling suction piping: i

a. Shutdown Cooling Demineralized Water Supply Isolation Valve, Ell-F083. ,

b. Shutdown Cooling Demineralized Water Supply 1 solation Valve, Ell-F084 .

31. Monitor shutdown cooling demineralized water supply piping flow noise by listening to the pipe.

32. When flow into the shutdown cooling suction pipe stops, then close the following valves:

a. Shutdown Cooling Demineralized Water Supply Isolation Valve, Ell-F083.

b. Shutdown Cooling Demineralized Water Supply 1 solation Valve, Ell-F084

33. Close Shutdown Cooling outboard Suction Throttle Valve, Ell-F008.

34. Open Shutdown Cooling Inboard Suction Isolation Valve, Ell-F009. g I

CAUTION I

t DC Limitorque valves are limited to a duty cycle of three starts in five minutes followed by a fif ty-minute cooldown period. Any valve actuation l whether in the form of a throttle action, a continuous stroke, or an j Failure to adhere to r auto-actuated movement is considered a motor start. I the duty cycle requirements may cause DC valve motor failures. f

- 35. While closely monitoring RPV water level on Panel P603 and locally monitoring RHR piping for water hammer, slowly throttle open Shutdown Cooling Outboard Suction ,

Throttle Valve, Ell-F008.

Page 27 of 170 20P-17 Rev. 106

x I k -

e 5.0 STARTUP 5.4 Shutdown Cooline Mode (Continued) ,

5.4.2 Procedural Steps (Continued)

36. If water hammer occurs, then perform the following:

a. Close Shutdown Cooling Outboard Suction Throttle Valve, E11-F008.

b. Request Technical Support to examine affected l piping.

c. When Technical Support has determined that shutdown cooling may recommence, slowly throttle open l Shutdown Cooling Outboard Suction Throttle Valve, Ell-F008.

37. If RPV water level drops more than 10 inches, then l perform the following:

c. Close Shutdown Cooling Inboard Suction Isolation Valve, Ell-F009.

b. Close Shutdown Cooling outboard Suction Throttle Valve, E11-F008.

c. Refill Shutdown cooling suction piping by repeating Steps 5.4.2.30 through 5.4.2.36. l.

38. If RPV water level is stable and NO water hammer is apparent, then fully open Shutdown Cooling Outboard l Suction Throttle Valve, E11-F008.

39. Close RHR HX 2A(2B) Bypass Valve, Ell-F048A(B). l

40. If Loop B of RHR is being placed in Shutdown Cooling, and the alternate flush to Radwaste is desired, then perform the following:

a. N/A Steps 5.4.2.41 through 5.4.2.46.c.

b. Go to Section 8.5 of this procedure.

41. With RHR to Torus Drain Controller, E11-HC-604A(B) in l MANUAL, slowly throttle open RHR to Torus Drain Control Valve, E11-CV-F053A(B), to commence RHR pump upstream l

piping flush to the suppression pool.

20P-17 Rev. 106 Page 28 of 170 l

)

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l% s) l l

j 5,0 STARTUP 5.4 Shutdown Cooline Mode (Continued) ,

5.4.2 Procedural Steps (Continued)

42. Maintain RPV water level between 182 and 192 inches i

during flush to the suppression pool as follows:

a. If RPV water level CANNOT be maintained above 182 I

inches, then close Ell-F0llA(B).

b. When RPV water level is restored to normal, then reopen Ell-F0llA(B).

43. If suppression pool level approaches -27 inches, then i perform the following:

a. With RHR to Torus Drain Controller. Ell-HC-604A(B) in MANUAL, throttle closed RHR HX 2A(2B) RHR to Torus Drain Control Valve, Ell-CV-F053A(B) reduce flushing rate to the suppression pool.

1

b. Drain Suppression Pool to Radwaste in accordance l with Section 8.7, if necessary.

l

\

j l 44. While flushing RHR pump upstream piping to the suppression pool, verify RHR HX 2A(2B) outlet !

conductivity as read on Ell-CIS-R001A(B) on Instrument Rack H12-P018(P021) is less than 10 pmho/cm, OR direct E&RC to sample RHR HX outlet for conductivity.

45. If RER HX 2A(2B) shell side outlet temperature indicates l less than 194*F, OR outlet conductivity is above 10 l umho/cm, then perform the following:

a. Direct E&RC to sample RHR HX outlet for conductivity.

b. Continue flush until the conductivity is l satisfactory.

1 Page 29 of 170 20P-17 Rev. 106

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(_ J 5.0 STARTUP 5.4 Shutdown Coolinr. Mode (Continued) , l 1 5.4.2 Procedural Steps (Continued)

46. When BOTH of the following exist:

- RHR HX 2A(2B) inlet temperature as read on either E41-TR-R605 on Panel H12-P614, Point 1(2), or ERFIS shutdown cooling display', is within 110'F of RPV water temperature, AND

- Conductivity is less than 10 pmho/cm, Then shut down RHR pump upstream piping flush to !

suppression pool as follows:

a. Close RHR HX 2A(2B) Drain to Suppression Pool Valve, l Ell-F011A(B). l

b. With RHR to Torus Drain Controller, Ell-HC-604A(B) in MANUAL, close RER HX 2A(2B) RHR to Torus Drain Control Valve, E11-CV-F053A(B).

c. Place Control Switch, E11-541A(B), to CLOSE l position to deenergize the Air Supply to I Ell-F053A(B) RHR to Torus Drain Valve.

47. Open RHR HX 2A(2B) Bypass Valve, E11-F048A(B). l l

48. Verify RHR HX 2A(2B) Outlet Valve, E11-F003A(B), is I I

closed.

49. Fill and vent Loop A(B) of the RHR System per Section 8.1 of this procedure and return to Step 5.4.2.50.

50. When RHR Loop A(B) is filled and vented, then continue in this procedure. ,

I

51. Close RHR RX 2A(2B) Bypass Valve, E11 F048A(B). l l

52. Open Loop A(B) LPCI Inboard, Inj ection Valve , g Ell-F015A(B). ,

l

53. Close RHR HX 2A(2B) Inlet Valve, Ell-F047A(B), to minimize the hydraulic shock and initial cooldown rate created by starting the RHR pump.

20P-17 Rev. 106 Page 30 of 170 l ll

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5.0 STARTUP 5.4 Shutdown Cooline Mode (Continued) ,

5.4.2 Procedural Steps (Continued)

R17 CAUTION Failure to complete each valve operation before performing the next sequential step for Steps 5.4.2.54 to.5.4.2.55 may cause inadvertent draining of the reactor to the Suppression Pool.

l

54. Verify Loop A(B) Minimum Flow Bypass Valve, g

Ell-F007A(B), is closed.

55. Open Loop A(B) minimum flow bypass valve breaker, Compartment DFl(DL3) on 480V MCC 2XA(2XB).

56. Stop reactor recirculation pump in the loop to be i utilized for shutdown cooling per OP-02, and return to Step 5.4.2.57 of this procedure.

57. Ensure closed Reactor Recirculation Pump A(B) Discharge g Valve, B32-F031A(B).

58 Ensura open Reactor Recirculation Pump A(B) Suction valve, B32-F023A(B).

59. If recirculation loop temperature is more than 100*F I below reactor saturation temperature, then close Reactor Recirculation Pump A(B) Discharge Bypass Valve, B32-F032A(B).

60. If recirculation loop temperature is less than 100*F i below reactor saturation temperature, then ensure open Reactor Recirculation Pump A(B) Discharge Bypass Valve B32-F032A(B).

61. If one reactor recirculation pump is operating, then i establish and maintain RPV water level between 180 and g 200 inches on B21-LI-R605A(P603) or -R605B(P601).

62. If NO reactor recirculation pumps are operating, then g establish and maintain RPV water level between 200 and 220 inches on B21-LI-R605A(P603) or -R605B(P601), OR as directed by the Shift Supervisor.

63. Monitor reactor recirculation loop temperature on g B32-TR-R650(P603) to ensure reactor recirculation loop ,

heatup and cooldown rates do NOT exceed 100*F/hr. I 20P-17 Rev. 106 Page 31 of 170 l

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5.0 STARTVP t

5.4 Shutdown Coolinn Mode (Continued) ,

5.4.2 Procedural Steps (Continued) t

64. Monitor Reactor Saturation Temperature (TSAT) #" i I

determined from Reactor Steam Dome Pressure to ensure reactor vessel cooldown rates do NOT exceed 100'F/hr, OR 50*F/hr with recirculation loop temperature less than j 300*F. r i

CAUTION The maximum temperature of the RHR HX service water outier shall NOT exceed 18('F.

I

65. Monitor RHR RX SW discharge temperature on E41-TR-R605

l Point 5(6) on Panel H12-P614 to ensure temperature does not exceed 170*F.

R.11 Operation of RRR under minimum flow conditions should be minimized. l NOTE:

CAUTION ,

Failure to maintain RHR total flow rate at or above 4150 gpm will cause I overheating of the RHR pump.

r CAUTION The maximum RFV heatup or cooldown rate of 100*F in any 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period shall NOT be exceeded during heatup or cooldown with RHR System.

t CAUTION i Tbc croldown rate should be decreased to less than 50*F/hr when r+circu ation loop temperature is below 300*F.

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CAUTION l The differential temperature of 110*F between the RHP. HX outlet and l recirculation loop should NOT be exceeded during sh4tdown cooling.

t i

20P-17 Rev. 106 Page 32 of 170 l

O.

5.0 STARTUP 5.4 Shutdown Coolinz Mode (Continued) ,

5.4.2 Procedural Steps (Continued) l

\

CAUTION The RPV should NOT be cooled down below a temperature of 70*F until reactor vessel head closure studs have.been detensioned,

66. Start desired Loop A(B) RHR pump.

67. Immediately throttle open RHR HX 2A(2B) Bypass Valve, E11-F048A(B) and establish RRR total flow rate between 4500 and 6500 gpm.

68. If Reactor Recirculation Pump A(B) Discharge Bypass Valve, B32-F032A(B), is closed, then perform the following: ,

a. Bump open Reactor Recirculation Pump A(B) Discharge Bypass Valve, B32-F032A(B), as needed to maintain a recirculation loop heatup rate below 100'F/hr.

b. When reactor recirculation loop temperature is ,

j approximately equal to reactor vessel temperature, then fully open Reactor Recirculation Pump A(B4 Discharge Bypass Valve, 332-F032A(B).

CAUTION IF any of the four fuel assemblies ..urrounding aach in-core instrumentation string have been removed and NOT replaced with blade guides, THEN failure to limit RRR total flow to less than 6500 gpm may cause damage to in-core instrumentation from flow induced vibration.

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69. If an RER total flow rate between 5000 and 10,000 gpm is desired, then throttle open RHR HX 2A(2B) Bypass Valve, E11-F048A(B), to establish desired flowrate.

CAUTION RHR service water flow rate of 6000 gpm should NOT be exceeded until RPV l pressure is at or below 75 psig.

4 Page 33 of 170 l 20P-17 Rev. 106

e w

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5.0 STARTUP 5.4 Shutdown Coolint Mode (Continued) ,

5.4.2 Procedural Steps (Continued)

CAUTION During shutdown cooling, failure to maintain RHR Service Water pressure above RHR pressure may cause inleakage of radioactive water into the RHR Service Water System.

70. Throttle RHR HX 2A(2B) SW Discharge Valve, E11-PDV-F068A(B), to establish minimum Service Water flow rates to RHR HX 2A(2B) as follows:

a. IF BOTH RER Service Water Booster Pumps are operating (preferred), THEN establish 5000 to 5500 gpm flow,

b. IF one RHR Service Water Booster Pump is operating, THEN establish 2000 to 3000 gpm flow.

71. Open RHR HX 2A(2B) Inlet Valve, E11-F047A(B).

R15 NOTE: A target cooldown rate of 25*/hr is recommended by GE to minimize release of corrosion product activity during cooldown and depressurization.

72. Perform BOTH of the following to establish RPV cooldown rate and maintain desired flow rate:

a. Slowly throttle open RHR HX 2A(2B) Outlet Valve, E11-F003A(B), as necessary,

b. Slowly throttle closed RRR HX 2A(2B) Bypass Valve, E11-F048A(B), as necessary.

l i

b 20P-17 Rev. 106 i

b h

w ,8 5.0 STARTUP 5.4 Shutdown Cooline Mode (Continued) ,

5.4.2 Procedural Steps (Continued)

I

73. Monitor RPV temperature using one or more of the following methods to ensure RPV cooldown rates do NOT g exceed 100*F/hr, OR 50'F/hr with recirculation loop temperature below 300*F.

a. Reactor Saturation Temperature (T as determined from Reactor Steam Dome Pressure N ) conditions where T SAT ** "* '

b. Reactor Recirculation Suction Temperature as read on B32-TR-R650(P603) for conditions where i recirculation loop is NOT isolated with flow in that loop AND T gg is below 212'F. l

c. PRR HX 2A(2B) Inlet Temperature as read on i E41-TR-R605 Point 1(2) on Panel H12-P614 for conditions where RHR HX 2A(2B) is NOT isolated AND g T ** '

SAT

d. RHR HX 2A(2B) Outlet Temperature as read on E41-TR-R605 Point 3(4) on Panel H12-P614 for conditions where PER HX 2A(2B) is isolated AND TSAT l is below 212*F.

R10 74. Request the SRO to provide a normal temperature band to control RPV coolant temperature and log this band in the RO's log book,

75. If reactor recirculation pumps are NOT operating, AND RPV water level needs to be lowered below 200 inches for maintenance or testing with the steam separator installed, THEN pls e RHR in shutdown cooling two loop injection mode per Section 8.2 of this procedure and return to Step 5.4.2.76 of this section.

76. If shutdown cooling flow adjustments are required to maintain RPV cooldown or heatup rates, then refer to Section 8.13 as necessary.

77. Monitor normal operation of.RHR shutdown cooling per l Section 6.2 of this procedure. l NOTE: When shutdown cooling is NO longer required, RHR should be shut down per Section 7.2 of this procedure.

i j

20P-17 Rev. 106 Page 35 of IM l

w bw 5.0 STARTUP 5.4 Shutdown Cooline Mode (Continued) 5.4.2 Procedural Steps (Continued)

Date/ Time Completed Performed By (Print) Initials Reviewed By:

Shift Supervisor 20P-17 Rev. 106 Page 36 of 170

O O Figure 1 Thermal Power Limitations l

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28 % MINBIUM (19,351 k NME 30

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NATUR/ 1 CRC. LINE /

20 / / MINIMUM POl WER LINE

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15 20 25 30 35 40 CORE FLOW (MLBS/HR) 0 AOP-03.0 Rev. 5 l

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$w $w TABLE II - DETECTORS INCLUDED IN APRM CHANNEL (Cont'd)

Detector Drywell APRM Detector Drywell APRM Location Pent. Channel Location Pent. Channel

& Level No. & Level 04-37B X-100E C 12-37B X-100F D 36-37B X-100E C 44-37B X-100F D 20-21B X-100E C 28-21B X-100F D 36-05B X-100E C 28-37D X-100F D 20-37B X-100E C 12-21D X-100F D 04-21D X-100E C 44-21D X-100F D 36-21D X-100E C 28-05D X-100F D 20-05D X-100E C 20-45A X-100F D 12-45A X-100E C 04-29A X-100F D 44-45A X-100E C 36-29A X-100F D 28-29A X-100E C 20-13A X-100F D 12-13A X-100E C 36-45C X-100F D !44-13A X-100E C 20-29C X-100F D 28-45C X-100E C 36-13C X-100F D 12-29C X-100E C >

44-29C X-100E C 28-13C X-100E C Detector Drywell APRM Detector Drywell APRM Location Pent. Channel Location Pent. Channel

& Level No. & Level 12-45B X-100C E 20-45B X-100H F 44-45B X-100C E 04-29B X-100H F -

3 28-29B X-100C -E 36-29B X-100H F 12-13B X-100C E 20-13B X-100H F f 44-13D X-100C E 36-45D X-100H .F 28-45D X-100C E 20-29D X-100H F ,12-29D X-100C E 36-13D X-100H F 44-29D X-100C E 28-37A X-100H F 28-13D X-100C E 12-21A X-100H F 20-37A X-100C E 44-21A X-100H F 04-21A X-100C E 28-05A X-100H F 36-21A X-100C E 12-37C X-100H F f 20-05A X-100C E 44-37C X-100H F 04-37C X-100C E 28-21C X-100H F 36-37C X-100C E 20-21C X-100C E 36-05C X-100C E

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U. S. NUCLEAR REGULATORY COMMISSION SITE SPECIFIC EXAMINATION SENIOR OPERATOR LICENSE REGION 2 CANDIDATE'S NAME:

TACILITY: Brunswick 1 & 2 REACTOR TYPE: BWR-GE4 DATE ADMINISTERED: 94/10/24 INSTRUCTIONS TO CANDIDATE:

Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires a final grade of at least 80%. Examination papers will be picked up four (4) hours after the examination starts.

CANDIDATE'S TEST VALUE SCORE %

100.00 % TOTALS FINAL GRADE All work done on this examination is my own. I have neither given nor received aid.

Candidate's Signature

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SINIOR' REACTOR OPERATOR Page 2 ANSWER SHEET Multiple Choice (Circle or X your choice)

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

MULTIPLE CHOICE 023 a b c d I 001 a b c d 024 a b c d 002 a b c d 025 a b c d 003 a b c d 026 a b c d 1

004 a b c d 027 a b c d 005 a b c d 028 a b c d 006 a b c d 029 a b c d 007 a b c d 030 a b c d 008 a b c d 031 a b c d 009 a b c d 032 a b c d 010 a l' e d 033 a b c d 011 a b c d 034 a b c d 012 a b c d 035 a b c d 013 a b c d 036 a b c d 014 a b c d 037 a b c d 1

1 015 a b c d 038 a b c d l 016 a b c d 039 a b c d 017 a b c d 040 a b c d 018 a b c d 041 a b c d 019 a b c d 042 a b c d 020 a b c d 043 a b c d 021 a b c d 044 a b c d 022 a b c d 045 a b c d

,s ss S'ENIOR' REACTOR OPERATOR Page 3 l ANSWER SHEET Multiple Choice (Circle or X your choice)

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

l 046 a b c d 068 a b c d l l

l MULTIPLE CHOICE 069 a b c d 047 a b c d 070 a b c d l l

048 a b c d 071 a b c d 049 a b c d 072 a b c d 050 a b c d 073 a b c d 051 a b c d 074 a b c d 052 a b c d 075 a b c d 053 a b c d 076 a b c d 054 a b c d 077 a b c d 055 a b c d 078 a b c d 056 a b c d 079 a b c d 057 a b c d 080 a b c d 058 a b c d 081 a b c d l

059 a b c d 082 a b c d l l

060 a b c d 083 a b c d 061 a b c d 084 a b c d l l

062 a b c d 085 a b c d !

063 a b c d 086 a b c d 064 a b c d 087 a b c d l l

065 a b c d 088 a b c d 066 a b c d 089 a b c d 067 a b c d 090 a b c d l

l

g-g SENIOR' REACTOR OPERATOR Page 4 ANSWER SHEET Multiple Choice (Circle or X your choice)

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

091 a b c d MULTIPLE CHOICE 092 a b c d 093 a b c d 094 MATCHING a

b c

d MULTIPLE CHOICE 095 a b c d l 096 a b c d 097 a b c d 098 a b c d 099 a b c d

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

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Page 5 NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination, the following rules apply: